18th Fusion Energy Conference 4 - 10 October 2000 Sorrento, Italy (IAEA-CN-77) INTERNATIONAL ATOMIC ENERGY AGENCY Organized in co-operation with the Italian National Agency for New Technology, Energy and the Environment (ENEA) This CD-ROM contains the proceedings of the 18th International Conference on Fusion Energy 2000, held in Sorrento, Italy, 4-10 October 2000. The CD-ROM contains HTML files for navi- gation via a web browser, the papers in Portable Document Format (PDF) and Acrobat Reader 4.0 for Windows, Mac-OS and UNIX. In detail, the CD-ROM contains the following: Information on how to use this CD-ROM, fore- word and comments on this document by the producers; information about the conference organizers and site and a picture gallery; links to the Web page “WorldAtom” of the IAEA Physics Section and Fusion Energy Conference 2002 and some other web sites of interest; a list of all sessions; index lists of all submitted papers, of all authors of submitted papers and of keywords (all lists with direct links to the papers); and all submitted papers in PDF. 1
Transcript
18th Fusion Energy Conference
4 - 10 October 2000
Sorrento, Italy
(IAEA-CN-77)
INTERNATIONAL ATOMIC ENERGY AGENCY
Organized in co-operation with the
Italian National Agency for New Technology,Energy and the Environment (ENEA)
This CD-ROM contains the proceedings of the 18th International Conference on Fusion Energy2000, held in Sorrento, Italy, 4-10 October 2000. The CD-ROM contains HTML files for navi-gation via a web browser, the papers in Portable Document Format (PDF) and Acrobat Reader4.0 for Windows, Mac-OS and UNIX.
In detail, the CD-ROM contains the following: Information on how to use this CD-ROM, fore-word and comments on this document by the producers; information about the conferenceorganizers and site and a picture gallery; links to the Web page “WorldAtom” of the IAEAPhysics Section and Fusion Energy Conference 2002 and some other web sites of interest; a listof all sessions; index lists of all submitted papers, of all authors of submitted papers and ofkeywords (all lists with direct links to the papers); and all submitted papers in PDF.
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Contents
Foreword and Information about the CD-ROM 5
OPENING
Session AKMS — Artsimovich-Kadomtsev Memorial Session 7
Session OV — Overviews 9
OVERVIEWS
Session OV1 — Magnetic Fusion Overview 1 12
Session OV2 — Magnetic Fusion Overview 2 20
Session OV3 — Inertial Fusion Overview 25
Session OV4 — Magnetic Fusion Overview 3 29
Session OV5/EX1/TH1 — Current Drive, Heating and Fuelling 36
Session THP1 — Turbulence, Transport & Edge Physics 374
Session THP2 — MHD, Energetic Particles & Current Drive 402
POST DEADLINE PAPERS
Session PD — Post Deadline Papers 430
Session PDP — Post Deadline Posters 435
SUMMARIES
Session S — Summaries 444
INDEX PAGES
Paper Index 449
Keyword Index 453
Author Index 458
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Foreword and Information aboutthe CD-ROM
Foreword
The 18th International Atomic Energy Agency Fusion Energy Conference was held in Sorrento,Italy, 4–10 October 2000. This series of meetings, which began in 1961, has been held bienniallysince 1974. In addition to these biennial conferences, the IAEA organizes co-ordinated researchprojects (CRPs), technical committee meetings (TCMs), advisory group meetings (AGMs) andconsultants meetings (CMs) on fusion research topics. The objectives of the IAEA activitiesrelated to fusion research are to:
• Promote fusion energy development and worldwide collaboration, such as the InternationalThermonuclear Experimental Reactor (ITER);
• Support Member State activities in fusion research;
• Emphasize the safety and environmental advantages of fusion energy;
• Encourage the utilization of plasmas and fusion technology in industry and applications.
This 5 1/2-day Fusion Energy Conference, which was attended by 573 participants from overthirty countries and three international organizations, was organized by the IAEA in co-operationwith the Italian National Agency for New Technology, Energy and the Environment (ENEA),to which the IAEA wishes to express its gratitude. A total of 397 papers, including the post-deadline papers, were accepted for presentation at the conference by the Programme Committee,which met in May 2000 and on the second day of the conference. Overall, 389 papers, includingthe 4 summary talks, were presented in 22 oral and 8 poster sessions on magnetic confinementexperiments, inertial fusion energy, plasma heating and current drive, ITER engineering de-sign activities, magnetic confinement theory, innovative concepts, fusion technology, and safetyand environmental aspects. The Opening Session was designated the Artsimovich-KadomtsevMemorial Session, in honour of Academicians L.A. Artsimovich and B.B. Kadomtsev. Threeinvited lectures were given in the Opening Session by Nobel Prize Winner C. Rubbia, Presi-dent, ENEA, Rome, Italy; A. Arima, Member of the House of Councilors, Tokyo, Japan; and E.Velikhov, President, Russian Research Centre “Kurchatov Institute”, Moscow, Russia.
Using the CD-ROM
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Editorial Note
This publication has been prepared from the original material as submitted by the authors. Theviews expressed do not necessarily reflect those of the IAEA, the governments of the nominatingMember States or the nominating organizations. The use of particular designations of countriesor territories does not imply any judgement by the publisher, the IAEA, as to the legal statusof such countries or territories, of their authorities and institutions or of the delimitation oftheir boundaries. The mention of names of specific companies or products (whether or notindicated as registered) does not imply any intention to infringe proprietary rights, nor should itbe construed as an endorsement or recommendation on the part of the IAEA. The authors areresponsible for having obtained the necessary permission for the IAEA to reproduce, translateor use material from sources already protected by copyrights.
(OV/1) ITER-FEAT - The Future International BurningPlasma Experiment - Overview
R. Aymar1), V. Chuyanov1), M. Huguet2), Y. Shimomura2) for the ITER Joint Central Teamand Home Teams
1) ITER Garching Joint Work Site, Boltzmannstrasse 2, 85748 Garching, Germany2) ITER Naka Joint Work Site, Naka-machi, Ibaraki-ken, Japan
Abstract. The focus of effort in the ITER Engineering Design Activities (EDA) since 1998 hasbeen the development of a new design to meet revised technical objectives and a cost reductiontarget of about 50% of the previously accepted cost estimate. Drawing on the design solu-tions already developed and using the latest physics results and outputs from technology R&Dprojects, the Joint Central Team and Home Teams, working jointly, have been able to convergetowards a new design which will allow the exploration of a range of burning plasma conditions,with a capacity to progress towards possible modes of steady state operation. As such the newITER design, whilst having reduced technical objectives from its predecessor, will nonethelessmeet the programmatic objective of providing an integrated demonstration of the scientific andtechnological feasibility of fusion energy. The main features of the current design and of its pro-jected performance are introduced and the outlook for construction and operation is summarised.
Abstract. An overview of the FTU results during the period 1998–2000 is presented. FTU hasoperated up to the nominal parameters (B=8T, I=1.6MA) with good reliability. Using the high-speed, multiple pellet injection system in 8T/1.25MA discharges a phase lasting a few energyconfinement time has been achieved with improved confinement properties with peaked densityprofiles (n = 4× 1020m−3), Te ≈ Ti and Zeff ≈ 1.3. Up to 14 keV of electron temperaturehave been obtained at high density using electron cyclotron resonance heating (ECRH) on thecurrent ramp. The transport analysis shows a very low electron heat transport in the regionwith flat/hollow safety factor profile. Synergy studies have been performed with simultaneousinjection of lower hybrid and electron cyclotron waves in 7.2T discharges, well above the ECRHresonance. Clear evidence has been obtained of electron cyclotron wave absorption by the lowerhybrid produced electron tails. Stabilisation of m=2 tearing modes has been obtained usingECRH, with subsequent improvement of the energy confinement. Ion Bernstein wave injectionin high magnetic field (B=8T) plasmas has shown the reduction of the electron thermal con-ductivity in the region inside the absorption radius possibly due to the formation of an internaltransport barrier.
Y. Kamada1), H. Adachi1), H. Akasaka1), N. Akino1), K. Annou1), T. Arai1), K. Arakawa1),N. Asakura1), M. Azumi1), P. E. Bak1), M. Bakhtiari1), C. Z. Cheng2), S. Chiba1), Z. Cui3),S. A. Dettrick4), N. Ebihara1), G. Y. Fu2), T. Fujii1), T. Fujita1), H. Fukuda1), T. Fukuda1),A. Funahashi1), H. Furukawa1), L. R. Grisham4), K. Hamamatsu1), T. Hamano1), T. Hatae1),A. Hattori1), N. Hayashi1), S. Higashijima1), S. Hikida1), K. Hill2), S. Hiranai1), H. Hirat-suka1), H. M. Hoek1), A. Honda1), M. Honda1), Y. Hoshi1), N. Hosogane1), L. Hu5), H. Ichige1),S. Ide1), Y. Idomura1), K. Igarashi1), Y. I. Ikeda1), A. Inoue1), M. Isaka1), A. Isayama1),N. Isei1), S. Ishida1), K. Ishii1), Y. Ishii1), T. Ishijima6), M. Ishikawa6), A. Ishizawa1), K. Itami1),T. Itoh1), T. Iwahashi1), K. Iwasaki1), M. Iwase1), K. Kajiwara1), E. Kajiyama1), Y. Kamada1),A. Kaminaga1), T. Kashiwabara1), M. Kawai1), Y. Kawamata1), Y. Kawano1), M. Kazawa1),H. Kikuchi1), M. Kikuchi1), T. Kimura1), Y. Kishimoto1), S. Kitamura1), A. Kitsunezaki1),K. Kizu1), K. Kodama1), Y. Koide1), M. Koiwa1), S. Kokusen1), T. Kondoh1), S. Konoshima1),G. J. Kramer2), H. Kubo1), K. Kurihara1), G. Kurita1), M. Kuriyama1), Y. Kusama1), N. Ku-sanagi1), L. L. Lao7), P. Lee3), S. Lee1), A. W. Leonard7), J. Li5), M. A. Mahdavi7), J. Man-ickam2), K. Masaki1), H. Masui1), T. Matsuda1), M. Matsukawa1), T. Matsumoto1), D. R. Mikkelsen2),M. Z. Mironov8), Yukitoshi Miura1), Yushi Miura1), N. Miya1), K. Miyachi1), H. Miyata1),K. Miyata1), Y. Miyo1), T. Miyoshi1), K. Mogaki1), M. Morimoto1), A. Morioka1), S. Moriyama1),K. Nagashima1), S. Nagaya1), O. Naito1), Y. Nakamura1), T. Nakano1), R. Nazikian2), M. Nemoto1),S. V. Neudatchin9), Y. Neyatani1), H. Ninomiya1), T. Nishitani1), H. Nobusaka1), M. Noda1),T. Oba1), T. Ohga1), K. Ohshima1), A. Oikawa1), T. Oikawa1), M. Okabayashi2), T. Okabe1),J. Okano1), K. Omori1), S. Omori1), Y. Omori1), H. Oohara1), T. Oshima1), N. Oyama1),T. Ozeki1), T. W. Petrie7), G. Rewoldt2), J. A. Romero1), N. Sakamoto1), A. Sakasai1), S. Sakata1),T. Sakuma1), S. Sakurai1), T. Sasajima1), N. Sasaki1), M. Sato1), M. Seimiya1), H. Seki1),M. Seki1), Y. Shibata1), K. Shimada1), M. Shimada1), K. Shimizu1), M. Shimizu1), M. Shi-mono1), K. Shinohara1), S. Shinozaki1), H. Shirai1), M. Shitomi1), X. Song5), M. Sueoka1),A. Sugawara1), T. Sugie1), H. Sunaoshi1), Masaei Suzuki1), Mitsuhiro Suzuki1), S. Suzuki1),T. Suzuki1), Y. Suzuki1), M. Takahashi1), S. Takahashi1), S. Takano1), M. Takechi1), S. Takeji1),H. Takenaga1), Y. Taki1), T. Takizuka1), H. Tamai1), Y. Tanai1), T. Terakado1), M. Terakado1),K. Tobita1), S. Tokuda1), T. Totsuka1), R. Toyokawa1), K. Tsuchiya1), T. Tsuda1), T. Tsugita1),Y. Tsukahara1), K. Uehara1), T. Uehara1), N. Umeda1), Y. Uramoto1), H. Urano10), K. Ushi-gusa1), K. Usui1), S. Wang5), J. Yagyu1), M. Yamaguchi1), Y. Yamashita1), H. Yamazaki1),K. Yokokura1), I. Yonekawa1), H. Yoshida1), R. Yoshino1)
1) 1)Japan Atomic Energy Research Instutute, Naka Fusion Research Establishment, Japan2) Princeton Plasma Physics Laboratory, USA3) South Western Institute of Physics, China4) Australian National University, Australia5) Academia Sinica, China6) Tsukuba University, Japan7) General Atomics, USA8) Ioffe Institute, RF9) Kurchatov Institute, RF10) Hokkaido University, Japan
Abstract. With the main aim of providing physics basis for ITER and the steady-state toka-mak reactor, JT-60U has been optimizing operational concepts and extending discharge regimestoward simultaneous sustainment of high confinement, high βN, high bootstrap fraction, fullnoninductive current drive and efficient heat and particle exhaust utilizing variety of heating,current drive, torque input and particle control capabilities. In the two advanced operationregimes, the reversed magnetic shear (RS) and the weak magnetic shear (high-βp) ELMy Hmodes characterized by both internal (ITB) and edge transport barriers and high bootstrap cur-
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rent fractions fBS, discharges have been sustained near the steady-state current profile solutionsunder full noninductive current drive with proper driven current profiles (High-βp; HHy2 ∼ 1.4and βN ∼ 2.5 with N-NB, RS; HHy2 ∼ 2.2 and βN ∼ 2 with fBS ∼ 80%). Multiple pelletinjection has extended the density region with high confinement. These operational modes havebeen extended to the reactor relevant regime with small values of collisionality and normalizedgyroradius and Te ∼ Ti. In the RS regime, Qeq
DT = 0.5 has been sustained for 0.8s. Stabilityhas been improved in these regimes by suppression of the neoclassical tearing mode with localECCD and enhanced βN-values with wall stabilization. The ITB structure has been controlledby toroidal rotation profile modification and transport studies have revealed a semi-global na-ture of the ITB structure. The both-leg divertor pumping has enhanced He exhaust by ∼ 40%.Ar-puff experiments have improved confinement at high density with detached divertor due tohigh pedestal temperature Ti-ped. In H-modes, the core confinement degraded with decreasingTi-ped suggesting stiff core profiles. The operational region of grassy ELMs with small divertorheat load has been established at high triangularity, high q95 and high βp. The record valueof the neutral beam current drive efficiency of 1.55 × 1019A/m2/W has been demonstrated byN-NB. Abrupt large amplitude events causing neutron drop have been discovered with frequencyinside the TAE gap. Disruption studies have clarified that runaway current is terminated byMHD fluctuations when the surface q becomes 2 ∼ 3.
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(OV1/2) Overview of JET Results in Support of the ITERPhysics Basis
C. Gormezano1), the JET Team2)
1) Associazone EURATOM-ENEA sulla Fusione, Frascati, Italy2) Jet Joint Undertaking, Abingdon, England
Abstract. The JET experimental campaign has focused on studies in support of the ITERphysics basis. An overview of the results obtained is given both for the reference ITER scenario,the ELMy H-mode, and for advanced scenarios which in JET are based on Internal TransportBarriers. JET studies for the ELMy H-mode have been instrumental for the definition of ITER-FEAT. Positive elongation and current scaling in the ITER scaling law have been confirmed, butthe observed density scaling fits better a two term (core and edge) model. Significant progressin neo-classical tearing mode limits has been made showing that ITER operation seems to beoptimised. Effective helium pumping and divertor enrichment is found to be well within ITERrequirements. Target asymmetries and H-isotope retention are well simulated by modelling codestaking into account drift flows in the scrape-off plasmas. Striking improvements in fuelling ef-fectiveness have been found with the new high field pellet launch facility. Good progress hasbeen made on scenarios for achieving good confinement at high densities, both with RI modesand with high field side pellets. Significant development of advanced scenarios in view of theirapplication to ITER has been achieved. Integrated advanced scenarios are in good progress withedge pressure control (impurity radiation). An access domain has been explored showing in par-ticular that the power threshold increases with magnetic field but can be significantly reducedwhen Lower Hybrid current drive is used to produce target plasma with negative shear. The roleof ion pressure peaking on MHD has been well documented. Lack of sufficient additional heatingpower and interaction with the septum at high beta prevents assessment of beta limits (steadyplasmas achieved with βN up to 2.6). Plasmas with non-inductive current (INI/Ip = 60%), wellaligned with plasma current, high beta and good confinement have also been obtained.
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(OV1/3) Overview of Recent Experimental Results fromthe DIII-D Advanced Tokamak Program
S. L. Allen1), and the DIII-D Team2)
1) Lawrence Livermore National Laboratory, Livermore, California USA2) General Atomics, P.O. Box 85608, San Diego, California USA
Abstract. The goals of DIII-D Advanced Tokamak (AT) experiments are to investigate andoptimize the upper limits of energy confinement and MHD stability in a tokamak plasma, and tosimultaneously maximize the fraction of non-inductive current drive. Significant overall progresshas been made in the past 2 years, as the performance figure of merit βNH89P of 9 has beenachieved in ELMing H-mode for over 16τE without sawteeth. We also operated at βN ∼ 7 for over35 τE or 3τR, with the duration limited by hardware. Real-time feedback control of β (at 95% ofthe stability boundary), optimizing the plasma shape (e.g., δ, divertor strike- and X-point, dou-ble/single null balance), and particle control (ne/nGW ∼ 0.3, Zeff < 2.0) were necessary for thelong-pulse results. A new quiescent double barrier (QDB) regime with simultaneous inner- andedge- transport barriers and no ELMs has been discovered with βNH89P of 7. The QDB regimehas been obtained to date only with counter neutral beam injection. Further modification andcontrol of internal transport barriers (ITBs) has also been demonstrated with impurity injection(broader barrier), pellets, and ECH (strong electron barrier). The new Divertor-2000, a keyingredient in all these discharges, provides effective density, impurity and heat flux control inthe high-triangularity plasma shapes. Discharges at ne/nGW ∼ 1.4 have been obtained with gaspuffing by maintaining the edge pedestal pressure; this operation is easier with Divertor-2000.We are developing several other tools required for AT operation, including real-time feedbackcontrol of resistive wall modes (RWMs) with external coils, and control of neoclassical tearingmodes (NTMs) with electron cyclotron current drive (ECCD).
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(OV1/4) Overview of LHD Experiments
M. Fujiwara1), K. Kawahata1), N. Ohyabu1), O. Kaneko1), A. Komori1), H. Yamada1), N. Ashikawa2),L. R. Baylor8), S. K. Combs8), P. de Vries1), M. Emoto1), A. Ejiri4), P. W. Fisher8), H. Fun-aba1), M. Goto1), D. Hartmann9), K. Ida1), H. Idei1), S. Iio5), K. Ikeda1), S. Inagaki1), N. In-oue1), M. Isobe1), S. Kado4), K. Khlopenkov1), T. Kobuchi2), A. V. Krasilnikov10), S. Kubo1),R. Kumazawa1), F. Leuterer9), Y. Liang2), J. F. Lyon8), S. Masuzaki1), T. Minami1), J. Miya-jima, T. Morisaki1), S. Morita1), S. Murakami1), S. Muto1), T. Mutoh1), Y. Nagayama1),N. Nakajima1), Y. Nakamura1), H. Nakanishi1), K. Narihara1), K. Nishimura1), N. Noda1),T. Notake3), S. Ohdachi1), Y. Oka1), S. Okajima6), M. Okamoto1), M. Osakabe1), T. Ozaki1),R. O. Pavlichenko1), B. J. Peterson1), A. Sagara1), K. Saito3), S. Sakakibara1), R. Sakamoto1),H. Sanuki1), H. Sasao2), M. Sasao1), K. Sato1), M. Sato1), T. Seki1), T. Shimozuma1), M. Shoji1),H. Sugama1), H. Suzuki1), M. Takechi1), Y. Takeiri1), N. Tamura1), K. Tanaka1), K. Toi1),T. Tokuzawa1), Y. Torii3), K. Tsumori1), K. Y. Watanabe1), T. Watanabe1), T. Wateri1),I. Yamada1), S. Yamaguchi1), S. Yamamoto3), M. Yokoyama1), N. Yoshida7), Y. Yoshimura1),Y. Zhao11), R. Akiyama1), K. Haba1), M. Iima1), J. Kodaira1), T. Takita1), T. Tsuzuki1),K. Yamauchi1), H. Yonezu1), H. Chikaraishi1), S. Hamaguchi1), S. Imagawa1), N. Inoue1),A. Iwamoto1), S. Kitagawa1), Y. Kubota1), R. Maekawa1), T. Mito1), K. Murai1), A. Nishimura1),K. Takahata1), H. Tamura1), S. Yamada1), N. Yanagi1), K. Itoh1), K. Matsuoka1), K. Ohkubo1),I. Ohtake1), S. Satoh1), T. Satow1), S. Sudo1), S. Tanahashi1), K. Yamazaki1), Y. Hamada1),O. Motojima1)
1) National Institute for Fusion Science, Oroshi-cho 322-6, Toki 509-5292, Japan2) Graduate University for Advanced Studies, Hayama 240-0193, Japan3) Dep. of Energy Eng. and Science, Nagoya University, Nagoya 464-8603, Japan4) University of Tokyo, Tokyo 113, Japan5) Tokyo Institute of Technology, Meguro-ku, Tokyo 152-8550, Japan6) Chubu University, Kasugai-shi 487-8501, Japan7) Kyushu University, Kasuga-shi 816-8580, Japan8) Oak Ridge National Laboratory, Oak Ridge, Tennessee, 37831-8072 USA9) Max Plank Institute for Plasma Physics, D-85748, Garching, Germany10) Troitsk Institute of Nuclear Physics (TRINITI), Troitsk, Russia11) Institute of Plasma Physics, Academia Scinica, 230031, Hefei, Anhui, China
Abstract. Experimental studies on the Large Helical Device during the last two years arereviewed. After the start of LHD experiment in 1998, the magnetic field has been graduallyraised up to 2.89 T. The heating power has been increased, up to 4.2 MW for NBI, 1.3 MW forICRF, and 0.9 MW for ECRH. Upgrading the key hardware systems has led to the extensionof the plasma parameters to (i) higher Te [ Te(0) = 4.4 keV at 〈ne〉 = 5.3 × 1018m−3 andPabs = 1.8MW ], (ii) higher confinement [ τE = 0.3 s, Te(0) = 1.1 keV at 〈ne〉 = 6.5× 1019m−3
and Pabs = 2.0MW ] and (iii) higher stored energy Wdiap = 880kJ. High performance plasmas
have been realized in the inward shifted magnetic axis configuration (R=3.6m) where the heli-cal symmetry is recovered and the particle orbit properties are improved by trade off of MHDstability properties due to the appearance of the magnetic hill. The energy confinement wassystematically higher than that predicted by the International Stellerator Scaling 95 up to afactor of 1.6 and was comparable with ELMy H-mode confinement capability in tokamaks. Thisconfinement improvement is attributed to the configuration control (the inward shift of mag-netic axis) and to the formation of the high edge temperature. The achieved average beta valuereached 2.4 % at B=1.3 T, the highest beta value ever obtained in helical devices, and so far nodegradation of confinement by MHD phenomenon is observed. The inward shifted configurationhas also led to successful ICRF minority ion heating. ICRF power up to 1.3 MW was reliablyinjected into the plasma without significant impurity contamination and a plasma with a storedenergy of 200 kJ was sustained for 5 sec by ICRF alone. As another important result long pulsedischarges of more than 1 minute were successfully achieved separately with NBI heating of 0.5
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MW and with ICRF heating of 0.85 MW.
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(OV1/D) Discussion of Section OV1
The file contains the discussion contributions relating to OV1/1, OV1/2, OV1/3, OV1/4.
Abstract. Ion and electron temperatures in conventional H mode on ASDEX Upgrade are stiffand limited by a critical temperature gradient length ∆T/T as given by ion temperature gradient(ITG) driven turbulence. ECRH experiments indicate that Te profiles are also stiff as predictedby ETG turbulence with streamers. Accordingly, core and edge temperatures are proportionalto each other and plasma energy is proprtional to pedestal pressure for fixed density profiles.Density profiles are not stiff, and confinement improves with density peaking. Higher triangularshapes (δ < 0.45) show strongly improved confinement up to Greenwald density nGW due to in-creasing pedestal pressure, and H-mode density operation extends above nGW. Density peakingat nGW was achieved with controlled gas puff rates and first results from higher high field sidepellet velocities are promising. At nGW small type II ELMs provide good confinement with lowdivertor power loading. In advanced scenarios highest performance was achieved in improvedH-modes with HL−89PβN ≈ 7.2 at δ = 0.3, limited by neo-classical tearing modes (NTM) atlow central shear (qmin ≈ 1). The T profiles are still governed by ITG/TEM turbulence andconfinement is improved by density peaking. Ion internal transport barriers (ITB) dischargeswith reversed shear and L-mode edge are limited to βN ≤ 1.7 by ideal MHD modes and gotHL−89P ≤ 2.1. Turbulent transport is suppressed in agreement with ExB shear flow paradigm,and transport coefficients are at neo-classical ion transport level. Reactor-relevant ion and elec-tron ITBs with Te ≈ Ti ≈ 10keV were achieved by combining ion and electron heating (NI,ECRH). Full non-inductive current drive was achieved in integrated high preformance H-modescenario with ne = nGW, βp = 3.1 and HL−89P = 1.8, which developed ITBs with qmin ≈ 1.Central co-ECCD at low densities allowed high current drive fraction of > 80%, while counter-ECCD leads to negative central shear and electron ITB with Te(0) > 10 keV. MHD phenomena,especially fishbones, contribute to achieve quasi-stationary advanced discharge conditions andtrigger ITBs,. but also limit the operation of conventional and advanced scenarios. CompleteNTM stabilisation has been demonstrated using ECCD with 10% of heating power. Extension ofMHD limits is expected from using off-axis CD (tangential NI) and wall stabilisation. Presently,divertor shape is adapted to higher δ’s and tungsten covering of first wall is extended based onthe positive experience using tungsten on divertor and heat shield tiles.
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(OV2/2) Overview of Alcator C-Mod Recent Results
I. H. Hutchinson1), R. L. Boivin1), P. T. Bonoli1), C. Boswell1), R. Bravenec3), N. Bretz2),R. Chatterjee3), T. Chung1), E. Eisner3), C. Fiore1), S. Gangadhara1), K. Gentle3), J. A. Goetz1),R. S. Granetz1), M. J. Greenwald1), J. C. Hosea2), A. Hubbard1), J. Hughes1), Y.In) 5), J. Irby1),B. LaBombard1), Y. Lin1), B. Lipschultz1), R. Maqueda6), E. S. Marmar1), A. Mazurenko1),E. Nelson-Melby1), D. R. Mikkelsen2), D. Mossessian1), R. Nachtrieb1), R. Nazikian2), D. Pap-pas1), R. R. Parker1), T. Pedersen1), P. Phillips3), C. S. Pitcher1), M. Porkolab1), J. Rice1),W. Rowan3), G. Schilling2), J. A. Snipes1), G. Taylor2), J. L. Terry1), J. R. Wilson2), S. Wolfe1),S. Wukitch1), H. Yuh1), S. Zweben2), P. Acedo10), M. Brambilla4), B. A. Carreras8), R. Gandy5),G. A. Hallock3), W. Dorland7), D. Johnson2), N. Krashennikova1), C. K. Phillips2), T. Tutt1),C. Watts9), M. Umansky1)
1) MIT Plasma Science and Fusion Center, Cambridge, MA, USA 021392) Princeton Plasma Physics Lab., Princeton, NJ, USA 085433) University of Texas, Fusion Research Center, Austin, TX, USA 787124) Max Planck Institute for Plasma Physics, Garching, Germany5) University of Idaho, Moscow, ID, USA6) Los Alamos National Laboratory, Los Alamos, New Mexico, USA 875457) University of Maryland, College Park, MD, USA 20742-35118) Oak Ridge National Laboratory, Fusion Energy Division, Oak Ridge,TN, USA 378319) Auburn University, Auburn, Alabama, USA 3684910) Universidad Carlos III de Madrid, 28911 Leganes Madrid, Spain.
Abstract. Research on Alcator C-Mod tokamak focusses on exploiting compact, high den-sity plasmas to understand core transport and heating, the physics of the H-mode transportbarrier, and the dynamics of the scrape-off-layer and divertor. Rapid toroidal acceleration ofthe plasma core is observed during ohmic heated H-modes and indicates a momentum pinch orsimilar transport mechanism. Core thermal transport observations support a critical gradientinterpretation, but with gradients that disagree with present theoretical values. High resolutionmeasurements of the H-mode barrier have been obtained including impurity and neutral den-sities, and the instability apparently responsible for the favorable “Enhanced D-alpha” regimehas been identified. Divertor bypass dynamic control experiments have directly addressed theimportant questions surrounding main chamber recycling and the effect of divertor closure onimpurities and confinement. Future plans include quasi-steady-state Advanced Tokamak plas-mas using Lower Hybrid current drive.
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(OV2/3) Ergodic Divertor Experiments on the Route toSteady-state Operation of Tore Supra
P. Ghendrih1), TORE SUPRA Team
1) Association Euratom-CEA sur la Fusion, CEA Cadarache, 13108 St Paul lez Durance, France
Abstract. Ergodic Divertor operation on Tore Supra is characterised by good performance interms of divertor physics. Control of particle recirculation and impurity screening are relatedto the symmetry both poloidally and toroidally of the shell of open field lines and to its radialextent, ∆r ∼ 0.16m. Feedback control of the divertor plasma temperature has led to controlledradiative divertor experiments. In particular, good performance is obtained when the plasma iscontrolled to be a temperature comparable to the energy involved in the atomic processes, (15to 20 eV). For standard discharges with 5 MW total power and ICRH heating, the low parallelenergy flux ∼ 10 MW m−2 is reduced to ∼ 3MWm−2 with nitrogen injection. This is achievedat a modest cost in core dilution, ∆Zeff ∼ 0.3. Despite the large volume of open field lines(∼ 36%) the Ergodic Divertor does not reduce the possible current in the discharge since stabledischarges are achieved with qsep ∼ 2. It is shown that the reorganisation of the current profilein conjunction with a transport barrier in the electron temperature on the separatrix stabilisesthe (2,1) tearing mode. Confinement follows the standard L-mode confinement. In a few casesat high density and with no gas injection (wall fuelled discharges), “RI-like” modes are reportedwith modest increase in confinement (∼ 40%). Despite the lack of core fuelling on Tore Supra,high densities during ICRH pulses can be achieved with Greenwald fractions fG ∼ 1. Compati-bility with both ICRH and LH is demonstrated. In particular long pulse operation with flat topin excess of 20 s are achieved with LHCD.
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(OV2/D) Discussion of Section OV2
The file contains the discussion contributions relating to OV2/1, OV2/2, OV2/3.
(OV3/2) Advances of Direct Drive Schemes in Laser FusionResearch at ILE Osaka
T. Yamanaka1), K. Mima1), K. A. Tanaka2), Y. Kitagawa1), R. Kodama1), N. Izumi1), Y. Sen-toku1), H. Azechi1), K. Nishihara1), H. Shiraga1), M. Nakai1), K. Fujita1), K. Shigemori1),M. Heya3), Y. Ochi1), T. Norimatsu1), K. Nagai1), Y. Izawa1), M. Nakatsuka1), N. Miyanaga1),H. Fujita1), T. Jitsuno1), S. Sakabe2), H. Takabe1), M. Murakami1), H. Nagatomo1), A. Suna-hara1), T. Kawamura1), S. Nakai2), and C. Yamanaka4)
1) Institute of Laser Engineering, Osaka University, Osaka, Japan2) Faculty of Engineering, and Institute of Laser Engineering, Osaka University, Osaka, Japan3) Institute of Free Electron Laser, Osaka University, Osaka, Japan4) Institute for Laser Technology, Osaka, Japan
Abstract. ILE Osaka is concentrating on the physical elements of fast ignition aiming at theproof of principle for ignition-and-burn of direct-drive laser fusion. A 1PW laser will be intro-duced to fast ignition experiments by the middle of 2001. A high intensity plasma experimentalresearch system, HIPER, has been in operation for obtaining scientific data base relevant toignition target. By irradiating a intense short pulse onto a long scale length plasma observedare penetration of a relativistically self-focused laser beam into over-dense region without con-siderable energy loss in under-dense region, MeV electrons generation with conversion efficiencyof 25%, heating of compressed core plasma by irradiating a 100 ps, 1017W/cm2 pulse and 1 ps,1019W/cm2 pulse. In the hydrodynamic instability, the initial imprint of hydrodynamic insta-bility and Rayleigh-Taylor growth rate at wavelength less than 10 µm have been investigatedextensively.
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(OV3/3) The Heavy Ion Fusion Program in the USA
R. O. Bangerter1), R. C. Davidson2), W. B. Herrmannsfeldt3), J. Lindl4), B. G. Logan1)4),W. R. Meier4)
1) Lawrence Berkeley National Laboratory2) Princeton Plasma Physics Laboratory3) Stanford Linear Accelerator4) Lawrence Livermore National Laboratory1)2)4) Heavy Ion Fusion Virtual National Laboratory
Abstract. Inertial fusion energy research has enjoyed increased interest and funding. This hasallowed expanded programs in target design, target fabrication, fusion chamber research, tar-get injection and tracking, and accelerator research. The target design effort examines ways tominimize the beam power and energy and increase the allowable focal spot size while preservingtarget gain. Chamber research for heavy ion fusion emphasizes the use of thick liquid walls toserve as the coolant, breed tritium, and protect the structural wall from neutrons, photons, andother target products. Several small facilities are now operating to model fluid chamber dynam-ics. A facility to study target injection and tracking has been built and a second facility is beingdesigned. Improved economics is an important goal of the accelerator research. The acceleratorresearch is also directed toward the design of an Integrated Research Experiment (IRE). TheIRE is being designed to accelerate ions to >100 MeV, enabling experiments in beam dynamics,focusing, and target physics. Activities leading to the IRE include ion source development anda High Current Experiment (HCX) designed to transport and accelerate a single beam of ionswith a beam current of approximately 1 A, the initial current required for each beam of a fusiondriver. In terms of theory, the program is developing a source-to-target numerical simulationcapability. The goal of the entire program is to enable an informed decision about the promiseof heavy ion fusion in about a decade.
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(OV3/5) OMEGA ICF Experiments and Preparations forDirect Drive on NIF
R. L. McCrory1), R. E. Bahr1), R. Betti1), T. R. Boehly1), T. J. B. Collins1), R. S. Craxton1),J. A. Delettrez1), W. R. Donaldson1), R. Epstein1), J. Frenje2), V. Yu. Glebov1), V. N. Gon-charov1), O. V. Gotchev1), R. Q. Gram1), D. R. Harding1), D. G. Hicks∗2), P. A. Jaanimagi1),R. L. Keck1), J. Kelly1), J. P. Knauer1), C. K. Li2), S. J. Loucks1), L. D. Lund1), F. J. Mar-shall1), P. W. McKenty1), D. D. Meyerhofer1), S. F. B. Morse1), R. D. Petrasso2), P. B. Radha1),S. P. Regan1), S. Roberts1), F. Seguin2), W. Seka1), S. Skupsky1), V. A. Smalyuk1), C. Sorce1),J. M. Soures1), C. Stoeckl1), R. P. J. Town1), M. D. Wittman1), B. Yaakobi1), and J. D. Zuegel1)
1) Laboratory for Laser Energetics, University of Rochester, Rochester, NY, USA2) Plasma Science and Fusion Center, MIT, Boston, MA, USA∗ Currently at Lawrence Livermore National Laboratory, Livermore, CA, USA
Abstract. Direct-drive laser-fusion ignition experiments rely on detailed understanding andcontrol of irradiation uniformity, the Rayleigh-Taylor instability, and target fabrication. LLE isinvestigating various theoretical aspects of a direct-drive NIF ignition target based on an “all-DT” design: a spherical target of ∼3.4-mm diameter, 1 to 2 µm of CH wall thickness, and an∼340-µm DT-ice layer near the triple point of DT (∼19 K). OMEGA experiments are designedto address the critical issues related to direct-drive laser fusion and to provide the necessarydata to validate the predictive capability of LLE computer codes. The cryogenic targets to beused on OMEGA are hydrodynamically equivalent to those planned for the NIF. The currentexperimental studies on OMEGA address the essential components of direct-drive laser fusion:irradiation uniformity and laser imprinting, Rayleigh-Taylor growth and saturation, compressedcore performance and shell fuel mixing, laser plasma interactions and their effect on target per-formance, and cryogenic target fabrication and handling.
1) EURATOM/UKAEA Fusion Association, Culham Science Centre, Abingdon, Oxfordshire,OX14 3DB
Abstract. MAST is one of the new generation of large, purpose-built Spherical Tokamaksnow becoming operational, designed to investigate the properties of the ST in large, collisionlessplasmas. The first 6 months of MAST operations have been remarkably successful. Opera-tionally, both merging-compression and the more usual solenoid induction schemes have beendemonstrated, the former providing over 400kA of plasma current with no demand on solenoidflux. Good vacuum conditions and operational conditions, particularly after boronisation intrymethylated boron, have provided plasma current of over 1MA with central plasma temper-atures (Ohmic) of order 1keV. The Hugill and Greenwald limits can be exceeded, and H-modeachieved at modest additional NBI power. Moreover, particle and energy confinement show animmediate increase at the L-H transition, unlike START where this only became apparent at thehighest plasma currents. Halo currents are small, with low toroidal peaking factors, in accor-dance with theoretical predictions, and there is evidence of a resilience to the major disruption.
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(OV4/2) Overview of the Initial NSTX Experimental Re-sults
M. Ono1), M. Bell1), R. Bell1), T. Bigelow2), M. Bitter1), W. Blanchard1), D. Darrow1), E. Fredrick-son1), D. Gates1), L. R. Grisham1), J. C. Hosea1), S. Kaye1), R. Kaita1), S. Kubota9), H. Kugel1),D. Johnson1), B. LeBlanc1), R. Maingi2), R. Maqueda5), E. Mazzucato1), J. Menard1), D. Mueller1),B. A. Nelson4), C. Neumeyer1), F. Paoletti3), S. Paul1), Y.-K. M. Peng2), S. Ramakrish-nan1), R. Raman4), P. Ryan2), S. A. Sabbagh3), C. Skinner1), T. Stevenson1), D. Stutman7),E. Synakowski1), D. Swain2), G. Taylor1), A. Von Halle1), J. Wilgen2), M. Williams1), J. R. Wil-son1), R. Ackers19), R. E. Barry2), A. Bers12), J. Bialek3), P. T. Bonoli12), M. D. Carter2),J. Chrzanowski1), W. Davis1), E. J. Doyle9), L. Dudek1), R. Ellis1), P. Efthimion1), J. R. Fer-ron6), E. Fredd1), M. Finkenthal7), T. Gibney1), R. Goldston1), R. Hatcher1), R. Hawryluck1),H. Hayashiya14), K. Hill1), T. R. Jarboe4), S. C. Jardin1), H. Ji1), M. Kalish1), L. L. Lao6),K. C. Lee11), F. Levinton14), N. C. Luhmann11), P. Lamarche1), B. Mccormack1), R. Ma-jeski1), J. Manickam1), R. Marsala1), T. K. Mau10), S. Medley1), M. M. Menon2), O. Mitarai17),M. Nagata18), N. Nishino16), G. Oliaro1), H. Park1), R. Parsells1), T. Peebles9), G. Pearson1),C. K. Phillips1), R. I. Pinsker6), G. D. Porter13), A. K. Ram12), J. Robinson1), P. Roney1),L. Roquemore1), A. Rosenberg1), M. Schaffer6), S. Shiraiwa15), P. Sichta1), B. Stratton1),D. Stotler1), Y. Takase15), W. R. Wampler8), G. Wurden5), J. G. Yang20), X. Q. Xu13), L. Zeng9),W. Zhu3), S. Zweben1)
1) Princeton Plasma Physics Laboratory, Princeton, NJ, USA2) Oak Ridge National Laboratory, Oak Ridge, TN, USA,3) Columbia University, New York, NY, USA4) University of Washington, Seattle, WA, USA5) Los Alamos National Laboratory, Los Alamos, NM, USA6) General Atomics, San Diego, CA, USA7) Johns Hopkins University, Baltimore, MD, USA8) Sandia National Laboratory, New Mexico, USA9) UC Los Angeles, Los Angeles, CA, USA10) UC San Diego, San Diego, CA, USA11) UC Davis, Davis, CA, USA12) Massachusetts Institute of Technology, Cambridge, MA, USA13) Lawrence Livermore National Laboratory, Livermore, CA, USA14) Fusion Physics, and Technology, San Diego, CA, USA15) Univ. Tokyo, Tokyo, Japan16) Hiroshima Univ., Hiroshima, Japan17) Kyushu Tokai Univ., Kumamoto, Japan18) Himeji Inst. Technology, Okayama, Japan19) EURATOM/UKAEA. Culham, UK20) Korea Basic Science Institute, Taejeon, Korea
Abstract. The main aim of the National Spherical Torus Experiment (NSTX) is to estab-lish the fusion physics principles of the spherical torus (ST) concept. The NSTX device beganplasma operations in February 1999 and the plasma current Ip was successfully brought up tothe design value of 1 million amperes on December 14, 1999. The planned plasma shapingparameters, κ = 1.6 − 2.2 and δ = 0.2 − 0.4, were achieved in inner limited, single null anddouble null configurations. The CHI (Coaxial Helicity Injection) and HHFW (High HarmonicFast Wave) experiments were also initiated. A CHI injected current of 27 kA produced up to260 kA of toroidal current without using an ohmic solenoid. With an injection of 2.3 MW ofHHFW power, using twelve antennas connected to six transmitters, electrons were heated froma central temperature of 400 eV to 900 eV at a central density of 3.5× 1013cm−3 increasing theplasma energy to 59 kJ and the toroidal beta, βT to 10 %. Finally, the NBI system commencedoperation in Sept. 2000. The initial results with two ion sources (PNBI = 2.8MW) shows good
31
heating, producing a total plasma stored energy of 90 kJ corresponding to βT ≈ 18% at a plasmacurrent of 1.1 MA.
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(OV4/3) Operational Boundaries on the Stellarator W7-ASat the Beginning of the Divertor Experiments
R. Jaenicke1), M. Anton1), J. Baldzuhn1), S. Baumel1), R. Brakel1), R. Burhenn1), H. Callaghan1),G. Cattanei1), D. Dorst1), H. Ehmler1), A. Elsner1), M. Endler1), V. Erckmann1), Y. Feng1),S. Fiedler1), C. Fuchs1), F. Gadelmeier1), J. Geiger1), L. Giannone1), P. Grigull1), H. Hacker1),H. J. Hartfuß1), D. Hartmann1), D. Hildebrandt1), M. Hirsch1), E. Holzhauer1), F. Karger1),M. Kick1), J. Kisslinger1), S. Klose1), J. P. Knauer1), J. P. Koponen1), R. Koenig1), G. Kuehner1),H. Laqua1), L. Ledl1), H. Maassberg1), H. Niedermeyer1), K. McCormick1), D. Naujoks1),W. Ott1), F.-P. Penningsfeld1), S. Reimbold1), N. Ruhs1), N. Rust1), J. Saffert1), E. Sallander1),J. Sallander1), F. Sardei1), M. Schubert1), E. Speth1), U. Stroth1), F. Volpe1), F. Wagner1),A. Weller1), C. Wendland1), A. Werner1), E. Wursching1)
Abstract. During the last shutdown the Stellarator W7-AS underwent two major modifica-tions: First, the limiters were replaced by ten divertor modules, and the diagnostic set associatedwith the plasma boundary and target plate regions was greatly expanded. Secondly, the pre-viously counter tangential neutral beam injector box was shifted to a co-position. Thus, theheating efficiency should be considerably increased at low magnetic fields and high densities.After resuming experiments these improvements will be used to test the boundary island di-vertor concept and further expand operational boundaries during the remaining experimentaltime until permanent shutdown in 2002. The present operational boundaries are reviewed withrespect to the stability of high β and density limit discharges. Discharges with good confinementproperties will be discussed where further progress was achieved after installing control coils tomodify the size and properties of vacuum field islands. In contrast to the usual net-current freemode, W7-AS also allows operation at large toroidal currents. In this way disruption-like eventsin the presence of rather large external poloidal fields can be produced.
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(OV4/4) Review of Confinement and Transport Studies inthe TJ-II Flexible Heliac
C. Alejaldre, J. Alonso, L. Almoguera, E. Ascasıbar, A. Baciero, R. Balbın, M. Blaumoser,J. Botija, B. Branas, E. de la Cal, A. Cappa, R. Carrasco, F. Castejon, J. R. Cepero, C. Cremy,J. Doncel, S. Eguilior, T. Estrada, A. Fernandez, C. Fuentes, A. Garcıa, I. Garcıa-Cortes,J. Guasp, J. Herranz, C. Hidalgo, J. A. Jimenez, I. Kirpitchev, V. Krivenski, I. Labrador, F. La-payese, K. Likin, M. Liniers, A. Lopez-Fraguas, A. Lopez-Sanchez, E. de la Luna, R. Martın,L. Martınez-Laso, M. Medrano, P. Mendez, K. J. McCarthy, F. Medina, B. van Milligen,M. Ochando, L. Pacios, I. Pastor, M. A. Pedrosa, A. de la Pena, A. Portas, J. Qin, L. Rodrıguez-Rodrigo, A. Salas, E. Sanchez, J. Sanchez, F. Tabares, D. Tafalla, V. Tribaldos, J. Vega,B. Zurro
1) Laboratorio Nacional de Fusion, Asociacion EURATOM-CIEMAT, Madrid, Spain
Abstract. TJ-II is a four period, low magnetic shear stellarator (R = 1.5m, a < 0.22m, B0 ≤1.2T) which was designed to have a high degree of magnetic configuration flexibility. In the lastexperimental campaign, coupling of the full ECH power (PECRH ≤ 600kW) to the plasma hasbeen possible using two ECRH transmission lines which have different power densities. Bothhelium and hydrogen fuelled plasmas have been investigated. This paper reviews the latestphysics results in particle control, configuration effects, and transport and fluctuation studies.
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(OV4/D) Discussion of Section OV4
The file contains the discussion contributions relating to OV4/1, OV4/2, OV4/3, OV4/4.
H. Weisen1), S. Alberti1), C. Angioni1), K. Appert1), J. Bakos2), R. Behn1), P. Blanchard1),P. Bosshard1), R. Chavan1), S. Coda1), I. Condrea1), A. Degeling1), B. P. Duval1), D. Fasel1), J.-Y. Favez1), A. Favre1), I. Furno1), P. Gomez1), T. Goodman1), M. A. Henderson1), F. Hofmann1),R. R. Kayruthdinov3), P. Lavanchy1), J. B. Lister1), X. Llobet1), A. Loarte4), V. Lukash5),P. Gorgerat1), J.-P. Hogge1), P.-F. Isoz1), B. Joye1), J.-C. Magnin1), A. Manini1), B. Marletaz1),P. Marmillod1), Y. R. Martin1), A. Martynov1), J.-M. Mayor1), E. Minardi6), J. Mlynar1), J.-M. Moret1), P. Nikkola1), P. J. Paris1), A. Perez1), Y. Peysson7), Z. A. Pietrzyk1), V. Piffl8),R. A. Pitts1), A. Pochelon1), H. Reimerdes1), J. H. Rommers1), O. Sauter1), E. Scavino1),A. Sushkov4), G. Tonetti1), M. Q. Tran1) and A. Zabolotsky1)
1) Centre de Recherches en Physique des Plasmas, Ecole Polytechnique Federale de Lausanne,Association EURATOM-Suisse, 1015 Lausanne, Switzerland2) KFKI, Budapest, Hungary3) EFDA-CSU, Garching, Germany4) RRC, Moscow, Russia5) TRINITI, Troisk, Russia,6) Istituto di Fisica del Plasma “P.Caldirola”, Milano, Italy7) CEA, Cadarache, France8) IPP, Prague, Czech Republic
Abstract. The TCV tokamak (R = 0.88m, a < 0.25m, BT < 1.54T) is equipped with six0.5MW gyrotron sources operating at 82.7 GHz for second harmonic X-mode ECH. By dis-tributing the ECCD current sources over the discharge cross section, fully driven stationaryplasmas with Ip = 210kA, ne0 = 2× 1019m−3, Te0 ≈ 4keV, were obtained for the full dischargeduration of 2s. Highly peaked electron temperature profiles with Te0 up to 12keV were obtainedin central counter current drive scenarios with off-axis ECH. Absorption measurements using a118 GHz gyrotron have demonstrated the importance of suprathermal electrons for third har-monic absorption. A coupled heat-particle transport phenomenon known as “density pumpout”,which leads to the expulsion of particles from the plasma core, has been linked to the presenceof m=1 modes, suggesting that it is due to the existence of locally trapped particles associatedwith the loss of axisymmetry. Highly elongated discharges have been developed with Ohmicheating (κ < 2.8) and off-axis ECH. The latter exhibit considerably improved vertical stabilitydue to current profile broadening. A “gateway” for Elmy H-modes has been discovered, whichallows stationary Ohmic ELMy H-mode operation in over wide range of elongation, triangularityand density. Divertor detachment experiments suggest the existence of recombination pathwaysother than three-body or radiative processes.
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(OV5/2) Recent Experiments with ECRH/ECCD in T-10
D. A. Kislov1), the T-10 Team1)
1) RRC “Kurchatov Institute”, Moscow, Russia
Abstract. Experiments with ECRH performed during 1998-2000 are presented. An operatingregime with an edge transport barrier, interpreted as the H-mode, has been studied with bothon-axis and off-axis ECRH. Experimental data imply formation of an internal transport barriersimultaneously with the external transport barrier in the case of off-axis ECRH. Strong evolu-tion of plasma potential in the regions of the external and internal transport barriers has beenobserved using Heavy Ion Beam Probe diagnostics. The mode of improved confinement withpeaked density profiles and increased ion temperature has been observed after pellet injectioninto ECRH heated plasmas. Experimental study of q(r) profile control by ECCD has been per-formed. Instabilities, identified as the neoclassical tearing modes, have been found to limit betain the regimes with a high fraction of bootstrap current. The dependence of the critical β onthe q(r) profile has been observed. A systematic study of plasma turbulence has been startedusing Correlation Reflectometry diagnostics.
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(OV5/3) Steady-state Experiments on High Performance,Current Profile Control and Long Sustainment of LHCDPlasmas on the Superconducting Tokamak TRIAM-1M
H. Zushi1), S. Itoh1), K. Nakamura1), M. Sakamoto1), K. Hanada1), E. Jotaki1), Y. D. Pan2),M. Hasegawa1), S. Kawasaki1), H. Nakashima1)
1) Advanced Fusion Research Center Research Institute for Applied Mechanics, Kyushu Uni-versity, JAPAN2) Southwestern Institute of Physics, China
Abstract. TRIAM-1M (R0 = 0.8m, a×b = 0.12m× 0.18m and B = 8T) has the main missionto study the route toward a high field compact steady state fusion reactor. We have advancedsteady state operation (SSO) programme in tokamaks, studied a heating mechanism for thehigh ion temperature (HIT) mode with an internal transport barrier, obtained an enhancedcurrent drive (ECD) mode in an extended (higher power and higher density) operation regime,performed current density profile control experiments using multi-current drive sys-tems and in-vestigated effects of wall recycling, wall pumping and wall saturation on particle control. In HITmode a hysteresis relation between Ti and ne is found to be ascribed to different timescales forTi and ne to change. Excitation of plasma waves corresponding to ion heating is studied by bothmeasurements of electromagnetic and electrostatic waves and their analysis. Achieved plasmaparameters in ECD are as follows; ne is 4.3 × 1019m−3 , ILHCD is ∼ 70kA, Te and Ti are 0.8keV and 0.5 keV, respectively, and the stored energy is 1.9 kJ. The energy confinement time τEof 8–10 ms, HITER89−P ∼ 1.4, is achieved and the current drive efficiency ηCD = neICDR0/PLH
reaches ∼ 1× 1019Am−2/W at B = 7 T under the fully non-inductive condition. Power thresh-old and hysteresis nature are studied. Bi-directional current drive and superposed current driveexperiments have been carried out. In the former steady current reduction and peaking of j(r)are observed, but it is noticed that self-organization of j(r) occurs above a certain power ratio.In the latter broadening of j(r) can be obtained by increasing superposed RF power, however,self-organization of j(r) also occurs again at a certain power. Temporal behaviour of the recyclingcoefficient with two different time constants (∼ 3 s and ∼ 30 s) is analysed. The wall pumpingrates are evaluated to be ∼ 1.5 × 1016 atoms/s/m2 for low ne and ∼ 4 × 1017 atoms/s/m2 forhigh ne, respectively. In the high power and high density experiments, the wall saturation phe-nomenon affects the density control.
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(OV5/4) Overview on Chinese Tokamak Experimental Progress
J. K. Xie1), J. Li1), HT-7 Team1), Y. Liu2), HL-1M Team2), Y. Z. Wen3), KT-5C Team3),L. Wang4), CT-6B Team4)
1) Institute of Plasma Physics, Chinese Academy of Sciences, P.O.Box 1126 Hefei, China, 2300312) SouthWestern Institue of Physics, ChengDu, China3) University of Sciences and Technology of China, Hefei, China4) Institute of Physics, Chinese Academy of Sciences, Beijing, China
Abstract. Tokamak experiment research in China has made important progress. The mainefforts subjected to quasi-steady state operation, LHCD, plasma heating with ICRF, IBW, NBI,ECRH, fueling with pellet and supersonic molecular beam, first wall conditioning technique.Plasma parameters in experiments were much improved, such as ne = 8×1019m−3, plasma pulse>10Sec. ICRF boronization and conditioning made Zeff close to unit. Steady state full LH wavecurrent drive has been achieved for more than 3 seconds. LHCD ramp up and recharge have alsobeen demonstrated. The Best ηexp
CD ≈ 0.5(1 + 0.085 exp(4.8(BT − 1.45))neICDRp/PLH = 1019m−2A/W.Quasi steady state H-mode like plasma with density close to Greenwald limit was obtained byLHCD, in which energy confinement time was nearly 5 times longer than the Ohmic case. Thesynergy between IBW, pellet and LHCD was tested. Research on the mechanism of macro-turbulence has been extensively carried out experimentally. Ac operation of tokamak was suc-cessfully demonstrated.
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(EX1/1) Progress Towards Confinement Improvement Us-ing Current Profile Modification in the MST Reversed FieldPinch
C. B. Forest1), J. K. Anderson1), T. M. Biewer1), D. Brower2), B. E. Chapman1), P. K. Chat-topadhyay1), D. Craig1), N. C. Crocker1), D. J. Den Hartog1), G. Fiksel1), R. W. Harvey3),Y. Jiang2), N. E. Lanier1), R. O’Connell1), S. C. Prager1), E. Uchimoto4), J. S. Sarff1), A. P. Smirnov5),M. A. Thomas1)
1) MST Group, Department of Physics, University of Wisconsin, Madison WI USA2) University of California, Los Angeles, CA USA3) CompX, Del Mar, CA USA4) Department of Physics and Astronomy, University of Montana, Missoula, MT USA5) Moscow State University, Moscow, Russia
Abstract. Recent current profile modification experiments on the MST reversed field pinchhave resulted in improved performance and elucidated the role of the current profile in de-termining confinement. During transient experiments in which the current profile is modifiedinductively, new profile measurements show that both the electron thermal diffusivity and parti-cle diffusivity can decrease by more than an order of magnitude compared to standard plasmas.Concurrent with this improvement in energy confinement, density fluctuations associated withcore-resonant tearing modes are reduced by more than an order of magnitude over the entireplasma cross-section. Edge resonant modes (poloidal mode number m=0) are shown to affectconfinement and are controllable by current drive in the extreme plasma edge-indicating thesignificance of edge modes in addition to the core-resonant m=1 modes. Finally, experimentsare underway to demonstrate new non-inductive current profile control techniques using lowerhybrid waves and electron Bernstein waves.
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41
(TH/1) Generation and Sustainment of Plasma Rotationby ICRF Heating
F. W. Perkins1), R. White1)3), P. T. Bonoli2), V. S. Chan3)
1) Princeton Plasma Physics Laboratory, P.O. Box 451, Princeton, New Jersey USA2) Plasma Science and Fusion Center, MIT, Cambridge, Massachusetts3) General Atomics, P.O. Box 451, San Diego, California USA
Abstract. A mechanism is proposed and evaluated for driving rotation in tokamak plas-mas by minority ion-cyclotron heating, even though this process introduces negligible angularmomentum. The mechanism has two elements: First, angular momentum transport is gov-erned by a diffusion equation with a non-slip boundary condition at the separatrix. Second,Monte-Carlo calculations show that energized particles will provide a torque density sourcewhich has a zero volume integral but separated positive and negative regions. With such asource, a solution of the diffusion equation predicts the on-axis rotation frequency Ω to beΩ = (4qmaxWJ∗)eBR3a2ne(2π)2)−1(τM/τE) where |J∗| ≈ 5 − 10 is a nondimensional rotationfrequency calculated by the Monte-Carlo ORBIT code. Overall, agreement with experiment isgood, when the resonance is on the low-field-side of the magnetic axis. The rotation becomesmore counter-current and reverses sign on the high field side for a no-slip boundary. The velocityshear layer position is controllable and of sufficient magnitude to affect microinstabilities.
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42
(OV5/D) Discussion of Section OV5/EX1/TH1
The file contains the discussion contributions relating to OV5/1, OV5/2, OV5/3, OV5/4. EX1/1,TH1/1.
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43
Session OV6/ITER — ITER
Contents
(OV6/1) Key Engineering Features of the ITER-FEAT Magnet System andImplications for the R&D Programme . . . . . . . . . . . . . . . . . . . . 45
(OV6/1) Key Engineering Features of the ITER-FEAT Mag-net System and Implications for the R&D Programme
M. Huguet1) and the ITER Joint Central Team, Home Teams2)3)4)
1) ITER Naka Joint Work Site, Naka-machi, Japan2) EFDA Closing Unit, Garching, Germany3) JAERI Naka, Naka-machi, Japan4) Efremov Institute, St. Petersburg, Russian Federation
Abstract. The magnet design of the new ITER-FEAT machine comprises 18 Toroidal Field(TF) coils, a Central Solenoid (CS), 6 Poloidal Field (PF) coils and Correction Coils (CCs). Akey driver of this new design is the requirement to generate and control plasmas with a relativelyhigh elongation (k95 = 1.7) and a relatively high triangularity (δ95 = 0.35). This has lead to adesign where the CS is vertically segmented and self-standing and the TF coils are wedged alongtheir inboard legs. Another important design driver is to achieve a high operational reliabilityof the magnets, and this has resulted in several unconventional designs, and in particular, theuse of conductors supported in radial plates for the winding pack of the TF coils. A key me-chanical issue is the cyclic loading of the TF coil cases due to the out-of-plane loads which resultfrom the interaction of the TF coil current and the poloidal field. These loads are resisted bya combination of shear keys and “pre-compression” rings able to provide a centripetal preloadat assembly. The fatigue life of the CS conductor jacket is another issue as it determines theCS performance in terms of the flux generation. Two jacket materials and designs are understudy. Since 1993, the ITER magnet R&D programme has been focussed on the manufactureand testing of a CS and a TF model coil. During its testing, the CS model coil has successfullyachieved all its performance targets in DC and AC operations. The manufacture of the TFmodel coil is complete. The manufacture of segments of the full scale TF coil case is anotherimportant and successful part of this programme and is near completion. New R&D effort isnow being initiated to cover specific aspects of the ITER-FEAT design.
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45
(ITER/1) ITER-FEAT Operation
Y. Shimomura1), R. Aymar2), V. Chuyanov2), M. Huget1), H. Matsumoto2), T. Mizoguchi1),Y. Murakami1), A. Polevoi1), M. Shimada1) and the ITER Joint Central Team and Home Teams
1) ITER Joint Central Team, Naka Joint Work Site, Naka-machi, Ibaraki-ken, Japan2) ITER Joint Central Team, Garching Joint Work Site, Garching, Germany
Abstract. ITER is planned to be the first fusion experimental reactor in the world operatingfor research in physics and engineering. The first 10 years’ operation will be devoted primarily tophysics issues at low neutron fluence and the following 10 years’ operation to engineering testingat higher fluence. ITER can accommodate various plasma configurations and plasma operationmodes such as inductive high Q modes, long pulse hybrid modes, non-inductive steady-statemodes, with large ranges of plasma current, density, beta and fusion power, and with variousheating and current drive methods. This flexibility will provide an advantage for coping withuncertainties in the physics database, in studying burning plasmas, in introducing advanced fea-tures and in optimizing the plasma performance for the different programme objectives. Remotesites will be able to participate in the ITER experiment. This concept will provide an advantagenot only in operating ITER for 24 hours per day but also in involving the world-wide fusioncommunities and in promoting scientific competition among the Parties.
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46
(ITER/3) Divertor Design and its Integration into the ITER-FEAT Machine
G. Janeschitz1), A. Antipenkov1), G. Federici1), C. Ibbott1), A. S. Kukushkin1), P. Ladd1),E. Martin1), R. Tivey1)
1) ITER Joint Central Team, Boltzmannstrasse 2, 85748 Garching, Germany
Abstract. The physics of the edge and divertor plasma is strongly coupled with the divertor andthe fuel cycle design. Due to the limited space available the design as well as the remote main-tenance approach for the ITER divertor are highly optimized to allow maximum space for thedivertor plasma. Several auxiliary systems (e.g. in vessel viewing, glow discharge electrodes. . . )as well as a part of the pumping and fuelling system have to be integrated together with thedivertor into the lower level of the ITER machine. Two main options exist for the choice of theplasma-facing material in the divertor, i.e. W and CFC. Based on already existing R&D resultsone can be optimistic that the material choice will be mainly based on physics considerationsand material issues (e.g. C-T co-deposition). The requirements for the ITER fuel cycle arisefrom plasma physics as well as from the envisaged operation scenarios. Due to the complexdynamic relationship of the fuel cycle subsystems among themselves and with the plasma, codesare employed for their optimization. This paper elaborates these interacting issues and gives thelatest design status.
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47
(ITER/4) Progress of the ITER Central Solenoid ModelCoil Program
H. Tsuji1), K. Okuno2), R. Thome3), E. Salpietro4), S. Egorov5), N. Martovetsky6), M. Ricci7),R. Zanino8), G. Zahn9), A. Martinez10), G. Vecsey11), K. Arai12), T. Ishigooka13), T. Kato1),T. Ando1), Y. Takahashi1), H. Nakajima1), T. Hiyama1), M. Sugimoto1), N. Hosogane1), M. Mat-sukawa1), Y. Miura1), T. Terakado1), J. Okano1), K. Shimada1), M. Yamashita1), T. Isono1),N. Koizumi1), K. Kawano1), M. Oshikiri1), Y. Nunoya1), K. Matsui1), Y. Tsuchiya1), G. Nishi-jima1), H. Kubo1), T. Shimba1), E. Hara1), K. Imahashi1), Y. Uno1), T. Ohuchi1), K. Ohtsu1),J. Okayama1), T. Kawasaki1), M. Kawabe1), S. Seki1), K. Takano1), Y. Takaya1), F. Tajiri1),F. Tsutsumi1), T. Nakamura1), H. Hanawa1), H. Wakabayashi1), T. Shimizu1), K. Kuramochi1),T. Omine1), T. Tamiya1), J. Harada1), K. Nishii1), M. Huguet2), N. Mitchell2), D. Bessette2),J. Minervini3), R. Vieira3), P. Michael3), M. Takayasu3), G. Bevilacqua4), R. Maix4), R. Mana-han6), R. Jayakumar6), L. Savoldi8), W. Herz9), A. Ninomiya13)
1) Japan Atomic Energy Research Institute (JAERI), Naka, Japan2) ITER Joint Central Team, Naka, Japan3) Massachusetts Institute of Technology, Cambridge, MA, USA4) European Fusion Development Agreement, Garching, Germany5) D. V. Efremov Scientific Research Institute, St. Petersburg, Russia6) Lawrence Livermore National Laboratory, Livermore, USA7) ENEA C. R. Frascati, Frascati, Italy8) Politecnico di Torino, Torino, Italy9) Forschungszentrum Karlsruhe , Eggenstein-Leopoldshafen, Germany10) CEA Cadarache, St-Paul-Lez-Durance, France11) Ecole Polytechnique Federale de Lausanne, Villigen, Suisse12) Electrotechnical Laboratory, Tsukuba, Japan13) Seikei University, Musashino, Japan
Abstract. The world s largest pulsed superconducting coil was successfully tested by chargingup to 13 T and 46 kA with a stored energy of 640 MJ. The ITER Central Solenoid (CS) ModelCoil and CS Insert Coil were developed and fabricated through an international collaborationand their cool down and charging tests were successfully carried out by international test andoperation teams. In pulsed charging tests, where the original goal was 0.4T/s up to 13T, theCS Model Coil and the CS Insert Coil achieved ramp rates of 0.6T/s and 1.2T/s up to 13T,respectively. In addition, the CS Insert Coil was charged and discharged 10,003 times in the13-T background field of the CS Model Coil and no degradation of the operational temperaturemargin directly coming from this cyclic operation was observed. These test results fulfilled allthe goals of CS Model Coil development by confirming the validity of the engineering design anddemonstrating that we are now ready to construct the ITER coils with confidence.
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48
(ITER/5) ITER-FEAT Vacuum Vessel and Blanket DesignFeatures and Implications for the R&D Programme
K. Ioki1), W. Daenner2), K. Koizumi3), V. Krylov4), A. Cardella1), F. Elio1), M. Onozuka1),ITER Joint Central Team and Home Teams
1) ITER Joint Central Team, Garching Joint Work Site, Garching, Germany2) EU Home Team, EFDA CSU, Max-Planck Institute, Garching, Germany3) JA Home Team, Japan Atomic Energy Research Institute (JAERI), Japan4) RF Home Team, Efremov Institute, St. Petersburg, Russian Federation
Abstract. A tight fitting configuration of the VV to the plasma aids the passive plasma verticalstability, and ferromagnetic material in the VV reduces the TF ripple. The blanket modulesare supported directly by the VV. A full-scale VV sector model has provided critical informa-tion related to fabrication technology, and the magnitude of welding distortions and achievabletolerances. This R&D validated the fundamental feasibility of the double-wall VV design. Theblanket module configuration consists of a shield body to which a separate first wall is mounted.The separate first wall has a facet geometry consisting of multiple flat panels, where 3-D machin-ing will not be required. A configuration with deep slits minimizes the induced eddy currentsand loads. The feasibility and the robustness of solid HIP joining was demonstrated in R&D, bymanufacturing and testing several small and medium scale mock-ups and finally two prototypes.Remote handling tests and assembly tests of a blanket module have demonstrated the basicfeasibility of its installation and removal.
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49
(ITER/6) Status of R&D of the Plasma Facing Componentsfor the ITER Divertor
I. V. Mazul1), M. Akiba2), I. Arkhipov3), V. Barabash4), S. Chiocchio4), K. Ezato2), G. Fed-erici4), G. Janeschitz4), C. Ibbott4), A. Makhankov1), A. Markin3), M. Merola5), S. Suzuki2),R. Tivey4), M. Ulrickson6), G. Vieider5), C. H. Wu5)
1) Efremov Institute of Electrophysical Apparatus, 189631, St. Petersburg, Russia2) Japan Atomic Energy Research Institute, Naka-machi, Naka-gun, Ibaraki-ken, Japan3) Institute of Physical Chemistry, Moscow, Russia4) ITER Joint Central Team, Garching, Germany5) EFDA Close Support Unit, Garching, Germany6) Sandia National Laboratories, Albuquerque, USA
Abstract. The paper reports the progress made by the ITER Home Teams in the develop-ment of robust carbon and tungsten armoured plasma facing components for the ITER diver-tor. The activities on the development and study of armour materials, joining technologies,non-destructive evaluation techniques, high heat flux testing of manufactured components andneutron irradiation resistance studies are presented. The results of these activities confirm thefeasibility of the main divertor components. Examples of the fruitful collaboration between Par-ties and future R&D needs are also described.
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50
(OV6/D) Discussion of Section OV6/ITER
The file contains the discussion contributions relating to OV6/1, ITER/1, ITER/2(R), ITER/4,ITER/5.
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51
Session OV7/FT — FusionTechnology
Contents
(OV7/1) Design and Construction of the KSTAR Tokamak . . . . . . . . . . 53
(OV7/1) Design and Construction of the KSTAR Tokamak
G. S. Lee1) and the KSTAR Team
1) Korea Basic Science Institute, Korea
Abstract. The extensive design effort has been focused on two major aspects of the KSTARproject mission, steady-state operation capability and “advanced tokamak” physics. The steady-state aspect of mission is reflected in the choice of superconducting magnets, provision of activelycooled in-vessel components, and long-pulse current-drive and heating systems. The “advancedtokamak” aspect of the mission is incorporated in the design features associated with flexibleplasma shaping, double-null divertor and passive stabilizers, internal control coils, and a compre-hensive set of diagnostics. Substantial progress in engineering has been made on superconductingmagnets, vacuum vessel, plasma facing components, and power supplies. The new KSTAR ex-perimental facility with cryogenic system and de-ionized water-cooling and main power systemshas been designed, and the construction work has been on-going for completion in year 2004.
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53
(FT/4) The Helias Reactor
J. Kisslinger1), C. D. Beidler1), E. Harmeyer1), F. Herrnegger1), Yu. Igitkhanov1), A. Kendl1),Ya. I. Kolesnichenko2), V. V. Lutsenko2), C. Nuhrenberg1), I. Sidorenko1), E. Strumberger1),H. Wobig1), Yu. V. Yakovenko2)
1) Max-Planck Institut fur Plasmaphysik, EURATOM Association D-85740 Garching bei Munchen,Germany.2) Scientific Centre “Institute for Nuclear Reseach” 03680 Kyiv, Ukraine
Abstract. The Helias reactor is an upgraded version of the Wendelstein 7-X experiment. Astraightforward extrapolation of Wendelstein 7-X leads to HSR5/22, which has 5 field periodsand a major radius of 22 m. HSR4/18 is a more compact Helias reactor with 4 field periodsand 18 m major radius. Stability limit and energy confinement times are nearly the same asin HSR5/22, thus the same fusion power (3000 MW) is expected in both configurations. Neo-classical transport in HSR4/18 is very low, the effective helical ripple is below 1%. The paperdescribes the power balance of the Helias reactor, the blanket and maintenance concept. Thecoil system of HSR4/18 comprises 40 modular coils with NbTi-superconducting cables. Thereduction from 5 to 4 field periods and the concomitant reduction in size will also reduce thecost of the Helias reactor.
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54
(OV7/D) Discussion of Section OV7/FT
The file contains the discussion contributions relating to OV7/1, FT/1(R), FT/4.
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55
Session EX2 — Transport 1
Contents
(EX2/1) Pellet Fuelling and ELMy H-mode Physics at JET . . . . . . . . . . 57
Abstract. As the reference operating regime for ITER, investigations of the ELMy H-modehave received high priority in the JET experimental programme. Recent experiments have con-centrated in particular on operation simultaneously at high density and high confinement usinghigh field side (HFS) pellet launch. The enhanced fuelling efficiency of HFS pellet fuelling isfound to scale favourably to a large machine such as JET. The achievable density of ELMyH-mode plasmas in JET has been significantly increased using HFS fuelling although at theexpense of confinement degradation back to L-mode levels. Initial experiments using controlof the pellet injection frequency have shown that density and confinement can simultaneouslybe increased close to the values necessary for ITER. The boundaries of the available ELMyH-mode operational space have also been extensively explored. The power necessary to main-tain the high confinement normally associated with ELMy H-mode operation is found to besubstantially higher than the H-mode threshold power. The compatibility of ELMy H-modeswith divertor operation acceptable for a fusion device has been studied. Narrow energy scrape-off widths are measured which place stringent limits on divertor power handling. Deuteriumand tritium codeposition profiles are measured to be strongly in/out asymmetric. Successfulmodelling of these profiles requires the inclusion of the (measured) scrape-off layer flows and ofthe production in the divertor of hydrocarbon molecules with sticking coefficients below unity.Helium exhaust and compression are found to be within the limits sufficient for a reactor.
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57
(EX2/2) Confinement and Transport Studies of Conven-tional Scenarios in ASDEX Upgrade
F. Ryter1), J. Stober1), A. Stabler1), G. Tardini1), H.-U. Fahrbach1), O. Gruber1), A. Her-rmann1), F. Imbeaux2), A. Kallenbach1), M. Kaufmann1), B. Kurzan1), F. Leuterer1), M. Maraschek1),H. Meister1), A. G. Peeters1), G. V. Pereverzev1), A. Sips1), W. Suttrop1), W. Treutterer1),H. Zohm1)
1) Max-Planck-Institut fur Plasmaphysik, Garching, Germany2) CEA-Cadarache, DRFC, France
Abstract. Confinement studies of conventional scenarios, i.e. L and H modes, in ASDEX Up-grade indicate that the ion and electron temperature profiles are generally limited by a criticalvalue of ∇T/T . When this is the case the profiles are stiff: core temperatures are proportional topedestal temperatures. Transport simulations based on turbulence driven by Ion TemperatureGradient show good agreement with the ion experimental data. Studies specifically dedicated toelectron transport using Electron Cyclotron Heating in steady-state and modulated indicate thatelectron temperature profiles are also stiff, in agreement with recent calculations on transportdriven by ETG turbulence with streamers. In particular the predicted threshold and the increaseof the stiffness factor with temperature are found experimentally. The density profiles are notstiff. As a consequence of this profile behaviour, the plasma energy is proportional to pedestalpressure and improves with density peaking. The confinement time increases with triangularityand can be good at densities close to the Greenwald limit. In this operational corner and atq95 ≈ 4, the replacement of type-I ELMs by small ELMs of type-II provides good confinementwith much reduced peak power load on the divertor plates.
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58
(EX2/3) ITER Shaping and Elongation Experiments onJET
Abstract. The flexibility of the design of JET has made it well suited to exploring the effects ofdifferent plasma shapes and elongations on ELMy H-mode plasmas. Over the past few years, anumber of experiments [G. Sabeine; Nuclear Fusion 39 (1999) 1133] have measured such effectsand attempted to explain the underlying physics. Two of the principal results have been thestrong dependence of confinement on elongation and the improvement of confinement, for densi-ties close to the Greenwald density, at higher triangularities. This led to experiments in Autumn1999, at the request of the ITER project, to measure, independently, the effects of elongationand shaping at constant q and current. Here we draw together these experiments to producethe best measure yet of the elongation and triangularity scaling on JET. By including datafrom other machines, from the international confinement database [O. Kardaun; Plasma Phys.Control. Fusion 41 (1999) 429], we show this scaling’s impact on possible next step machines,especially by means of two term scaling laws.
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59
(EX2/4) Effect of Variation in Equilibrium Shape on ELM-ing H-mode Performance in DIII-D Diverted Plasmas
M. E. Fenstermacher1), T. H. Osborne2), T. W. Petrie2), C. J. Lasnier1), A. W. Leonard2),J. G. Watkins3), T. N. Carlstrom2), R. J. Groebner2), A. W. Hyatt2), R. J. La Haye2), M. A. Mah-davi2), G. D. Porter1), S. L. Allen1), J. A. Boedo4), N. H. Brooks2), R. J. Colchin5), R. Maingi5),M. E. Rensink1), T. L. Rhodes6), D. M. Thomas2), M. R. Wade5), W. P. West2), D. G. Whyte4),N. S. Wolf1), the DIII-D Team2)
1) Lawrence Livermore National Laboratory, Livermore, California2) General Atomics, San Diego, California3) Sandia National Laboratories, Albuquerque, New Mexico4) University of California-San Diego, La Jolla, California5) Oak Ridge National Laboratory, Oak Ridge, Tennessee6) University of California-Los Angeles, Los Angeles, California
Abstract. The changes in the performance of the core, pedestal, scrape-off-layer (SOL), anddivertor plasmas as a result of changes in triangularity, δ, up/down magnetic balance, and sec-ondary divertor volume were examined in shape variation experiments using ELMing H modeplasmas on DIII-D. In moderate density, unpumped plasmas, high δ ∼ 0.7 increased the energyin the H mode pedestal and the global energy confinement of the core, primarily due to anincrease in the margin by which the edge pressure gradient exceeded the value which would havebeen expected had it been limited by infinite-n ideal ballooning modes. In addition, a nearlybalanced double-null (DN) shape was effective for sharing the peak heat flux in the divertor inthese attached plasmas. For detached plasmas good heat flux sharing was obtained for a sub-stantial range of unbalanced DN shapes. Finally, the presence of a second X-point in unbalancedDN shapes did not degrade the plasma performance if it was sufficiently far inside the vacuumvessel. These results indicate that a high δ unbalanced DN shape has some advantages over asingle null shape for future high power tokamak operation.
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60
(EX2/5) Enhanced D-Alpha H-mode Studies in the AlcatorC-Mod Tokamak
E. S. Marmar1), R. L. Boivin1), C. Fiore1), J. A. Goetz1), R. S. Granetz1), M. J. Green-wald1), A. Hubbard1), J. Hughes1), I. H. Hutchinson1), J. Irby1), B. LaBombard1), Y. Lin1),B. Lipschultz1), A. Mazurenko1), D. Mossessian1), T. Sunn Pedersen1), M. Porkolab1), J. Rice1),G. Schilling2), J. A. Snipes1), G. Taylor2), J. L. Terry1), S. Wolfe1), S. Wukitch1)
1) MIT Plasma Science and Fusion Center, Cambridge, MA USA2) Princeton Plasma Physics Laboratory, Princeton, NJ USA
Abstract. A favorable regime of H-mode confinement, seen on the Alcator C-Mod tokamak isdescribed. Following a brief period of ELM-free H-mode, the plasma evolves into the EnhancedD-Alpha (EDA) H-mode which is characterized by very good energy confinement, the completeabsence of large, intermittent type I ELMs, finite impurity and majority species confinement,and low radiated power fraction. Accompanying the EDA H-mode, a quasi-coherent (QC) edgemode is observed, and found to be responsible for particle transport through the edge confine-ment barrier. The QC-mode is localized within the strong density gradient region, and haspoloidal wavenumber kθ ' 5cm−1 and lab-frame frequency of ' 100 kHz. Parametric studiesshow that the conditions which promote EDA include moderate safety factor (q95 > 3.5), hightriangularity (δ > 0.35) and high target density (ne > 1.2× 20m−3). EDA H-mode is readilyobtained in purely ohmic and well as in ICRF auxiliary-heated discharges.
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61
(EX2/6) Core and Edge Confinement Studies with DifferentHeating Methods in JET
F. G. Rimini1), G. Saibene2), the JET Team3)
1) Ass. Euratom-CEA -Departement de Recherches sur la Fusion Controlee, CEN-CadaracheSt.Paul-lez-Durance, F13108, France2) EFDA Close Support Unit, Garching, Germany3) see appendix to IAEA-CN-77/OV1/2, The JET Team (presented by C. Gormezano)
Abstract. A detailed comparison of core and edge confinement with different heating methods, NBI and ICRH, has been carried out in the ELMy H-mode regime in JET with the Gas Boxdivertor. Transport in the core and characteristics of the edge pedestal have been assessed indischarges at 2.0 MA / 2.6 T at total input power level of 11-12 MW. The thermal core con-finement has been found to be higher by about 10% in ICRH dominated discharge. Althoughthis difference is well within the uncertainties of the thermal confinement estimation, it hasbeen consistently found in similar experiments in the past and it may be related to the morepeaked power deposition provided by ICRF heating. Local transport analysis carried out withthe TRANSP code indicates that, independently of the NBI vs. ICRH mix, ion conduction lossesare the dominant energy loss channel. Unlike previous experiments, the gas flow and density ofNBI and ICRH discharges have been closely matched. In these conditions it has been found thatboth types of heating yield similar values of edge density and temperature and produce similarELMs. The fact that the edge fast ion concentration can be varied from roughly 0.4% up to 4%without producing significant changes in the edge pedestal parameters is an indication that fastions do not always play a dominant role in the edge stabilization between ELMs, especially indischarges with strong gas fuelling.
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62
(EX2/7) Relation Among Potential, Fluctuation Changeand L/H Transition in the JFT-2M Tokamak
Y. Miura1), T. Ido2), K. Kamiya1), Y. Hamada2) and JFT-2M Group1)
1) Japan Atomic Energy Research Institute, Ibaraki, 311-0193, Japan2) National Institute for Fusion Science, Gifu, 509-5292 Japan
Abstract. Potential and density/temperature fluctuation at L-H transition are measured by theheavy ion beam probe (HIBP) on JFT-2M. It has been observed that the time scale of potentialchange is as fast as 10-100µs when the input power (Pin) is larger than the L-H threshold power(Pth). When Pin ∼ Pth , the confinement is improved gradually step by step with sawteethcrashes accompanied with the decrease of potential. After a few sawteeth crashes, potentialdrops rapidly to the level of the ELM-free H-mode. From the gradual change of the potentialwith assuming that dEr/dr is a key to form and to sustain the transport barrier, the criterionof the dEr/dr is less than (1.2± 0.4)× 103kV/m2. At an ELM just before H-L transition, thepotential inside the separatrix also shows the rapid positive jump. The time scale of the positivejump and its recover to its negative value is about 40µsec and 150µsec, respectively. Before theH-L transition, the time between ELMs and/or dithering transition becomes shorter and theplasma finally goes back to L-mode.
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63
(EX2/D) Discussion of Section EX2
The file contains the discussion contributions relating to EX2/1, EX2/2, EX2/3, EX2/4, EX2/5,EX2/6, EX2/7.
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64
Session EX3 — Stability 1
Contents
(EX3/1) The Physics of Neoclassical Tearing Modes and their Stabilisationby ECCD in ASDEX Upgrade . . . . . . . . . . . . . . . . . . . . . . . . 66
(EX3/2) High Performance and Stability in COMPASS-D . . . . . . . . . . . 67
Abstract. The physics understanding of Neoclassical Tearing Modes (NTMs) gained throughexperiments and modelling at ASDEX Upgrade is presented. The onset βN for NTM is foundto be proportional to the normalised ion gyroradius ρ∗ and independent of the normalised colli-sionality ν∗ for a wide range of ν∗. This scaling is in accordance with both polarisation currentand χ⊥/χ‖ model, if for the latter, the heat flux limit on parallel heat conductivity is taken intoaccount. Analysis of the structure and dynamics of NTMs validates the negative ∆′. Completestabilisation using ECCD has been demonstrated at βN = 2.5 and with about 10 % of the totalheating power. The results are in good qualitative agreement with modelling using the Ruther-ford equation and in quantitative agreement with a 2 dimensional nonlinear cylindrical tearingmode code. A precise positioning of the ECCD microwave beam, so far achieved by feed-forwardvariation of Bt, is required for efficient stabilisation.
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66
(EX3/2) High Performance and Stability in COMPASS-D
B. Lloyd1), L. C. Appel1), K. B. Axon1), C. A. Bunting1), R. J. Buttery1), P. G. Carolan1),N. J. Conway1), G. Cunningham1), J. Dowling1), M. R. Dunstan1), T. Edlington1), A. R. Field1),S. J. Fielding1), S. J. Manhood1), K. G. McClements1), H. Meyer1), A. W. Morris1), M. R. O’Brien1),T. Pinfold1), V. Shevchenko1), K. Stammers1), M. Tournianski1), M. Valovic1), M. Walsh2),C. D. Warrick1), H. R. Wilson1)
1) EURATOM/UKAEA Fusion Association, Culham Science Centre,Abingdon, Oxon, OX14 3DB, United Kingdom2) Walsh Scientific Ltd, Culham Science Centre, Abingdon,Oxon, OX14 3EB, United Kingdom
Abstract. COMPASS-D is a compact, adaptable, D-shaped tokamak equipped with powerfulheating (ECRH) and current drive (LHCD/ECCD) systems allowing access to both H-mode(ELMy and ELM-free) and quasi-stationary high beta regimes under conditions of dominantelectron heating (Te > Ti), negligible external momentum input and no central fuelling. Con-trol and avoidance of neo-classical tearing modes (NTMs) has enabled quasi-stationary high beta(βN ∼ 2, βp > 1) discharges to be sustained for sim 20 energy confinement times and a durationcorresponding to 20% of that of a nominal ITER discharge, when normalised to the currentdiffusion time. Controlled seeding of NTMs by external application of resonant magnetic per-turbations has enabled NTM onset criteria to be carefully explored and compared with theory;observed island evolutions follow theoretical expectations. Off-axis lower hybrid current drive(LHCD) has been reliably used to completely stabilise NTMs in high beta discharges. Detailedmodelling has shown that the stabilising effect is consistent with a reduction in the stability in-dex ∆′, although other stabilisation mechanisms may also contribute. High frequency energeticparticle driven instabilities (∼400kHz), which exhibit frequency-sweeping (‘chirping’), have, forthe first time, been observed with high power ECRH as the sole source of auxiliary heating.
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(EX3/3) Mode Coupling Trigger of Tearing Modes in ECWHeated Discharges in FTU
S. Cirant1), E. Lazzaro1), A. Bruschi1), P. Buratti2), R. Coelho3), M. De Benedetti2), G. Granucci1),A. Jacchia1), S. Nowak1), G. Pucella2), G. Ramponi1), P. Smeulders2), A. Smolyakov4), C. Sozzi1),O. Tudisco2), B. Angelini2), M. L. Apicella2), G. Apruzzese2), E. Barbato2), L. Bertalot2),A. Bertocchi2), G. Bracco2), G. Buceti2), A. Cardinali2), C. Centioli2), R. Cesario2), S. Ciattaglia2),V. Cocilovo2), F. Crisanti2), R. De Angelis2), B. Esposito2), D. Frigione2), L. Gabellieri2), F. Gan-dini1), G. Gatti2), E. Giovannozzi2), C. Gormezano2), M. Grolli2), H. Kroegler2), M. Leigheb2),G. Maddaluno2), M. Marinucci2), G. Mazzitelli2), P. Micozzi2), F. P. Orsitto2), D. Pacella2),L. Panaccione2), M. Panella2), V. Pericoli-Ridolfini2), L. Pieroni2), S. Podda2), G. B. Righetti2),F. Romanelli2), S. E. Segre2), A. Simonetto1), E. Sternini2), N. Tartoni2), A. A. Tuccillo2),V. Vitale2), G. Vlad2), V. Zanza2), M. Zerbini2), F. Zonca2)
1) Associazione EURATOM-ENEA-CNR, Istituto di Fisica del Plasma, CNR, Milano, Italy2) Associazione EURATOM-ENEA sulla Fusione, C.R. Frascati, Frascati, Roma, Italy3) Associacao EURATOM/IST, Centro de Fusao Nuclear, 1096 Lisbon, Portugal4) University of Saskatchewan, Saskatoon, Canada5) INFM and Universita di Roma “Tor Vergata”, Roma Italy
Abstract. In recent ECRH experiments on FTU tokamak the destabilization of (2,1) tearingmodes, coupled to the (1,1) mode, has been observed. The dynamics of the rotating islandcan be described by a Rutherford-type equation, where mode coupling, inertial effects and re-sistive walls effects are included. TM dynamics is influenced by the position of the absorbinglayer with respect to the island, to the extent that the mode can be completely suppressed ifabsorption occurs at the island radial position, with a precision in the order of the island size.EC absorption leading to TM stabilization is localized on the island by mechanical steering ofthe EC beam, and not by adjusting the toroidal field or forcing the plasma column in the rightposition for stabilization. Stabilization is achieved at a minimum ECRH power of ≈ 0.15POH.TM stabilization improves significantly the energy confinement in the plasma core.
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68
(EX3/5) New Insights into MHD Dynamics of MagneticallyConfined Plasmas from Experiments in RFX
P. Martin1), S. Martini1), V. Antoni1), L. Apolloni1), M. Bagatin1), W. Baker1), O. Barana1),R. Bartiromo1), P. Bettini1), A. Boboc1), T. Bolzonella1), A. Buffa1), A. Canton1), S. Cap-pello1), L. Carraro1), R. Cavazzana1), G. Chitarin1), S. Costa1), F. D’Angelo1), S. Dal Bello1),A. De Lorenzi1), D. Desideri1), D. F. Escande2), L. Fattorini1), P. Fiorentin1), P. Franz1),E. Gaio1), L. Garzotti1), L. Giudicotti1), F. Gnesotto1), L. Grando1), S. C. Guo1), P. Inno-cente1), A. Intravaia1), R. Lorenzini1), A. Luchetta1), G. Malesani1), G. Manduchi1), G. Mar-chiori1), L. Marrelli1), E. Martines1), A. Maschio1), A. Masiello1), F. Milani1), M. Moresco1),A. Murari1), P. Nielsen1), M. O’Gorman1), S. Ortolani1), R. Paccagnella1), R. Pasqualotto1),B. Pegourie3), S. Peruzzo1), R. Piovan1), N. Pomaro1), A. Ponno1), G. Preti1), M. E. Puiatti1),G. Rostagni1), F. Sattin1), P. Scarin1), G. Serianni1), P. Sonato1), E. Spada1), G. Spizzo1),M. Spolaore1), C. Taliercio1), G. Telesca1), D. Terranova1), V. Toigo1), L. Tramontin1), M. Val-isa1), N. Vianello1), M. Viterbo1), L. Zabeo1), P. Zaccaria1), P. Zanca1), B. Zaniol1), L. Zanotto1),E. Zilli1), G. Zollino1)
1) Consorzio RFX, Associazione Euratom – ENEA sulla Fusione, Corso Stati Uniti 4, 35127Padova, Italy2) UMR 6633 CNRS-Universite de Provence, Avenue Normandie-Niemen, 13397 Marseille Cedex20, France3) Association EURATOM-CEA sur la Fusion Controlee, C. E. Cadarache, 13108 Saint-Paul-lez-Durance, France
Abstract. The experimental and theoretical activity performed in the RFX experiment hasallowed a deeper insight into the MHD properties of the RFP configuration. A set of successfulexperiments has demonstrated the possibility of influencing both the amplitude and the spec-trum of the magnetic fluctuations which characterise the RFP configuration. A new regime (QSHstates) where the dynamo mechanism works in a nearly laminar way and a helical core plasmais produced has been investigated. With these studies a reduction of the magnetic chaos hasbeen obtained. The continuos rotation of wall locked resistive tearing modes has been obtainedby an m = 0 rotating perturbation. This perturbation induces rotation of m = 1 non-linearlycoupled modes.
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(EX3/6) Equilibrium Properties of Ohmically Heated Plas-mas in NSTX
S. A. Sabbagh1), S. Kaye2), J. Menard2), F. Paoletti1), M. Bell2), R. Bell2), J. Bialek1), M. Bit-ter2), E. Fredrickson2), D. Gates2), A. H. Glasser5), H. Kugel2), L. L. Lao3), B. LeBlanc2),R. Maingi4), R. Maqueda5), E. Mazzucato2), D. Mueller2), M. Ono2), S. Paul2), M. Peng4),C. Skinner2), D. Stutman6), G. Wurden5), W. Zhu1)
1) Columbia University, New York, NY, USA2) Princeton University, Princeton, NJ, USA3) General Atomics, San Diego, CA, USA4) Oak Ridge National Laboratory, Oak Ridge, TN, USA5) Los Alamos National Laboratory, Los Alamos, NM, USA6) Johns Hopkins University, Baltimore, MD, USA
Abstract. Research in the National Spherical Torus Experiment, NSTX, has been conductedto establish spherical torus plasmas to be used for high-β , auxiliary heated experiments. Thedevice has a major radius R0 = 0.86 m, a midplane half-width of 0.7 m, and has been oper-ated with toroidal magnetic field B0 ≤ 0.3 T and Ip ≤ 1.0 MA. The evolution of the plasmaequilibrium is analyzed between shots with an automated version of the EFIT code. Limiter,double-null, and lower single-null diverted configurations have been sustained for several energyconfinement times. Plasma stored energy has reached 92 kJ (βt = 17.8%) with neutral beamheating. Plasma elongation of 1.6 ≤ κ ≤ 2.0 and triangularity in the range 0.25 ≤ δ ≤ 0.45have been sustained, with values of κ = 2.5 and δ = 0.6 being reached transiently. The recon-structed magnetic signals are fit to the corresponding measured values with low error. Aspectsof the plasma boundary, pressure, and safety factor profiles are supported by measurements fromnon-magnetic diagnostics. Plasma densities have reached 0.8 and 1.2 times the Greenwald limitin deuterium and helium plasmas, respectively, with no clear limit encountered. Instabilitiesincluding sawteeth and reconnection events (REs), characterized by Mirnov oscillations, andperturbation of the Ip, κ, and li evolution, have been observed. A low q limit was observed andis imposed by a low toroidal mode number kink instability.
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(EX3/D) Discussion of Section EX3
The file contains the discussion contributions relating to EX3/1, EX3/2, EX3/3, EX3/4(R),EX3/5, EX3/6.
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71
Session EX4 — PhysicsIntegration, Operation
Contents
(EX4/1) Sustainment of High Confinement in JT-60U Reversed Shear Plasmas 73
(EX4/1) Sustainment of High Confinement in JT-60U Re-versed Shear Plasmas
T. Fujita1), Y. Kamada1), S. Ide1), S. Takeji1), Y. Sakamoto1), A. Isayama1), T. Suzuki1),T. Oikawa1), T. Fukuda1)
1) Japan Atomic Energy Research Institute, Naka Fusion Research Establishment, Naka-machi,Ibaraki-ken, Japan
Abstract. Experimental results towards long sustainment of JT-60U reversed shear plasmas,where high confinement is achieved owing to strong internal transport barriers (ITBs), are re-ported. In a high current plasma with an L-mode edge, deuterium-tritium-equivalent fusionpower gain, Qeq
DT = 0.5 was sustained for 0.8 s (∼ energy confinement time) by adjusting plasmabeta precisely using feedback control of stored energy. In a high triangularity plasma with anELMy H-mode edge, the shrinkage of reversed shear region was suppressed and quasi steadysustainment of high confinement was achieved by raising the poloidal beta and enhancing thebootstrap current peaked at the ITB layer. High bootstrap current fraction (∼ 80%) was ob-tained in a high q regime (q95 ∼ 9), which leaded to full non-inductive current drive condition.The normalized beta (βN) of ∼ 2 and H-factor of H89 ∼ 3.5 (HH98y2 ∼ 2.2) were sustained for2.7 s (∼ 6 times energy confinement time).
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(EX4/2) Internal Barrier Discharges in JET and their Sen-sitivity to Edge Conditions
Abstract. Experiments in JET have concentrated on steady state discharges with internaltransport barriers. The internal transport barriers are formed during the current rise phase ofthe discharge with low magnetic shear in the centre and with high additional heating power.In order to achieve stability against disruptions at high pressure peaking, typical for ITB dis-charges, the pressure profile can be broadened with a H-mode transport barrier at the edge ofthe plasma. However, the strong increase in edge pressure during an ELM free H-mode weakensthe internal transport barrier due to a reduction of the rotational shear and pressure gradientat the ITB location. In addition, type I ELM activity, associated with a high edge pedestalpressure, leads to a collapse of the ITB with the input powers available in JET. The best ITBdischarges are obtained with input power control to reduce to core pressure, and with the edgeof the plasma controlled by argon gas dosing. These discharges achieve steady conditions forseveral energy confinement times with H97 confinement enhancement factors of 1.2-1.6 at lineaverage densities around 30%-40% of the Greenwald density. This is at much lower density (typ-ically factor 2 to 3) compared to standard H-mode discharges in JET. Increasing the density,using additional deuterium gas dosing or shallow pellet fueling has not been successful so far. Apossible route to higher densities should maintain the type III ELM’s towards high edge density,giving scope for future experiments in JET.
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(EX4/3) Long-Pulse High-Performance Discharges in theDIII-D Tokamak
T. C. Luce1), M. R. Wade2), P. A. Politzer1), S. L. Allen3), M. E. Austin4), D. R. Baker1),B. Bray1), D. P. Brennen5)1), K. H. Burrell1), T. A. Casper3), M. S. Chu1), J. C. DeBoo1),E. J. Doyle6), J. R. Ferron1), A. M. Garofalo7), P. Gohil1), I. A. Gorelov8), C. M. Green-field1), R. J. Groebner1), W. W. Heidbrink9), C.-L. Hsieh1), A. W. Hyatt1), R. Jayakumar3),J. E. Kinsey10), R. J. La Haye1), L. L. Lao1), C. J. Lasnier3), E. Lazarus2), A. W. Leonard1),Y. R. Lin-Liu1), J. Lohr1), M. A. Mahdavi1), M. A. Makowski3), M. Murakami2), C. C. Petty1),R. I. Pinsker1), R. Prater1), C. L. Rettig6), T. L. Rhodes6), B. W. Rice3), E. J. Strait1), T. S. Tay-lor1), D. M. Thomas1), A. D. Turnbull1), J. G. Watkins11), W. P. West1), and K. L. Wong8)
1) General Atomics, P.O. Box 85608, San Diego, California USA2) Oak Ridge National Laboratory, Oak Ridge, Tennessee USA3) Lawrence Livermore National Laboratory, Livermore, California USA4) University of Texas-Austin, Austin, Texas USA5) ORISE, Oak Ridge, Tennessee USA6) University of California-Los Angeles, California USA7) Columbia University, New York, New York USA8) Princeton Plasma Physics Laboratory, Princeton, New Jersey USA9) University of California, Irvine, California USA10) Lehigh University, Bethlehem, Pennsylvania USA11) Sandia National Laboratories, Albuquerque, New Mexico USA
Abstract. Significant progress in obtaining high performance discharges for many energy con-finement times in the DIII-D tokamak has been realized since the previous IAEA meeting. Inrelation to previous discharges, normalized performance ∼ 10 has been sustained for > 5τEwith qmin > 1.5. (The normalized performance is measured by the product βNH89 indicatingthe proximity to the conventional β limits and energy confinement quality, respectively.) TheseH-mode discharges have an ELMing edge and β . 5%. The limit to increasing β is a resistivewall mode, rather than the tearing modes previously observed. Confinement remains good de-spite the increase in q. The global parameters were chosen to optimize the potential for fullynon-inductive current sustainment at high performance, which is a key program goal for theDIII-D facility in the next two years. Measurement of the current density and loop voltageprofiles indicate ∼ 75% of the current in the present discharges is sustained non-inductively.The remaining ohmic current is localized near the half radius. The electron cyclotron heatingsystem is being upgraded to replace this remaining current with ECCD. Density and β control,which are essential for operating advanced tokamak discharges, were demonstrated in ELMingH-mode discharges with βNH89 ∼ 7 for up to 6.3 s or ∼ 34τE. These discharges appear to be inresistive equilibrium with qmin ∼ 1.05, in agreement with the current profile relaxation time of1.8 s.
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(EX4/4) Performance, Heating, and Current Drive Scenar-ios of ASDEX Upgrade Advanced Tokamak Discharges
R. C. Wolf1), J. Hobirk1), G. Conway1), O. Gruber1), A. Gude1), S. Gunter1), K. Kirov1),B. Kurzan1), M. Maraschek1), P. McCarthy2), H. Meister1), F. Leuterer1), G. V. Pereverzev1),E. Poli1), F. Ryter1) and the ASDEX Upgrade Team
1) Max-Planck Insitut fur Plasmaphysik, EURATOM-Association, D-85748 Garching2) Physics Department, University College Cork, Association Euratom-DCU, Cork Ireland
Abstract. Various approaches to high performance and steady state operation are presented.Strong neutral beam heating in the current ramp leads to internal transport barriers (ITBs) inconjunction with negative central magnetic shear. In high βp discharges full non-inductive cur-rent drive has been achieved transiently. At Greenwald density values of βp = 3.1, βN = 2.8, andHITER89−P = 1.8 have been reached simultaneously. In plasmas with ITBs and L-mode edgeadditional electron cyclotron resonance heating (ECRH) and current drive (ECCD) facilitateshigh core confinement with Te ≈ Ti. Central ECCD has been applied in low density, low currentdischarge. For ECCD in co-current direction a current drive fraction of 82% is calculated. In caseof counter-ECCD negative central shear with qmin > 1 leads to the formation of an electron ITB.
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(EX4/5) Experimental Studies toward Long-Pulse Steady-state Operations in LHD
N. Noda1), Y. Nakamura1), Y. Takeiri1), T. Mutoh1), R. Kumazawa1), M. Sato1), K. Kawa-hata1), S. Yamada1), T. Shimozuma1), Y. Oka1), A. Iiyoshi, R. Sakamoto1), Y. Kubota1),S. Masuzaki1), S. Inagaki1), T. Morisaki1), H. Suzuki1), N. Ohyabu1), K. Adachi1), K. Akaishi1),N. Ashikawa2), H. Chikaraishi1), P. de Vries, M. Emoto1), H. Funaba1), M. Goto1), S. Ham-aguchi1), K. Ida1), H. Idei1), K. Ikeda1), S. Imagawa1), N. Inoue1), M. Isobe1), A. Iwamoto1),S. Kado, O. Kaneko1), S. Kitagawa1), K. Khlopenkov1), T. Kobuchi2), A. Komori1), S. Kubo1),Y. Liang2), R. Maekawa1), T. Minami1), T. Mito1), J. Miyazawa1), S. Morita1), K. Murai1),S. Murakami1), S. Muto1), Y. Nagayama1), H. Nakanishi1), K. Narihara1), A. Nishimura1),K. Nishimura1), A. Nishizawa1), T. Notake1), S. Ohdachi1), M. Okamoto1), M. Osakabe1),T. Ozaki1), R. O. Pavlichenko1), B. J. Peterson1), A. Sagara1), K. Saito3), S. Sakakibara1),H. Sasao2), M. Sasao1), K. Sato1), T. Seki1), M. Shoji1), H. Sugama1), K. Takahata1), M. Takechi2),H. Tamura1), N. Tamura2), K. Tanaka1), K. Toi1), T. Tokuzawa1), Y. Torii3), K. Tsumori1),K. Y. Watanabe1), T. Watanabe1), T. Watari1), N. Yanagi1), I. Yamada1), H. Yamada1), S. Ya-maguchi1), S. Yamamoto3), T. Yamamoto3), M. Yokoyama1), Y. Yoshimura1), I. Ohtake1),R. Akiyama1), K. Haba1), M. Iima1), J. Kodaira1), K. Tsuzuki1), K. Itoh1), K. Matsuoka1),K. Ohkubo1), S. Satoh1), T. Satow1), S. Sudo1), S. Tanahashi1), K. Yamazaki1), O. Motojima1),Y. Hamada1), M. Fujiwara1)
1) National Institute for Fusion Science, Toki 509-5292, Japan2) Department of Fusion Science, The Graduate University for Advanced Studies, Toki 509-5292,Japan3) Dept. of Energy Engineering and Science, Graduate School of Engineering, Nagoya University,Nagoya 464, Japan
Abstract. Stable discharges longer than one minute have been obtained in LHD with all theheating schemes including electron cyclotron heating (ECH). Plasma is sustained with neutralbeam injection (NBI) or with ion cyclotron resonance frequency (ICRF) with 0.5–1 MW. Centralplasma temperature is higher than 1.5 keV with a density of 1 − 2 × 1019m−3 until the end ofthe pulse. Full installation of the carbon divertor has contributed to this achievement. Thisgives a sufficient base for physics and technology studies from the next campaign. The longpulse operation indicates new possibilities in diagnostics and in physics studies. Higher accu-racy and reliability is obtained with diagnostics parameter scan, longer integration of signalsor two-dimensional measurement. The mechanism of a slow oscillation called breathing is dis-cussed. Hydrogen recycling analysis has been carried out and preliminary results are obtained.Based on these results, the future program is divided into two categories, that is, i) physics andtechnology experiments utilizing long-pulse discharges up to 5 minutes, and ii) extension of thepulse-length toward one hour.
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(EX4/6) Confinement in TPE-RX Reversed Field Pinch
Y. Yagi1), T. Bolzonella2), A. Canton2), K. Hayase1), Y. Hirano1), S. Kiyama1), H. Koguchi1),Y. Maejima1), J.-A. Malmberg3), H. Sakakita1), Y. Sato1), S. Sekine1), T. Shimada1), K. Sug-isaki1)
1) Electrotechnical Laboratory, Tsukuba 305-8568 Japan2) Consorzio RFX, Padua 35127 Italy3) Royal Institute of Technology, Stockholm 100 44 Sweden
Abstract. Characteristics of the confinement properties of a reversed field pinch (RFP), theTPE-RX (R/a = 1.72/0.45 m, R and a are major and minor radii), are presented for the plasmacurrent, Ip of 0.2–0.4 MA. TPE-RX has been operational since 1998, and Ip = 0.5 MA andduration time of up to 0.1 s have been obtained separately. It is found that Ip/N (=12× 10−14
Am, N is the line density) is higher than those of other RFPs and poloidal beta, βp, and energyconfinement time, τE, are 5–10% and 0.5–1 ms, respectively, which are comparable with thoseof other RFPs of comparable sizes (RFX and MST). Pulsed poloidal current drive has recentlybeen tested and the result has shown a twofold improvement of βp and τE. The improvement isdiscussed in terms of the dynamic trajectories in the F −Θ plane, where F and Θ are reversaland pinch parameters, respectively. The global confinement properties are compared betweenthe locked and nonlocked discharges, which yields a better understanding of the mode-lockingphenomena in RFP plasmas.
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78
(EX4/D) Discussion of Section EX4
The file contains the discussion contributions relating to EX4/1, EX4/2, EX4/3, EX4/4, EX4/5,EX4/6.
(EX5/1) Physics of Confinement Improvement of Plasmaswith Impurity Injection in DIII-D
M. Murakami1), G. R. McKee2), G. L. Jackson3), G. M. Staebler3), D. A. Alexander4), D. R. Baker3),G. Bateman5), L. R. Baylor1), J. A. Boedo6), N. H. Brooks3), K. H. Burrell3), J. R. Cary4)7),R. H. Cohen8), R. J. Colchin1), J. C. DeBoo3), E. J. Doyle9), D. R. Ernst10), T. E. Evans3),C. Fenzi2), C. M. Greenfield3), D. E. Greenwood1), R. J. Groebner3), J. Hogan1), W. A. Houl-berg1), A. W. Hyatt3), R. J. La Haye3), R. Jayakumar8), T. C. Jergigan1), R. A. Jong8), J. E. Kin-sey5), A. H. Kritz5), L. L. Lao3), C. J. Lasnier8), M. A. Makowski8), A. Messiaen11), J. Man-drekas12), R. A. Moyer6), J. Ongena11), A. Pankin5), T. W. Petrie3), C. C. Petty3), C. L. Rettig9),T. L. Rhodes9), B. W. Rice8), D. W. Ross13), J. C. Rost14), S. Shasharina4), W. M. Stacey12),H. E. St John3), P. Strand1), R. D. Sydora15), T. S. Taylor3), D. M. Thomas3), M. R. Wade1),R. E. Waltz3), W. P. West3), K. L. Wong10), L. Zeng9), and the DIII-D Team
1) Oak Ridge National Laboratory, Oak Ridge, Tennessee USA2) University of Wisconsin, Madison, Wisconsin USA3) General Atomics, P.O. Box 85608, San Diego, California USA4) Tech-X Corporation, Boulder, Colorado USA5) Lehigh University, Bethlehem, Pennsylvania USA6) University of California, San Diego, California USA7) University of Colorado, Boulder, Colorado USA8) Lawrence Livermore National Laboratory, Livermore, California USA9) University of California, Los Angeles, California USA10) Princeton Plasma Physics Laboratory, Princeton, New Jersey USA11) KMS/ERM, Brussels, Belgium12) Georgia Institute of Technology, Atlanta, Georgia USA13) University of Texas, Austin, Texas USA14) Massachussetts Institute of Technology, Cambridge, Massachusetts USA15) University of Alberta, Edmonton, Canada
Abstract. External impurity injection into L-mode edge discharges in DIII-D has producedclear increases in confinement (factor of 2 in energy confinement and neutron emission), re-duction in all transport channels (particularly ion thermal diffusivity to the neoclassical level),and simultaneous reduction of long-wavelength turbulence. Suppression of the flux wavelengthturbulence and transport reduction are attributed to synergistic effects of impurity-induced en-hancement of ExB shearing rate and reduction of toroidal drift wave turbulence. A promptreduction of density fluctuations and local transport at the beginning of impurity injection ap-pears to result from an increased gradient of toroidal rotation enhancing the ExB shearing.Transport simulations carried out using the National Transport Code Collaboration Demonstra-tion Code with a GyroLandau fluid model, GLF23, indicate ExB shearing suppression is thedominant transport suppression mechanism.
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(EX5/2) Improved Energy Confinement at Trans Green-wald Densities in Discharges with a Radiating Edge in theTokamak TEXTOR-94
B. Unterberg1), G. Mank1), A. Messiaen2), J. Ongena2), S. Brezinsek1), V. Dreval4), P. Du-mortier2), R. Jaspers3), D. Kalupin2), H. R. Koslowski1), A. Kramer-Flecken1), A. Kreter1),M. Lehnen1), A. Pospieszczyk1), J. Rapp1), U. Samm1), B. Schweer1), G. Sergienko1), S. Solda-tov4), M. Z. Tokar1), G. Van Wassenhove2), R. R. Weynants2) and the TEXTOR Team1)2)3)
1) Institut fur Plasmaphysik, Forschungszentrum Julich, Ass. “EURATOM- FZ Julich”,D-52425 Julich, Germany2) Laboratoire de Physique des Plsamas - Laboratorium voor Plasmafysica, Ass. “EURATOM-Belgian State”, ERM- KMS, B-1000 Brussels, Belgium3) FOM Instituut voor Plasmafysica Rijnhuizen, Ass. “EURATOM-FOM”, NL-3430 BE, Nieuwegein,The Netherlands4) Nuclear Fusion Institute, Russian Research Centre “Kurchatov Institute”, Kurchatov Square1, 123182 Moscow, Russia
Abstract. Confinement quality as good as in the ELM-free H-mode at plasma densities sub-stantially above the Greenwald density (up to ne/nGW = 1.4) has been obtained in dischargeswith a radiating boundary in the tokamak TEXTOR-94. This is achieved by optimising the gasfuelling rate of RI-mode discharges to avoid both a confinement back transition at the beta limitor a confinement degradation to L-mode levels as a consequence of a too strong gas puffing. Asuccessful increase of the density to values well above nGW without degradation is obtainedif the plasma density and the neutral pressure at the edge can be kept low as a result of amoderate gas fuelling. In discharges with a strong gas fuelling, high plasma edge density andneutral pressure builds up and the toroidal plasma rotation at the edge just inside the LCFSis reduced. Furthermore, measurements of density fluctuation spectra at the plasma boundaryindicate a qualitative change of edge turbulence with a significant increase of fluctuations below50 kHz. Under these conditions the edge density and the recycling flux at the main limiterstart to increase prior to the global degradation. Modelling of the profile evolution after stronggas fuelling with a 1-D particle transport code shows the re-appearance of the ion temperaturegradient driven mode in the plasma bulk, which is first suppressed in the transition from L-to RI-mode after impurity injection, and supports the experimental finding that the strong gasfuelling is the reason for the degradation.
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82
(EX5/3) High-Radiation and High-Density Experiments inJT-60U
H. Kubo1), S. Sakurai1), N. Asakura1), S. Konoshima1), H. Tamai1), S. Higashijima1), A. Saka-sai1), H. Takenaga1), K. Itami1), K. Shimizu1), T. Fujita1), Y. Kamada1), Y. Koide1), H. Shi-rai1), T. Sugie1), T. Nakano1), N. Oyama1), H. Urano2), T. Ishijima3), K. Hill4), D. R. Ernst4),A. W. Leonard5), the JT-60U Team1)
1) Japan Atomic Energy Research Institute, Naka-gun, Japan2) Hokkaido University, Sapporo, Japan3) Nagoya University, Nagoya, Japan4) Princeton Plasma Physics Laboratory, Princeton, USA5) General Atomics, San Diego, USA
Abstract. In order to obtain confinement improved plasmas with high radiation at high den-sity, Ar gas was injected into ELMy H-mode plasmas in JT-60U. A confinement improvementof HH98(y,2) ∼ 1 was obtained with a high radiation loss power fraction (∼ 80 %) at an electrondensity of ∼ 0.65nGW. The HH-factor was about 50% higher than that in plasmas without Arinjection.
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83
(EX5/4) Radiating Edge Plasma Experiments on JET
G. P. Maddison1), M. Brix2), R. V. Budny3), M. Charlet1), I. Coffey1), G. Cordey1), P. Du-mortier4), S. K. Erents1), N. C. Hawkes1), M. von Hellermann5), D. L. Hillis6), J. Hogan6),L. D. Horton7), L. C. Ingesson5), S. Jachmich2), G. L. Jackson8), A. Kallenbach7), H. R. Koslowski2),K. D. Lawson1), A. Loarte9), G. Matthews1), G. R. McKee8)10), A. Meigs1), A. Messiaen4), F. Mi-lani1), P. Monier-Garbet11), M. Murakami6)8), M. F. F. Nave12), J. Ongena4), M. E. Puiatti13),E. Rachlew14), J. Rapp2), S. Sharapov1), G. M. Staebler8), M. Stamp1), J. D. Strachan3),G. Telesca13), M. Z. Tokar2), B. Unterberg2), M. Valisa13), K.-D. Zastrow1), and EFDA-JET2000 workprogramme contributors
1) EURATOM/UKAEA Fusion Association, Culham, UK2) IPP, Forschungszentrum Julich GmbH, EURATOM Association, D-52425 Julich, Germany3) Princeton Plasma Physics Laboratory, Princeton University, NJ 08543, USA4) LPP/ERM-KMS, Association EURATOM-Belgian State, B-1000 Brussels, Belgium5) FOM-IVP, EURATOM Association, Postbus 1207, NL-3430 BE Nieuwegein, Netherlands6) ORNL, Oak Ridge, TN 37831-8072, USA7) Max-Planck IPP, EURATOM Association, D-85748 Garching, Germany8) DIII-D National Fusion Facility, San Diego, CA 92186-5698, USA9) EFDA-CSU, D-85748 Garching, Germany10) University of Wisconsin-Madison, Madison, Wisconsin, USA11) CEA Cadarache, F-13108 St Paul lez Durance, France12) CFN, EURATOM-IST Associacao,1096 Lisbon, Portugal13) Consorzio RFX, Corso Stati Uniti 4, 35127 Padova, Italy14) Association Euratom-NFR, KTH, Stockholm, Sweden
Abstract. Scaling to larger tokamaks of high confinement plasmas with radiating edges is beingstudied through internationally collaborative experiments on JET. Three different configurationshave been explored. A small number of limiter L-mode discharges have most closely repeated theapproach used on TEXTOR-94. Divertor L-modes at intermediate density have pursued tran-siently improved states found on DIII-D. An original scheme has also examined impurity seedingof higher density ELMy H-modes, formed either in a novel pumped-limiter like arrangement,or again in divertor geometry. While strongly peaked density profiles thought to be importantin TEXTOR-94 have not generally been produced, nevertheless beneficial effects have similarlyemerged in JET. Confinement up to H-mode quality, together with radiation fractions of ≈40%, has briefly been obtained in divertor L-modes. Most notably, ELMy H-mode confinementhas been sustained at densities close to the Greenwald level, with little change of central Zeff
but up to ≈ 60% radiation, in long, “afterpuff” phases following the end of main gas fuelling.Outstanding products of normalized confinement and normalized density H97 · fGwd ≈ 0.9 haveconsequently been achieved. Marked reductions in the frequency of accompanying ELMs aregenerally also induced.
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84
(EX5/5) Helium Exhaust and Forced Flow Effects withBoth-leg Pumping in W-shaped Divertor of JT-60U
A. Sakasai, H. Takenaga, S. Higashijima, H. Kubo, T. Nakano, H. Tamai, S. Sakurai, N. Akino,T. Fujita, N. Asakura, K. Itami, K. Shimizu, the JT-60U Team
Naka Fusion Research Establishment, Japan Atomic Energy Research Institute, Naka-machi,Japan
Abstract. The W-shaped divertor of JT-60U was modified from inner-leg pumping to both-leg pumping. After the modification, the pumping rate was improved from 3% with inner-legpumping to 5% with both-leg pumping in a divertor-closure configuration, which means bothseparatrixes close to the divertor slots. Efficient helium exhaust was realized in the divertor-closure configuration with both-leg pumping. A global particle confinement time of τ∗He = 0.4sand τ∗He/τE = 3 was achieved in attached ELMy H-mode plasmas. The helium exhaust efficiencywith both-leg pumping was extended by 45% as compared with inner-leg pumping. By usingcentral helium fueling with He-beam injection, the helium removal from the core plasma insidethe internal transport barrier (ITB) in reversed shear plasmas in the divertor-closure config-uration was investigated for the first time. The helium density profiles inside the ITB werepeaked as compared with those in ELMy H-mode plasmas. In the case of low recycling diver-tor, it was difficult to achieve good helium exhaust capability in reversed shear plasmas withITB. However, the helium exhaust efficiency was improved with high recycling divertor. Carbonimpurity reduction was observed by the forced flow with gas puff and effective divertor pumping.
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85
(EX5/6) Cross-Field Transport in the SOL: Its Relationshipto Main Chamber and Divertor Neutral Control in AlcatorC-Mod
B. Lipschultz1), B. LaBombard1), J. A. Goetz1), C. S. Pitcher1), N. Asakura2), R. L. Boivin1),J. Hughes1), A. Kallenbach3), G. M. McCracken4), G. Matthews4), D. Mossessian1), J. Rice1),J. L. Terry1), M. Umansky5)
1) M.I.T. Plasma Science and Fusion Center, 175 Albany St., Cambridge, MA 02139 USA2) Japan Atomic Energy Research Institute, Naka-Machi, Naka-gun, Ibareki-ken, Japan3) MPI fur Plasmaphysik, EURATOM Association, D-85748 Garching, Germany4) UKAEA Fusion, Culham Science Centre, Abingdon, Oxon., OX14 3EA5) Laboratory for Laser Energetics, 250 East River Rd., Rochester, NY 14623 USA
Abstract. The sources of neutrals at the outer midplane of the plasma are discussed. We findthat both the flux of neutrals escaping the divertor through leaks and ion recycling at mainchamber surfaces appear to contribute. The ion flux to the walls is larger than the flux enteringthe divertor and comparable to recycling at the divertor plate. The cause of these high wallion fluxes is an enhancement of cross-field particle transport which gives rise to substantial con-vective heat transport at higher densities. We have further explored main chamber recyclingand impurity transport utilizing a novel divertor ‘bypass’, which connects the outer divertorplenum to the main chamber. We find that leakage of neutrals (fuel and recycling impurities)from the divertor appears to be determined primarily by the conductance through the divertorstructure, thus indicating that tight baffling would be desirable in a reactor for fuel and heliumash compression.
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86
(EX5/D) Discussion of Section EX5
The file contains the discussion contributions relating to EX5/1, EX5/2, EX5/3, EX5/4, EX5/5.
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87
Session EX6 — Transport 2
Contents
(EX6/1) Electron Transport and Improved Confinement on Tore Supra . . 89
(EX6/2) Progress Towards Increased Understanding and Control of InternalTransport Barriers (ITBs) on DIII-D . . . . . . . . . . . . . . . . . . . . 90
(EX6/1) Electron Transport and Improved Confinement onTore Supra
G. T. Hoang1), C. Bourdelle1), X. Garbet1), T. Aniel1), G. Giruzzi1), M. Ottaviani1), W. Hor-ton2), P. Zhu2), R. V. Budny3)
1) Association EURATOM-CEA CEA-Cadarache, France2) Institute for Fusion Studies, The University of Texas, Austin, USA3) Princeton Plasma Physics Laboratory, Princeton, NJ08543 USA
Abstract. Magnetic shear is found to play an important role for triggering various improvedconfinement regimes through the electron channel. A wide database of hot electron plasmas(Te > 2Ti) heated by fast wave electron heating (FWEH) is analyzed for electron thermal trans-port. A critical gradient is clearly observed. It is found that the critical gradient linearly increaseswith the ratio between local magnetic shear (s) and safety factor (q). The Horton model, basedon the electromagnetic turbulence driven by the electron temperature gradient (ETG) mode, isfound to be a good candidate for electron transport modeling.
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89
(EX6/2) Progress Towards Increased Understanding andControl of Internal Transport Barriers (ITBs) on DIII-D
E. J. Doyle1), C. M. Greenfield2), M. E. Austin3), L. R. Baylor4), K. H. Burrell2), T. A. Casper5),J. C. DeBoo2), D. R. Ernst6), C. Fenzi7), P. Gohil2), R. J. Groebner2), W. W. Heidbrink8),G. L. Jackson2), T. C. Jernigan4), J. E. Kinsey9), L. L. Lao2), M. A. Makowski5), G. R. McKee10),M. Murakami4), W. A. Peebles1), R. Prater2), C. L. Rettig1), T. L. Rhodes1), J. C. Rost10),G. M. Staebler2), B. W. Stallard5), E. J. Strait2), E. Synakowski6), D. M. Thomas2), M. R. Wade4),R. E. Waltz2), and L. Zeng1)
1) Dept. of Electrical Engineering and IPFR, University of California, Los Angeles, Califor-nia USA2) General Atomics, San Diego, California USA3) University of Texas at Austin, Austin, Texas USA4) Oak Ridge National Laboratory, Oak Ridge, Tennessee USA5) Lawrence Livermore National Laboratory, Livermore, California USA6) Princeton Plasma Physics Laboratory, Princeton, New Jersey USA7) University of Wisconsin-Madison, Madison, Wisconsin USA8) Univeristy of California, Irvine, California USA9) Lehigh University, Bethlehem, Pennsylvania USA10) Massachusetts Institute of Technology, Cambridge, Massachusetts USA
Abstract. Substantial progress has been made towards both understanding and control ofinternal transport barriers (ITBs) on DIII-D, resulting in the discovery of a new sustained highperformance operating mode termed the Quiescent Double-Barrier (QDB) regime. The QDBregime combines core transport barriers with a quiescent, ELM-free H-mode edge (termed QH-mode), giving rise to separate (double) core and edge transport barriers. The core and edgebarriers are mutually compatible and do not merge, resulting in broad core profiles with an edgepedestal. The QH-mode edge is characterized by ELM-free behavior with continuous multihar-monic MHD activity in the pedestal region, and has provided density and impurity control for3.5 s (> 20τE) with divertor pumping. QDB plasmas are long-pulse high-performance candi-dates, having maintained a βNH89 product of 7 for 5 energy confinement times (Ti ≤ 16 keV,βN ≤ 2.9, H89 ≤ 2.4, τE ≤ 150 ms, DD neutron rate Sn ≤ 4× 1015s−1). The QDB regime hasonly been obtained in counter-NBI discharges (injection anti-parallel to plasma current) with di-vertor pumping. Other results include successful expansion of the ITB radius using (separately)both impurity injection and counter-NBI, and the formation of ITBs in the electron thermalchannel using both ECH and strong negative central shear (NCS) at high power. These resultsare interpreted within a theoretical framework in which turbulence suppression is the key to ITBformation and control, and a decrease in core turbulence is observed in all cases of ITB formation.
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90
(EX6/3) ECRH Results during Current Ramp-Up and Post-Pellet Injection in FTU Plasma
G. Bracco1), A. Bruschi1), P. Buratti1), S. Cirant2), F. Crisanti1), B. Esposito1), D. Frigione1),E. Giovannozzi1), G. Giruzzi4), G. Granucci2), V. Krivenski5), C. Sozzi2), O. Tudisco1), V. Zanza1),F. Alladio1), B. Angelini1), M. L. Apicella1), G. Apruzzese1), E. Barbato1), L. Bertalot1),A. Bertocchi1), G. Buceti1), A. Cardinali1), S. Cascino6), C. Castaldo1), C. Centioli1), R. Ce-sario1), P. Chuillon1), S. Ciattaglia1), V. Cocilovo1), R. De Angelis1), M. De Benedetti1), E. dela Luna5), F. De Marco1), B. Francioni6), L. Gabellieri1), G. Gatti1), C. Gormezano1), F. Gra-vanti1), M. Grolli2), F. Iannone1), H. Kroegler2), M. Leigheb1), G. Maddaluno1), G. Mafia1),M. Marinucci1), G. Mazzitelli1), P. Micozzi1), F. Mirizzi1), S. Nowak2), F. P. Orsitto1), D. Pacella1),L. Panaccione1), M. Panella1), F. Papitto1), V. Pericoli-Ridolfini1), L. Pieroni1), S. Podda1),F. Poli1), G. Pulcella6), G. Ravera1), G. B. Righetti1), F. Romanelli1), M. Romanelli1), A. Russo6),F. Santini1), Sassi 1), S. E. Segre3), A. Simonetto2), P. Smeulders1), S. Sternini1), N. Tartoni1),P. E. Travisanutto1), A. A. Tuccillo1), V. Vitale1), G. Vlad1), M. Zerbini1), F. Zonca1)
1) Associazione Euratom-ENEA sulla Fusione, CR Frascati, CP 65, Frascati, Roma, Italy2) Associazione Euratom-ENEA-CNR, Istituto di Fisica del Plasma, Milano, Italy3) INFM and Dipartimento di Fisica, II Universita di Roma “Tor Vergata”, Roma, Italy4) Association Euratom-CEA sur la Fusion, DRFC/STPF, CEA/Cadarache, France5) Asociacion Euratom/CIEMAT para Fusion, CIEMAT, Madrid, Spain6) ENEA Fellow
Abstract. Recent ECRH experiments in FTU have provided new results in two plasma scenar-ios, both characterized by the absence of the sawtooth activity and by flat or reversed q profiles.The first is the current ramp-up phase where low density plasmas have been heated up to highelectron temperature. When the heating is localized on the plasma axis, the high additionalpower density has produced the evidence of a deformation of the bulk of the local electron dis-tribution function, which is in agreement with the results of a detailed kinetic simulation. Whenoff-axis heating is applied, no clear evidence of non-diffusive energy transport has been found.In the second scenario, ECRH has been applied on the high density plasma produced by pelletinjection, resulting in strong ion heating as shown by the increase of the neutron yield. Theanalysis of this scenario shows that, when the post pellet phase is MHD quiescent, an enhancedenergy confinement regime can be obtained with ECRH as found previously in ohmically heatedpost-pellet plasmas.
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91
(EX6/4) Characteristics of Internal Transport Barrier inJT-60U Reversed Shear Plasmas
Y. Sakamoto1), Y. Kamada1), S. Ide1), T. Fujita1), H. Shirai1), T. Takizuka1), Y. Koide1),T. Fukuda1), T. Oikawa1), T. Suzuki1), K. Shinohara1), R. Yoshino1) and the JT-60U Team1)
1) Naka Fusion Research Establishment, Japan Atomic Energy Research Institute, Japan
Abstract. Characteristics of internal transport barrier (ITB) structure are studied and theactive ITB control has been developed in JT-60U reversed shear plasmas. The following resultsare found. Outward propagation of the ITB with steep Ti gradient is limited to the minimumsafety factor location (ρqmin). However the ITB with reduced Ti gradient can move to the out-side of ρqmin. Lower boundary of ITB width is proportional to the ion poloidal gyroradius at theITB center. Furthermore the demonstration of the active control of the ITB strength based onthe modification of the radial electric field shear profile is successfully performed by the toroidalmomentum injection in different directions or the increase of heating power by neutral beams.
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92
(EX6/5) Relating Nondimensional Scaling of Microturbu-lence to Global Confinement and Transport
G. R. McKee1), C. C. Petty2), R. E. Waltz2), C. Fenzi1), R. J. Fonck1), K. H. Burrell2),D. R. Baker2), E. J. Doyle3), X. Garbet4), J. E. Kinsey5), T. C. Luce2), R. A. Moyer6), C. L. Ret-tig3), T. L. Rhodes3), D. R. Ross7), G. M. Staebler2), R. D. Sydora8), M. R. Wade9)
1) University of Wisconsin-Madison, Madison, WI, United States of America2) General Atomics, San Diego, California, USA3) University of California, Los Angeles, California, USA4) Association Euratom-CEA sur la Fusion, Cadarache, France5) Lehigh University, Bethlehem, Pennsylvania, USA6) University of California, San Diego, California, USA7) University of Texas-Austin, Austin, Texas, USA8) University of Alberta, Edmonton, Alberta, Canada9) Oak Ridge National Laboratory, Oak Ridge, Tennessee, USA
Abstract. Plasma turbulence characteristics, including radial correlation lengths, decorrela-tion times, amplitude profile, and flow velocity, have been measured during a ρ∗ scan on DIII-Dwhile all other transport-relevant dimensionless quantities (β, ν∗, κ, q,Te/Ti ...) are held nearlyconstant. The turbulence is measured by examining the correlation properties of the local long-wavelength (k⊥ρI ≤ 1) density fluctuations, measured with beam emission spectroscopy. Theradial correlation length of the turbulence, Lc,r, is shown to scale with the local ion gyroradius,with Lc,r ≈ 5ρI, while the decorrelation times scale with the local acoustic velocity as τc ∼ a/cs.The turbulent diffusivity parameter, D ∼ (L2
c,r/τc), scales in a roughly gyro-Bohm-like fashion,as predicted by the gyrokinetic equations governing turbulent transport. The experimental one-fluid power balance heat diffusivity scaling and that from GLF23 modeling compare reasonablywell.
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93
(EX6/6) Observation of Bifurcation Property of Radial Elec-tric Field Using a Heavy Ion Beam Probe in CHS
A. Fujisawa1), H. Iguchi1), T. Minami1), Y. Yoshimura1), H. Sanuki1), K. Itoh1), M. Isobe1),C. Suzuki1), S. Nishimura1), K. Tanaka1), M. Osakabe1), I. Nomura1), K. Ida1), S. Okamura1),K. Toi1), S. Kado1), R. Akiyama1), A. Shimizu1), C. Takahashi1), M. Kojima1), K. Matsuoka1),Y. Hamada1), M. Fujiwara1)
1) National Institute for Fusion Science, Oroshi-cho, Toki, Japan 509-5292
Abstract. Bifurcation nature of potential profile of a toroidal helical plasma is investigatedin the Compact Helical System (CHS), using a heavy ion beam probe. The measurements re-veal that there exist three main branches of potential profiles in electron cyclotron resonance(ECR) heated plasmas with low density of ne ∼ 0.5× 1013cm−3. The branches with higher cen-tral potential exhibit a rather strong radial electric field shear that should result in fluctuationreduction and formation of transport barrier. Lissajous expression is useful to extract the bifur-cation characteristics of potential structure.
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94
(EX6/7) Energy Confinement and Thermal Transport Char-acteristics of Net-Current Free Plasmas in Large HelicalDevice
H. Yamada1), K. Y. Watanabe1), K. Yamazaki1), S. Murakami1), S. Sakakibara1), K. Nar-ihara1), K. Tanaka1), M. Osakabe1), K. Ida1), N. Ashikawa2), P. de Vries1), M. Emoto1),H. Funaba1), M. Goto1), H. Idei1), K. Ikeda1), S. Inagaki1), N. Inoue1), M. Isobe1), S. Kado1),O. Kaneko1), K. Kawahata1), K. Khlopenkov1), A. Komori1), S. Kubo1), R. Kumazawa1), S. Ma-suzaki1), T. Minami1), J. Miyazawa1), T. Morisaki1), S. Morita1), S. Muto1), T. Mutoh1),Y. Nagayama1), N. Nakajima1), Y. Nakamura1), H. Nakanishi1), K. Nishimura1), N. Noda1),T. Notake3), T. Kobuchi2), Y. Liang2), S. Ohdachi1), N. Ohyabu1), Y. Oka1), T. Ozaki1),R. O. Pavlichenko1), B. J. Peterson1), G. Rewoldt4), A. Sagara1), K. Saito3), R. Sakamoto1),H. Sasao2), M. Sasao1), K. Sato1), M. Sato1), T. Seki1), T. Shimozuma1), M. Shoji1), H. Sugama1),H. Suzuki1), M. Takechi1), Y. Takeiri1), N. Tamura2), K. Toi1), T. Tokuzawa1), Y. Torii3),K. Tsumori1), I. Yamada1), S. Yamaguchi1), S. Yamamoto3), M. Yokoyama1), Y. Yoshimura1),T. Watari1), K. Itoh1), K. Matsuoka1), K. Ohkubo1), I. Ohtake1), S. Satoh1), T. Satow1),S. Sudo1), S. Tanahashi1), T. Uda1), Y. Hamada1), O. Motojima1), M. Fujiwara1)
1) National Institute for Fusion Science, Toki, Japan2) Department of Fusion Science, Graduate Univ. for Advanced Studies, Hayama, Japan3) Department of Energy Engineering and Science, Nagoya University, Japan4) Princeton Plasma Physics Laboratory, Princeton Univ. Princeton, USA
Abstract. The energy confinement and thermal transport characteristics of net-current freeplasmas in the much smaller gyro-radii and collisionality regimes than before have been inves-tigated in the Large Helical Device (LHD). The inward shifted configuration that is superiorfrom the theoretical aspect of neoclassical transport has revealed a systematic confinement im-provement on a standard configuration. The improvement of energy confinement times on theinternational stellarator scaling 95 occurs with a factor of 1.6 ± 0.2 for an inward shifted con-figuration. This enhancement is primarily due to the broad temperature profile with a highedge value. A simple dimensional analysis involving LHD and other medium sized heliotronsyields strongly gyro-Bohm dependence (τEΩ ∝ ρ∗−3.8) of energy confinement times. It shouldbe noted that this result is attributed to comprehensive treatment of LHD for systematic con-finement enhancement and that the medium sized heliotrons have narrow temperature profiles.The core stored energy still indicates the dependence of τEΩ ∝ ρ∗−2.6 when data only fromLHD is processed. The local heat transport analysis of dimensionally similar discharges exceptfor ρ∗ suggests that the heat conduction coefficient lies between Bohm and gyro-Bohm in thecore and changes towards strong gyro-Bohm in the peripheral region. Since the inward shiftedconfiguration has a geometrical feature suppressing the neoclassical transport, confinement im-provement can be maintained in the collisionless regime where the ripple transport is important.The stiffness of the pressure profile coincides with enhanced transport in the peaked densityprofile obtained by pellet injection.
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95
(EX6/D) Discussion of Section EX6
The file contains the discussion contributions relating to EX6/1, EX6/2, EX6/3, EX6/4, EX6/5,EX6/6, EX6/7.
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96
Session EX7/TH5/IC — Stability2
Contents
(EX7/1) High Beta Plasmas and Internal Barrier Dynamics in JET Dis-charges with Optimised Shear . . . . . . . . . . . . . . . . . . . . . . . . . 98
(EX7/1) High Beta Plasmas and Internal Barrier Dynamicsin JET Discharges with Optimised Shear
P. Buratti1) and the JET Team2)
1) Associazione EURATOM-ENEA sulla Fusione, Frascati, Italy2) Abingdon, Oxfordshire, United Kingdon
Abstract. High βN conditions have been sustained in JET discharges with Internal TransportBarrier by controlling edge conditions and pressure peaking. The behaviour of ideal and resistiveMHD instabilities at high βN has been studied as a function of pressure peaking. Duration ofthe high bN phase was limited by interaction with the septum of the Gas Box divertor. ITBtriggering and evolution are in good agreement with the condition for turbulence suppressionby ExB shear flow.
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98
(EX7/2) Resistive Instabilities in Reversed Shear Dischargesand Wall Stabilization on JT-60U
S. Takeji1), S. Tokuda1), T. Fujita1), T. Suzuki1), A. Isayama1), S. Ide1), Y. Ishii1), Y. Kamada1),Y. Koide1), T. Matsumoto1), T. Oikawa1), T. Ozeki1), Y. Sakamoto1), the JT-60U Team1)
1) Japan Atomic Energy Research Institute, Ibaraki, Japan
Abstract. Resistive instabilities and wall stabilization of ideal low toroidal mode number, n,kink modes are investigated in JT-60U reversed shear discharges. Resistive interchange modeswith n = 1 are found to appear in reversed shear discharges with large pressure gradient atthe normalized beta, βN, of about unity or even lower. The resistive interchange modes appearas intermittent burst-like magnetohydrody-namic (MHD) activities and higher n ≤ 3 modesare observed occasionally in higher βN regime. No clear degradation of the plasma stored en-ergy is observed by the resistive interchange modes themselves. It is also found that resistiveinterchange modes can lead to major collapse owing to a coupling with tearing modes at theouter mode rational surface over the minimum safety factor. Stability analysis revealed thatstability parameter of tearing modes, ∆′, at the outer mode rational surface is affected by thefree-boundary condition. The result is consistent with the experimental evidence that majorcollapse tends to occur when plasma edge safety factor, q∗, is near integer values. Stabilizationof ideal low n kink modes by the JT-60U wall is demonstrated. Magnetohydrodynamic pertur-bations that are attributed to resistive wall modes are observed followed by major collapse inwall-stabilized discharges.
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99
(EX7/3) MHD Phenomena in Advanced Scenarios on AS-DEX Upgrade and the Influence of Localised Electron Heat-ing and Current Drive
S. Gunter1), A. Gude1), J. Hobirk1), M. Maraschek1), A. G. Peeters1), S. D. Pinches1), S. Saarelma2),S. Schade1), R. C. Wolf1), ASDEX Upgrade Team1)
1) MPI fur Plasmaphysik, D-85748 Garching, Germany, EURATOM-Association2) Helsinki Univ. of Tech., EURATOM-TEKES Association, FIN-02015 HUT, Finland
Abstract. MHD instabilities in advanced tokamak scenarios on the one hand are favourableas they can contribute to the stationarity of the current profiles and act as a trigger for theformation of internal transport barriers. In particular fishbone oscillations driven by fast parti-cles arising from neutral beam injection (NBI) are shown to trigger internal transport barriersin low and reversed magnetic shear discharges. During the whistling down period of the fish-bone oscillation the transport is reduced around the corresponding rational surface, leading toan increased pressure gradient. This behaviour is explained by the redistribution of the reso-nant fast particles resulting in a sheared plasma rotation due to the return current in the bulkplasma, which is equivalen to a radial electric field. On the other hand MHD instabilities limitthe accessible operating regime. Ideal and resistive MHD modes such as double tearing modes,infernal modes and external kinks degrade the confinement or even lead to disruptions in AS-DEX Upgrade reversed shear discharges. Localized electron cyclotron heating and current driveis shown to significantly affect the MHD stability of this type of discharges.
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100
(EX7/4) Effects of Toroidal Currents upon Magnetic Con-figurations and Stability in Wendelstein 7-AS
A. Weller1), M. Anton1), R. Brakel1), J. Geiger1), M. Hirsch1), R. Jaenicke1), S. Klose1), E. Sal-lander2), A. Werner1), W7-AS Team1), IPP-NBI Group1)
1) Max-Planck-Institut fur Plasmaphysik, Garching, Germany2) Alfven Laboratory, Royal Institute of Technology, Stockholm, Sweden
Abstract. The proposal of new concepts for current carrying hybrid stellarators has raisedthe issue if current driven instabilities, in particular major disruptions, may be suppressed ormitigated by the externally provided poloidal magnetic field. In W7-AS the internal toroidalcurrents such as bootstrap and Okhawa currents are cancelled by opposite currents driven induc-tively or by electron cyclotron current drive (ECCD). In this way the edge rotational transformis controlled, and net current-free stable plasmas are maintained. On the other hand, the currentdrive systems provide a flexible tool to investigate current driven instabilities as well as variousissues concerning the effect of magnetic shear on confinement and MHD mode behaviour. Thestability studies in the presence of significant toroidal currents have been made in the accessiblerange of the external rotational transform ι-ext = 0.30. . . 0.56 involving the low order rationalsurfaces ι- = 1/2, 3/2, 3/4 and 1. In addition the rational surfaces ι- = 1/3 and 1/4 couldbe accessed by reverse current drive. Target plasmas heated by electron cyclotron resonanceheating (ECRH), neutral beam injection (NBI) or both were investigated in order to assess towhich extent the stability depends on particular current density profiles. Disruption-like events,preceded by tearing mode activity, have been observed in a wide range of the external rotationaltransform. The mode structures have been analyzed by X-ray tomography, electron cyclotronemission (ECE) diagnostics and magnetic measurements. The experimental data are roughlyconsistent with stability calculations on the basis of a cylindrical ∆′-analysis. In contrast tothe tokamak case the plasma equilibrium is maintained even after a thermal collapse enablinga recovery of plasma energy and inductive current. The improved positional stability can resultin the formation of very large magnetic islands. Severe disruption-like effects may be controlledby excluding relevant rational surfaces, in particular ι- = 1/2, from the outer plasma region.
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101
(EX7/5) Ultra-High-Beta Spherical Tokamak with Abso-lute Minimum-B in TS-3 Spherical Torus Experiments
Y. Ono1), M. Inomoto1), Y. Ueda1), M. Katsurai1), T. Matsuyama1), Y. Murata1), T. Tawara1)
1) High Temperature Plasma Center, University of Tokyo
Abstract. An absolute minimum-B configuration was obtained when applied external toroidalmagnetic field transformed a field-reversed configuration (FRC) into an ultra-high-beta (> 50%)spherical tokamak (ST). High-power heating (≈ 5–30MW) of magnetic reconnection was usedto form the oblate FRC and to increase thermal pressure (beta) of STs. High-beta STs witha variety of pressure and q profiles revealed their stability boundaries and possible ballooningcollapses. The high-beta ST with maximum pressure gradient pΨ in the periphery was main-tained stably, while another high-beta ST with maximum pΨ in the core collapsed due to high-n(n > 8) localized modes.
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102
(TH5/1) Recent Development of Theory for W7-X
J. Nuhrenberg1), M. Drevlak1), R. Hatzky1), R. Kleiber1), A. Koenies1), P. Merkel1), D. Monti-cello2), C. Nuhrenberg1), A. Reiman2), T. M. Tran3)
1) Max-Planck-Institut fur Plasmaphysik, Germany2) Princeton Plasma Physics Laboratory, USA3) Centre de Recherches en Physique des Plasmas, Suisse
Abstract. The Stellarator Theory Division at the Greifswald Branch has been concentratingon widening the scope of theoretical work related to W7-X. Some of such areas are the qualityof finite-β magnetic surfaces, energetic-particle-driven Alfven eigenmode, resistively driven driftand ion-temperature-gradient-driven instabilities. New results pertaining to these issues are: i)MHD GAEs in W7-X-type equilibria are discovered and the first 3d drift-kinetic formulationof interaction with energetic particles is developed as nonlinear eigenvalue problem analyticallyand computationally; ii) for the first time, mode structures of globally calculated resistive driftinstabilities with poloidal mode numbers of O(103) and exhibiting the relative importance oftoroidal vs. helical coupling are obtained in a toroidal stellarator; iii) for the first time, nonlinearsaturation levels of kinetic ITG modes are obtained with energy conservation in a θ-pinch, iv)fixed-boundary-PIES W7-X-type high-β equilibria – well converged on an NEC-SX5 – show even5/5 islands to be very thin.
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103
(IC/1) Physics Issues in the Design of Low Aspect Ratio,High-Beta, Quasi-Axisymmetric Stellarators
M. C. Zarnstorff1), L. A. Berry2), A. H. Boozer3), A. Brooks1), W. A. Cooper4), M. Drevlak5),E. Fredrickson1), G. Y. Fu1), R. Goldston1), R. Hatcher1), S. Hirshman2), W. A. Houlberg2),S. Hudson1), M. Isaev6), C. Kessel1), L. P. Ku1), E. Lazarus2), J. Lewandowski1), Z. Lin1),J. F. Lyon2), R. Majeski1), P. Merkel5), M. Mikhailov6), D. R. Mikkelsen1), W. Miner7), D. Mon-ticello1), H. Mynick1), G. H. Neilson1), B. E. Nelson2), C. Nuhrenberg5), N. Pomphrey1),M. Redi1), W. Reiersen1), A. Reiman1), P. Rutherford1), R. Sanchez8), J. Schmidt1), D. A. Spong2),P. Strand2), D. Strickler2), S. Subbotin6), P. Valanju7), R. White1)
1) Princeton Plasma Physics Laboratory, Princeton, NJ 08543 USA2) Oak Ridge National Laboratory, Oak Ridge, TN 37831 USA3) Columbia University, New York, NY 10027 USA4) Ecole Polytechnique Federale de Lausanne, Lausanne, Switzerland5) Max Planck Institute for Plasma Physics, Greifswald, Germany6) Kurchatov Institute, Moscow, Russia7) University of Texas at Austin, Austin, TX 78712 USA8) Universidad Carlos III de Madrid, Spain
Abstract. Compact stellarators have the potential to combine the best features of the stel-larator and the advanced tokamak, offering steady state operation without current drive andpotentially without disruptions at an aspect ratio similar to tokamaks. A quasi-axisymmetricstellarator is developed that is consistent with the bootstrap current and passively stable to theballooning, kink, Mercier, vertical, and neoclassical tearing modes at β = 4.1% without needfor conducting walls or external feedback. The configuration has good flux surfaces and fast ionconfinement. Thermal transport analysis indicates that the confinement should be similar totokamaks of the same size, allowing access to the β-limit with moderate power. Coils have beendesigned to reproduce the physics properties. Initial analysis indicates the coils have consider-able flexibility to manipulate the configuration properties. Simulations of the current evolutionindicate the kink-mode can remain stable during the approach to high-beta.
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104
(EX7/D) Discussion of Section EX7/TH5/IC
The file contains the discussion contributions relating to EX7/1, EX7/2, EX7/3, IC/1, EX7/4.
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105
Session EX8 — Current Drive,Heating & Fuelling
Contents
(EX8/1) Highly Localized Electron Cyclotron Heating and Current Driveand Improved Core Transport in DIII-D . . . . . . . . . . . . . . . . . . 107
(EX8/1) Highly Localized Electron Cyclotron Heating andCurrent Drive and Improved Core Transport in DIII-D
R. Prater1), M. E. Austin2), S. Bernabei3), K. H. Burrell1), R. W. Callis1), W. P. Cary1),J. S. deGrassie1), C. Fuchs4), C. M. Greenfield1), Y. Gorelov1), R. W. Harvey5), J. C. Hosea3),A. Isayama6), J. Jayakumar7), R. J. La Haye1), L. L. Lao1), R. A. Legg1), Y. R. Lin-Liu1),J. Lohr1), T. C. Luce1), M. A. Makowski7), C. C. Petty1), R. I. Pinsker1), D. Ponce1), S. G. E. Pronko1),S. Raftopoulos3), E. J. Strait1), and K. L. Wong3)
1) General Atomics, San Diego, California USA2) University of Texas-Austin, Austin, Texas USA3) Princeton Plasma Physics Laboratory, Princeton, New Jersey USA4) Max Planck Institute fur Plasmaphysik, Garching, Federal Republic of Germany5) CompX, Del Mar, California USA6) Japan Atomic Energy Research Institute, Naka-machi, Naka-gun, IbaRaki-ken, Japan7) Lawrence Livermore National Laboratory, Livermore, California USA
Abstract. Electron cyclotron heating (ECH) and current drive (ECCD) are widely recognizedas methods of depositing highly localized power and current in a plasma, based on calculationsof wave absorption. The experiments reported here demonstrate through direct analysis of thepoloidal field pitch angles measured by the motional Stark effect diagnostic that ECCD canbe as localized as theory predicts. This very narrow profile of driven current has been verifiedeven for ELMing H-mode discharges, and observation of full stabilization of neoclassical tearingmodes tends to corroborate the calculations of ECCD far off axis even in plasmas with largeMHD activity present. The electron heating by EC waves can have dramatic effects on theplasma, creating high central electron temperatures even with very modest ECH power andgenerating a strong transport barrier in the electron fluid in discharges with strongly reversedcentral magnetic shear. The electron diffusivity is much smaller than ion neoclassical diffusivityin the narrow barrier which develops just outside the heating location.
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107
(EX8/2) High Power Lower Hybrid Current Drive Experi-ment in TORE SUPRA Tokamak
Y. Peysson1) and the TORE SUPRA Team1)
1) Association Euratom-CEA pour la Fusion Controlee, CEA Cadarache, France
Abstract. A review of the Lower Hybrid (LH) current drive experiments carried out on theTORE SUPRA tokamak is presented. This work highlights the issues for an effective applicationof the LH wave at high power in reactor relevant conditions. Very promising performances havebeen obtained with the new launcher that is designed to couple up to 4 MW during 1000 s at apower density of 25 MWm−2. The heat load on the guard limiter of the antenna and the fastelectron acceleration in the near electric field of the grill mouth remain at a low level, whilethe mean reflection coefficient never exceeds 10%. The powerful diagnosis capabilities of thehard x-ray (HXR) fast electron bremsstrahlung tomography has led to significant progresses inthe understanding of the LH wave dynamics. The role of the fastest electrons driven by theLH wave is clearly identified. From HXR measurements, an increase of the LH current driveefficiency with the plasma current is predicted and confirmed by a direct determination at zeroloop voltage. LH power absorption is observed to be off-axis in almost all plasma conditions,and its radial width clearly depends of antenna phasing conditions. A correlation between theHXR profiles and the onset of an improved core confinement is identified in fully non-inductivedischarges. This regime ascribed to some vanishing of the magnetic shear is found to be tran-sient and usually ends when the minimum of the safety factor becomes very close to 2, leadingto a large MHD activity. Experimental observations and numerical simulations suggest thatLH power is absorbed in a few number of passes. However, besides toroidal mode coupling,additional mechanisms may likely contribute to a spectral broadening to the LH wave.
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108
(EX8/3) Reactor Relevant Current Drive and Heating byN-NBI on JT-60U
T. Oikawa1), Y. Kamada1), A. Isayama1), T. Fujita1), T. Suzuki1), N. Umeda1), M. Kawai1),M. Kuriyama1), L. R. Grisham2), Y. I. Ikeda1), K. Kajiwara1), K. Ushigusa1), K. Tobita1),A. Morioka1), M. Takechi1), T. Itoh1) and the JT-60U Team
1) Japan Atomic Energy Research Institute, Naka, Japan2) Princeton Plasma Physics Laboratory, USA.
Abstract. Current drive capability of negative-ion-based NBI (N-NBI) in JT-60U has beenextended to the reactor relevant regime. Driven current profile and current drive efficiency havebeen evaluated in the high temperature regime of central electron temperature Te(0) ∼ 10keV,and reasonable agreement with theoretical prediction has been confirmed in this regime. N-NBdriven current reached near 1MA with injection power of 3.75MW at beam energy of 360keV.The current drive efficiency of 1.55× 1019Am−2W−1 approaching to the ITER requirement wasachieved in the highβp H mode plasma with Te(0) ∼ 13keV. This current drive performancerealized the sustainment of high beta (βN = 2.5) and high confinement (HHy2 = 1.4) plasma infull current-driven condition at the plasma current of 1.5MA. The influence of instabilities onN-NB current drive was studied. When a burst like instability driven by N-NB occurred in thecentral region, reductions in loop voltage near magnetic axis and neutron production rate dueto loss of beam ions were observed although the lost driven current was at most ∼ 7% of thetotal driven current. When neoclassical tearing instability appeared in high beta plasmas, lossof beam ions was enhanced with increasing activity of instability and depended on the locationof instability.
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109
(EX8/4) The Performance of ICRF Heated Plasmas in LHD
T. Watari1), T. Mutoh1), R. Kumazawa1), T. Seki1), K. Saito2), Y. Torii2), Y. Zhao3), D. Hart-mann4), H. Idei1), S. Kubo1), K. Ohkubo1), M. Sato1), T. Shimozuma1), Y. Yoshimura1),K. Ikeda1), O. Kaneko1), Y. Oka1), M. Osakabe1), Y. Takeiri1), K. Tsumori1), N. Ashikawa5),P. de Vries1), M. Emoto1), A. Fukuyama6), H. Funaba1), M. Goto1), K. Ida1), S. Inagaki1),N. Inoue1), M. Isobe1), K. Itoh1), S. Kado1), K. Kawahata1), T. Kobuchi1), K. Khlopenkov1),A. Komori1), A. V. Krasilnikov7), Y. Liang5), S. Masuzaki1), K. Matsuoka1), T. Minami1),J. Miyazawa1), T. Morisaki1), S. Morita1), S. Murakami1), S. Muto1), Y. Nagayama1), Y. Naka-mura1), H. Nakanishi1), K. Narihara1), K. Nishimura1), N. Noda1), A. T. Notake2), S. Ohdachi1),N. Ohyabu1), H. Okada6), M. Okamoto1), T. Ozaki1), R. O. Pavlichenko1), B. J. Peterson1),A. Sagara1), S. Sakakibara1), R. Sakamoto1), H. Sasao1), M. Sasao5), K. Sato1), S. Satoh1),T. Satow1), M. Shoji1), S. Sudo1), H. Suzuki1), M. Takechi1), N. Tamura5), S. Tanahashi1),K. Tanaka1), K. Toi1), T. Tokuzawa1), K. Y. Watanabe1), T. Watanabe1), H. Yamada1), I. Ya-mada1), S. Yamaguchi1), S. Yamamoto2), K. Yamazaki1), M. Yokoyama1), Y. Hamada1), O. Mo-tojima1), M. Fujiwara1)
1) National Institute for Fusion Science, 322-6 Oroshi-cho, Toki, 509-5292 Japan2) Department of Energy Engineering and Science, Nagoya University, 464-8603, Japan3) Institute of Plasma Physics, Academia Scinica, 230031, Hefei, Anhui, China4) Max Planck Institute for Plasma Physics, D-85748,Garching, Germany5) Department of Fusion Science, School of Mathematical and Physical Science, Graduate Uni-versity for Advanced Studies, Hayama, 240-0193, Japan6) Kyoto University, 606-8187, Kyoto, Japan7) Troisk Institute for Innovating and Fusion Research(TRINITI), Troisk, Russia
Abstract. An ICRF Heating experiment was conducted in the third campaign of the LHD in1999. 1.35 MW of ICRF power was injected into the plasma and 200kJ of stored energy wasobtained, which was maintained for 5 sec only by ICRF power after the termination of the ECH.The impurity problem was so completely overcome that the pulse length was easily extendedto 68 sec at a power level of 0.7 MW. The utility of a liquid stub tuner in steady state plasmaheating was demonstrated in this shot. The energy confinement time of the ICRF heated plasmahas the same dependences on plasma parameters as the ISS95 stellarator scaling with a multi-plication factor of 1.5, which is a high efficiency comparable to NBI. Such an improvement inperformance was obtained by applying various measures, including 1)scanning of the magneticfield intensity and minority concentration, 2)improvement of particle orbit due to a shift of themagnetic axis, and 3) reduction of impurities by means of Ti-gettering and the use of carbondivertor plates. In the optimized heating regime, ion heating turned out to be the dominantheating mechanism, different from that of in CHS and W7-AS. Due to the high quality of theheating and the extended parameter range far beyond that of previous experiments, the exper-iment can be regarded as the first complete demonstration of ICRF heating in stellarators.
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110
(EX8/5) Non-Inductive Current Generation in NSTX Us-ing Coaxial Helicity Injection
R. Raman1), T. R. Jarboe1), D. Mueller2), M. Schaffer3), R. Maqueda4), B. A. Nelson1),S. A. Sabbagh5), M. Bell2), R. Ewig1), E. Fredrickson2), D. Gates2), J. C. Hosea2), S. C. Jardin2),H. Ji2), R. Kaita2), S. Kaye2), H. Kugel2), L. L. Lao3), R. Maingi6), J. Menard2), M. Ono2),D. Orwis1), S. Paul2), M. Peng6), C. Skinner2), J. Wilgen6), S. Zweben2), and the NSTX Re-search Team2)
1) University of Washington, Seattle, WA, USA2) Princeton Plasma Physics Laboratory, Princeton, NJ, USA3) General Atomics, San Diego, CA, USA4) Los Alamos National Laboratory, Los Alamos, NM, USA5) Columbia University, New York, NY, USA6) Oak Ridge National Laboratory, Oak Ridge, TN, USA
Abstract. Coaxial Helicity Injection (CHI) on the National Spherical Torus Experiment (NSTX)has produced 240kA of toroidal current without the use of the central solenoid. Values of thecurrent multiplication ratio (CHI produced toroidal current / injector current) up to 10 wereobtained, in agreement with predictions. The discharges which lasted for up to 200ms, limitedonly by the programmed waveform are more than an order of magnitude longer in duration thatany CHI discharges previously produced in a Spheromak or a Spherical Torus (ST).
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111
(EX8/6) Calculations and Measurements of Rotating Mag-netic Field Current Drive in FRCs
A. L. Hoffman1), R. D. Brooks1), E. Crawford1), H. Y. Guo1), K. E. Miller1), R. D. Milroy1),J. T. Slough1), S. Tobin2)
1) Redmond Plasma Physics Laboratory, University of Washington, Seattle, Washington, USA2) Los Alamos National Laboratory, Los Alamos, New Mexico, USA
Abstract. Rotating Magnetic Fields (RMF) have been demonstrated to drive currents in manyrotamak experiments, but use with an FRC confined in a flux conserver imposes special con-straints. The strong current drive force results in a near zero density at the separatrix, andthe high average beta condition requires the current to be carried in an edge layer near theseparatrix. The RMF can only penetrate into this layer by driving the azimuthal electron ve-locity synchronous with the RMF frequency. Build-up or maintenance of the flux throughoutthe FRC occurs due to the torque imposed on the electrons in this layer exceeding the totalresistive torque due to electron-ion friction. Current is maintained on the inner flux surfaces byan inward radial flow. Particle balance is maintained by a swirling axial flow from inner to outerfield lines. This process is seen using a new numerical code, and the resultant flux build-up andcalculated profiles are demonstrated on the STX and TCS RMF FRC formation experiments.
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112
(EX8/D) Discussion of Section EX8
The file contains the discussion contributions relating to EX8/1, EX8/2, EX8/3, EX8/4, EX8/5.
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113
Session EX9/TH6 — EnergeticParticles
Contents
(EX9/1) Study of Energetic Ion Transport in the Large Helical Device . . . 115
(EX9/1) Study of Energetic Ion Transport in the Large He-lical Device
M. Sasao1), S. Murakami1), M. Isobe1), A. V. Krasilnikov2), S. Iiduka3), K. Itoh1), N. Nakajima1),M. Osakabe1), K. Saito4), T. Seki1), Y. Takeiri1), T. Watari1), H. Yamada1), N. Ashikawa5),P. de Vries1), M. Emoto1), H. Funaba1), M. Goto1), K. Ida1), H. Idei1), K. Ikeda1), S. Inagaki1),N. Inoue1), S. Kado1), O. Kaneko1), K. Kawahata1), K. Khlopenkov1), T. Kobuchi5), A. Ko-mori1), S. Kubo1), R. Kumazawa1), S. Masuzaki1), T. Minami1), J. Miyazawa1), T. Morisaki1),S. Morita1), S. Muto1), T. Mutoh1), Y. Nagayama1), Y. Nakamura1), H. Nakanishi1), K. Nar-ihara1), K. Nishimura1), N. Noda1), T. Notake4, ) Y. Liang5), S. Ohdachi1), N. Ohyabu1),Y. Oka1), T. Ozaki1), B. J. Peterson1), R. O. Pavlichenko1), A. Sagara1), S. Sakakibara1),R. Sakamoto1), H. Sasao5), K. Sato1), M. Sato1), T. Shimozuma1), M. Shoji1), H. Suzuki1),M. Takechi1), N. Tamura5), K. Tanaka1), K. Toi1), T. Tokuzawa1), Y. Torii4), K. Tsumori1), ,I. Yamada1), S. Yamaguchi1), S. Yamamoto4), M. Yokoyama1), Y. Yoshimura1), K. Y. Watan-abe1), O. Motojima1), and M. Fujiwara1)
1) National Institute for Fusion Science, Toki, 509-5292, Japan2) Troitsk Institute for Innovating and Fusion Research, Troitsk, Russia3) Department of Nuclear Engineering, Nagoya University, 464-8603, Japan4) Department of Energy Engineering and Science, Nagoya University, 464-8603, Japan5) Department of Fusion Science, School of Mathematical and Physical Science, Graduate Uni-versity for Advanced Studies, Hayama, 240-0193, Japan
Abstract. The confinement property of high energy ions and the role of ripple induced transporthave been studied in the Large Helical Device (LHD). Tangential beam particles are injectedat 90 - 150 kV by negative-ion-based neutral beam injectors, and perpendicular high energyparticles are generated by Ion Cyclotron Range of Frequency (ICRF) heating. Energy distribu-tions of high energy ions have been measured by fast neutral particle analyzers based on naturaldiamond detectors, specially developed for this experiment. Time evolution of perpendiculartail temperature in decaying phase after the ICRF termination indicates that the fast particlesdeeply trapped in helical ripples are well confined longer than the collisional relaxation time,when ne > 1019m−3 in the inward shifted configuration (Rax = 3.6m). In lower density plasma,however, faster decay than the classical prediction is observed. The tangential energy spectrain a low density steady state plasma of the standard configuration (Rax = 3.75m) also show thedeviation from 2D Fokker-Plank simulation. The global 5-D transport simulation code includingthe ripple induced transport shows good agreement with measured spectra, and this suggests animportant role of ripple induced transport in the standard configuration of LHD.
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115
(EX9/2) Runaway Current Termination in JT-60U
H. Tamai1), R. Yoshino1), S. Tokuda1), G. Kurita1), Y. Neyatani1), M. Bakhtiari2), R. Khayrut-dinov3), V. Lukash4), M. N. Rosenbluth5), and the JT-60U Team1)
1) Japan Atomic Energy Research Institute, Naka-machi, Ibaraki-ken, 311-0193, Japan2) Utsunomiya University, Utsunomiya, Tochigi, 321-8585, Japan3) Troitsk Institute, 142092 Moscow, Russia4) RRC Kurchatov Institute, 123182, Moscow, Russia5) General Atomics, San Diego, California 92186-5698, USA
Abstract. Spontaneous termination of runaway current, generated at the plasma disruption,is found in JT-60U during the vertical plasma displacement event, where the safety factor at theplasma surface, qs decreases. For all shots with runaway electron generation, runaway currenttermination starts with the appearance of a spike in magnetic fluctuation and finishes at qs ≥ 2.Growth rates of the spikes in the magnetic fluctuations decrease by an order of magnitude duringthe termination of runaway current. When each magnetic fluctuation with a slow growth rateappears, runaway current decays, and heat flux pulses are generated. Halo current during therunaway termination is small. Halo current is generated after the runaway termination, andreaches the maximum level at qs ∼ 1.
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116
(EX9/3) Demonstration of Ripple Reduction by FerriticSteel Board Insertion in JFT-2M
H. Kawashima1), M. Sato1), K. Tsuzuki1), Y. Miura1), N. Isei1), H. Kimura1), T. Nakayama2),M. Abe2), D. Darrow3) and the JFT-2M Group1)
1) Japan Atomic Energy Research Institute, Tokai, Naka, Ibaraki Japan2) Hitachi Ltd., Hitachi, Ibaraki Japan3) Princeton Plasma Physics Laboratory, Princeton, New Jersey U.S.A
Abstract. In the JFT-2M tokamak, application testing of low activation ferritic steel to plasmahas been investigated, (so called Advanced Material Tokamak Experiment (AMTEX) program).In the first stage, toroidal field ripple reduction was examined by ferritic steel boards (FBs) in-sertion between toroidal field coils and vacuum vessel. It is demonstrated that the FB insertionis effective to reduce the toroidal field ripple and to reduce the losses of fast ions produced bytangential co-NBI. By optimizing the FB thickness, such that the fundamental mode ripple isminimized to be 0.07% at the shoulder part, the ripple-trapped loss is reduced to be almostnegligible. It is indicated that the reductions of the fundamental mode ripple and the ripplebanana diffusion coefficient at the shoulder part are most effective to reduce the ripple ion losses.Ripple loss reduction by FBs is also confirmed with the perpendicular beam injection. The FBinsertion gives no deteriorative effect on the plasma production and control.
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117
(EX9/4) Transition from Ion Root to Electron Root in NBIHeated Plasmas in LHD
K. Ida1), H. Funaba1), S. Kado2), K. Narihara1), K. Tanaka1), Y. Takeiri1), Y. Nakamura1),N. Ohyabu1), K. Yamazaki1), M. Yokoyama1), S. Murakami1), N. Ashikawa1), P. de Vries1),M. Emoto1), M. Goto1), H. Idei1), K. Ikeda1), S. Inagaki1), N. Inoue1), M. Isobe1), K. Itoh1),O. Kaneko1), K. Kawahata1), K. Khlopenkov1), A. Komori1), S. Kubo1), R. Kumazawa1),Y. Liang3), S. Masuzaki1), T. Minami1), J. Miyazawa1), T. Morisaki1), S. Morita1), T. Mu-toh1), S. Muto1), Y. Nagayama1), H. Nakanishi1), K. Nishimura1), N. Noda1), T. Notake4),T. Kobuchi1), S. Ohdachi1), K. Ohkubo1), Y. Oka1), M. Osakabe1), T. Ozaki1), R. O. Pavlichenko1),B. J. Peterson1), A. Sagara1), K. Saito4), S. Sakakibara1), R. Sakamoto1), H. Sanuki1), H. Sasao1),M. Sasao1), K. Sato1), M. Sato1), T. Seki1), T. Shimozuma1), M. Shoji1), H. Suzuki1), S. Sudo1),N. Tamura3), K. Toi1), T. Tokuzawa1), Y. Torii4), K. Tsumori1), T. Yamamoto1), H. Yamada1),I. Yamada1), S. Yamaguchi1), S. Yamamoto4), Y. Yoshimura1), K. Y. Watanabe1), T. Watari1),Y. Hamada1), O. Motojima1), M. Fujiwara1)
1) National Institute for Fusion Science, 322-6, Oroshi-cho, Toki, 509-5292, Japan2) present address: High temperature Plasma Center, Univ. of Tokyo, Tokyo, 113-8656,Japan3) Department of Fusion Science, School of Mathematical and Physical Science, Graduate Uni-versity for Advanced Studies, Hayama, 240-0193, Japan4) Department of Energy Engineering and Science, Nagoya University, 464-8603, Japan
Abstract. Recent Large Helical Device (LHD) experiments revealed that the transition fromion root to electron root occurred for the first in neutral beam heated discharges, where thereis no non-thermal electrons exist. The measured values of the radial electric field were found tobe in qualitative agreement with those estimated by neoclassical theory. For the configurationwith a magnetic axis of 3.75m, where the ion transport loss was comparable to the neoclassicalion loss, a clear reduction of ion thermal diffusivity was observed after the mode transition fromion root to electron root as predicted by neoclassical theory. On the other hand, for the inwardshifted configuration (Rax = 3.6m), where the neoclassical ion loss is reduced below the anoma-lous loss, no change in the ion thermal diffusivity was observed.
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118
(TH6/1) Fast Particle Effects on the Internal Kink, Fish-bone and Alfven Modes
N. N. Gorelenkov1), S. Bernabei1), C. Z. Cheng1), G. Y. Fu1), K. Hill1), S. Kaye1), G. J. Kramer1),Y. Kusama2), K. Shinokhara2), R. Nazikian1), T. Ozeki2), W. Park1)
1) Prineceton Plasma Physics Laboratory, P.O.Box 451, Princeton, NJ, USA, 08543-04512) Naka Fusion Research Establishment, Japan Atomic Energy Research Institute, Naka-machi,Naka-gun, Ibaraki-ken, 311-0193, Japan.
Abstract. The issues of linear stability of low frequency perturbative and nonperturbativemodes in advanced tokamak regimes are addressed based on recent developments in theory,computational methods, and progress in experiments. Perturbative codes NOVA and ORBITare used to calculate the effects of TAEs on fast particle population in spherical tokamak NSTX.Nonperturbative analysis of chirping frequency modes in experiments on TFTR and JT-60U ispresented using the kinetic code HINST, which identified such modes as a separate branch ofAlfven modes - resonance TAE (R-TAE). Internal kink mode stability in the presence of fastparticles is studied using the NOVA code and hybrid kinetic-MHD nonlinear code M3D.
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119
(TH6/2) Fokker-Planck Simulation Study of Alfven Eigen-mode Burst
Y. Todo1), T. Watanabe1), Hyoung-Bin Park1), T. Sato1)
1) National Institute for Fusion Science
Abstract. Recurrent bursts of toroidicity-induced Alfven eigenmodes (TAEs) are reproducedwith a Fokker-Planck-magnetohydrodynamic simulation where a fast-ion source and slowingdown are incorporated self-consistently. The bursts take place at regular time intervals and thebehaviors of all the TAEs are synchronized. The fast-ion transport due to TAE activity spatiallybroadens the classical fast-ion distribution and significantly reduces its peak value. Only a smallchange of the distribution takes place with each burst, leading to loss of a small fraction of thefast ions. The system stays close to the marginal stability state established through the interplayof the fast-ion source, slowing down, and TAE activity.
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120
(EX9/D) Discussion of Section EX9/TH6
The file contains the discussion contributions relating to EX9/1, EX9/2, EX9/3, TH6/1, TH6/2,EX9/4.
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121
Session EXP1 — PhysicsIntegration
Contents
(EXP1/01) Low Loop Voltage Start-Up in HT-7 Tokamak with Ion CycleRange Frequency (ICRF) . . . . . . . . . . . . . . . . . . . . . . . . . . . 123
(EXP1/02) First Results from Operation of the H-1NF Heliac at IncreasedMagnetic Field and Power . . . . . . . . . . . . . . . . . . . . . . . . . . . 124
(EXP1/01) Low Loop Voltage Start-Up in HT-7 Tokamakwith Ion Cycle Range Frequency (ICRF)
J. R. Luo1), HT-7 Team
1) Institute of Plasma Physics, Academia Sinica,P.R.China
Abstract. The low voltage startup plasma has been studied in the HT-7 (R = 1.22m, a = 0.27m)tokamak with Ion Cyclotron Range of Frequency (ICRF) wave and Lower Hybrid Wave (LHW)for pre-ionization. Successful low loop voltage startup experiment results had been obtained withICRF (about 150kW) and LHW (about 200kW), for example, E ≈ 0.5V/m (with), E ≈ 2.5V/m(without). The ICRF and LHW assisted startup experiments on HT-7 have been successful inshowing the positive effect on plasma pre-ionization and target plasma formation; reduction ofthe volt-second consumption during the period of the breakdown and the ramp-up; reductionof the loop voltage in the breakdown; and setting of a beneficial ramp-up rate for the wholesuperconducting tokamak.
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123
(EXP1/02) First Results from Operation of the H-1NF He-liac at Increased Magnetic Field and Power
J. H. Harris1), B. D. Blackwell1), J. Howard1), M. G. Shats1), C. Charles1), S. M. Collis1),F. J. Glass1), A. Gough1), C. A. Michael1), H. Punzmann1), W. M. Solomon1), G. B. Warr1),G. G. Borg1)
1) Australian National University, Canberra, Australia
Abstract. The H-1 heliac is being developed as a national research facility. A new powersupply system which allows operation to fields ≤ 1 T has been commissioned; the ripple in thesupply is controlled to < 0.01% to eliminate induced currents. Up to 100 kW of rf power at 7MHz is used at present to produce plasmas using helicon waves. Multiple diagnostic studies ofthese plasmas and comparison experiments with a linear helicon device suggest that the nearfields of the rf antennas result in ion temperatures that increase at the edge of the plasma,even though the antennas are 3-4 cm outside the last closed flux surface. Probe and spectro-scopic results indicate that there the mass flow velocities are much less than the ExB velocity.Thus, radial force balance holds in detail, and the ambipolar radial electric field balances theion pressure gradient. In L-mode plasmas, tomographic interferometry and probe studies showlow-mode-number coherent oscillations in electron density and electron and ion temperaturesthat are suppressed at the L-H transition.
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124
(EXP1/03) Plasma Formation and First OH Experimentsin GLOBUS-M Tokamak
V. K. Gusev1), S. V. Aleksandrov1), T. A. Burtseva2), I. N. Chugunov1), A. V. Dech1), G. A. Gavrilov1),V. E. Golant1), Yu. A. Kostsov2), S. V. Krikunov1), E. A. Kuznetsov4), R. G. Levin1), V. B. Mi-naev1), A. Mineev2), O. A. Minyaev3), E. E. Mukhin1), A. N. Novokhatskii1), Yu. V. Petrov1),K. A. Podushnikova1), E. Rumyantsev2), N. V. Sakharov1), V. V. Semenov1), V. M. Sharapov5),G. Yu. Sotnikova1), V. S. Uzlov1), V. Vasiliev2), M. I. Vildjunas1), V. A. Yagnov4)
1) A.F. Ioffe Physico-Technical Institute, Russian Academy of Science, St. Petersburg, Rus-sia2) D.V. Efremov Institute of Electrophysical Apparatus, St. Petersburg, Russia3) Ioffe Fusion Technology, Ltd., St. Petersburg, Russia4) TRINITI, Troitsk, Moscow Region, Russia5) Institute of Physical Chemistry, Moscow, Russia
Abstract. The paper reports results of experimental campaigns on plasma ohmic heating,performed during 1999–2000 on the spherical tokamak Globus-M. Later experimental resultswith tokamak fed by thyristor rectifiers are presented in detail. The toroidal magnetic field andplasma pulse duration in these experiments were significantly increased. The method of straymagnetic field compensation is described. The technology of vacuum vessel conditioning, includ-ing boronization of the vessel performed at the end of the experiments, is briefly discussed. Alsodiscussed is the influence of ECR preioniziation on the breakdown conditions. Experimentaldata on plasma column formation and current ramp-up in different regimes of operation withthe magnetic flux of the central solenoid (CS) limited to ∼ 100 mVs are presented. Ramp-up ofthe plasma current of 0.25 MA for the time interval ∼ 0.03 s with about 0.02 s flat-top at thetoroidal field (TF) strength of 0.35 T allows the conclusion that power supplies, control systemand wall conditioning work well. The same conclusion can be drawn from observation of plasmadensity behavior the density is completely controlled with external gas puff and the influence ofthe wall is negligible after boronization. The magnetic flux consumption efficiency is discussed.The results of magnetic equilibrium simulations are presented and compared with experiment.
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125
(EXP1/04) High Performance H-mode Plasmas at Densi-ties above the Greenwald Limit
M. A. Mahdavi1), T. H. Osborne1), A. W. Leonard1), M. S. Chu1), E. J. Doyle2), M. E. Fen-stermacher3), G. R. McKee4), G. M. Staebler1), T. W. Petrie1), M. R. Wade5), S. L. Allen3),J. A. Boedo6), N. H. Brooks1), R. J. Colchin5), T. E. Evans1), C. M. Greenfield1), G. D. Porter3),R. C. Isler5), R. J. La Haye1), C. J. Lasnier3), R. Maingi5), R. A. Moyer6), M. Schaffer1),P. G. Stangeby7), J. G. Watkins8), W. P. West, D. G. Whyte6), and N. S. Wolf3)
1) General Atomics, San Diego, California USA2) University of California-Los Angeles, Los Angeles, California USA3) Lawrence Livermore National Laboratory, Livermore, California USA4) University of Wisconsin-Madison, Madison, Wisconsin USA5) Oak Ridge National Laboratory, Oak Ridge, Tennessee USA6) University of California-San Diego, La Jolla, California USA7) University of Toronto Institute for Aerospace Studies, Toronto, Canada8) Sandia National Laboratories, Albuquerque, New Mexico USA
Abstract. Densities up to 40 percent above the Greenwald limit are reproducibly achieved inhigh confinement (HITER89p = 2) ELMing H-mode discharges. Simultaneous gas fueling anddivertor pumping were used to obtain these results. Confinement of these discharges, similar tomoderate density H-mode, is characterized by a stiff temperature profile, and therefore sensitiveto the density profile. A particle transport model is presented that explains the roles of divertorpumping and geometry for access to high densities. Energy loss per ELM at high density isa factor of five lower than predictions of an earlier scaling, based on data from lower densitydischarges.
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126
(EXP1/05) Flux Consumption Optimization and the Achieve-ment of 1 MA Discharges on NSTX
J. Menard1), B. LeBlanc1), S. A. Sabbagh2), M. Bell1), R. Bell1), E. Fredrickson1), D. Gates1),S. C. Jardin1), S. Kaye1), H. Kugel1), R. Maingi3), R. Maqueda4), D. Mueller1), S. Paul1),C. Skinner1), D. Stutman5), and the NSTX Research Team
1) Plasma Physics Laboratory, Princeton University, Princeton, New Jersey, USA2) Department of Applied Physics and Applied Mathematics, Columbia University, New York,New York, USA3) Oak Ridge National Laboratory, Oak Ridge, Tennessee, USA4) Los Alamos National Laboratory, Los Alamos, New Mexico, USA5) Johns Hopkins University, Baltimore, Maryland, USA
Abstract. The spherical tokamak (ST), because of its slender central column, has very limitedvolt-second capability relative to a standard aspect ratio tokamak of similar plasma cross-section.Recent experiments on the National Spherical Torus Experiment (NSTX) have begun to quantifyand optimize the ohmic current drive efficiency in a MA-class ST device. Sustainable ramp-ratesin excess of 5MA/sec during the current rise phase have been achieved on NSTX, while fasterramps generate significant MHD activity. Discharges with IP exceeding 1MA have been achievedin NSTX with nominal parameters: aspect ratio A = 1.3− 1.4, elongation κ = 2− 2.2, triangu-larity δ = 0.4, internal inductance li = 0.6, and Ejima coefficient CE = 0.35. Flux consumptionefficiency results, performance improvements associated with first boronization, and comparisonsto neoclassical resistivity are described.
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127
(EXP1/06) Drift Orbit and Magnetic Surface Measurementsin the Helically Symmetric Experiment
J. N. Talmadge1), V. Sakaguchi1), F. S. B. Anderson1), D. T. Anderson1), A. F. Almagri1)
1) HSX Plasma Laboratory, University of Wisconsin-Madison, Madison WI USA
Abstract. HSX is a toroidal quasihelically-symmetric stellarator with negligibly small toroidalcurvature. Vacuum magnetic surfaces at 1 kG are measured using low-energy electron beamsthat strike a fluorescent mesh. The images are recorded with a CCD camera and show no ob-servable evidence of island structures inside the separatrix. The experimental determination ofthe rotational transform agrees with numerical calculations to within 1%. A simple analyticexpression is derived in Boozer coordinates to relate the drift orbits of passing particles to themagnetic field spectrum. This expression is used to analyze images of high-energy electron orbitsin HSX, using a neural network to map the lab coordinate system into Boozer coordinates. Atvery low magnetic field strengths (90 G) where the b11 component due to the earth s field isnot ignorable, this spectral component and the dominant helical term b41 can be experimentallydetermined. The data does not show the magnitude and direction of the orbit shift that wouldbe expected from the standard toroidal curvature term that exists in other toroidal devices.The results also confirm for the first time that quasihelical stellarators have a large effectivetransform that is responsible for small drifts of particles off a flux surface.
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128
(EXP1/07) Initial Plasmas in the UCLA Electric Tokamak
R. J. Taylor1), J.-L. Gauvreau1), M. Gilmore1), P.-A. Gourdain1), D. J. LaFonteese1), L. W. Schmitz1)
1) University of California at Los Angeles, CA, USA 90095-1597
Abstract. The UCLA Electric Tokamak (ET), a low field ITER sized device, has been oper-ating with well equilibrated clean plasmas since January 2000. The operating scenario is stillevolving as the magnetic configuration and the power supplies undergo refinements. The goalof equilibrating near unity beta plasmas will require 10 second long discharges at 3 kV tem-peratures in a toroidal field of 0.25 Tesla due to current shaping requirements. Short, 0.9 sec,discharges are now routinely obtained with kTe, kTi ∼ 120eV at a toroidal field of 0.1 Tesla. Thedischarges are feedback controlled in up/down position and in plasma current. Biased electrodedriven H-modes have been obtained and compare well to the results obtained on CCT and to the“neoclassical bifurcation” theory. Very successful second harmonic ion heating has been demon-strated with an ICRF antenna outside of the vacuum system and 50% single pass absorption.These discharges also indicate that edge bifurcation can be achieved by RF alone due to fast ionlosses. The remaining critical item needed for the exploration of unity beta plasma stability isthe demonstration of RF current profile shaping near the Troyon limit. We expect that ion-ionhybrid mode conversion (high field side launch) will allow current drive at low beta. This canthen be supplemented by high harmonic current drive at higher beta. Ultimately, near ignitionconditions could be reached if magnetic omnigeneity (classical transport physics) were obtainedat a toroidal field of 1 Tesla. The test of this concept is to be carried out at 0.25 Tesla in thecoming year, if RF current profile shaping can be achieved and supplemented by bootstrap anddiffusion driven current.
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129
(EXP1/08) Initial Results from the TST-2 Spherical Toka-mak
Y. Takase1), A. Ejiri1), N. Kasuya1), T. Mashiko1), S. Shiraiwa1), L. M. Tozawa1), T. Akiduki1),H. Kasahara1), Y. Nagashima1), H. Nozato1), H. Wada1), H. Yamada1), T. Yamada1), K. Yam-agishi1)
1) The University of Tokyo, Tokyo, Japan
Abstract. A new spherical tokamak TST-2 was constructed at the University of Tokyo andstarted operation in September 1999. Reliable plasma initiation is achieved with typically 1 kWof ECH power at 2.45 GHz. Plasma currents of up to 90 kA and toroidal fields of up to 0.2 Thave been achieved during the initial experimental campaign. The ion temperature is typically100 eV. Internal reconnection events (IREs) are often observed. The internal magnetic fieldmeasured at r/a = 2/3 indicated growth of fluctuations up to the 4th harmonic, suggesting theexistence of modes with several different mode numbers. In the presence of a toroidal field anda vertically oriented mirror field, noninductively driven currents of order 1 kA were observedwith 1 kW of ECH power. The driven current increased with decreasing filling pressure, downto 3× 10−6 torr. A study of high harmonic fast wave (HHFW) excitation and propagation hasbegun. Initial results indicate highly efficient wave launching.
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130
(EXP1/09) First Plasmas in Heliotron J
T. Obiki1), T. Mizuuchi1), K. Nagasaki1), H. Okada1), S. Besshou2), F. Sano1), K. Kondo2),Y. Liu1), Y. Nakamura2), K. Hanatani1), M. Nakasuga2), M. Wakatani2), T. Hamada1), Y. Man-abe1), H. Shidara1), O. Yamagishi2), K. Aizawa2), W. L. Ang1), Y. I. Ikeda1), Y. Kawazome2),T. Kobayashi1), S. Maeno2), T. Takamiya1), M. Takeda1), K. Tomiyama2), Y. Ijiri1), T. Senju1),K. Yaguchi1), K. Sakamoto1), K. Toshi1), M. Shibano1)
1) Institute of Advanced Energy, Kyoto University, Gokasho, Uji, Japan.2) Graduate School of Energy Science, Kyoto University, Gokasho, Uji, Japan.
Abstract. Results obtained in the initial experiment phase of Heliotron J are reported. Theelectron beam mapping of the magnetic surfaces has revealed that the observed surfaces arein basic agreement with the calculated ones based on the measured ambient field around thedevice. For 53.2-GHz second harmonic ECH hydrogen plasmas, a fairly wide resonance rangefor breakdown by the TE02 mode has been observed in Heliotron J as compared with that ofHeliotron E. With ECH injection powers up to ∼ 400kW, diamagnetic stored energies up to ∼0.7 kJ were obtained without the optimized density control.
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131
(EXP1/10) Progress in Long Sustainment and High DensityExperiments with Potential Confinement on GAMMA 10
K. Yatsu1), T. Cho1), M. Hirata1), H. Hojo1), M. Ichimura1), K. Ishii1), A. Itakura1), I. Katanuma1),J. Kohagura1), Y. Nakashima1), T. Saito1), T. Tamano1), S. Tanaka1), Y. Tatematsu1), M. Yoshikawa1)
1) Plasma Research Center, University of Tsukuba, Tsukuba, Japan
Abstract. The improvement of potential confinement reported in the last IAEA meeting was at-tained by axisymmetrization of heating pattern of electron cyclotron resonance heating (ECRH).It was experimentally shown that the axisymmetrization of ECRH really produced axisymmet-ric potential profile. GAMMA 10 experiments have advanced in longer sustainment and highdensity operation of potential confinement. Experiments for long sustainment of potential con-finement were carried out in order to study problems of steady state operation of a tandem mirrorreactor. A confining potential was sustained for 150 ms by sequentially injecting two (ECRH)powers in the plug region. It was difficult before to increase the central cell density higher thanabout 2.5× 1012cm−3 with and/or without potential confinement due to some density limitingmechanism. In order to overcome this problem, a new higher frequency ion cyclotron rangeof frequency (ICRF) system (RF3: 36-76 MHz) has been installed. A higher density plasmahas been produced with RF3. In addition to RF3, neutral beam injection (NBI) in the an-chor cell became effective by reducing neutral gas from beam injectors. Potential confinementexperiments have advanced to higher central cell densities up to 4 × 1012cm−3 with RF3 andNBI. A 20% density increase due to the potential confinement was obtained in the high densityexperiments.
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132
(EXP1/11) Real Time Plasma Feedback Control: An Overviewof Tore-Supra Achievements
G. Martin1), J. Bucalossi1), A. Ekedahl1), C. Gil1), C. Grisolia1), D. Guilhem1), J. Gunn1),F. Kazarian1), D. Moulin1), J. Y. Pascal1), F. Saint-Laurent1)
1) Association Euratom-CEA, CEA Cadarache, FRANCE
Abstract. Stable and reliable fusion plasma operation requires increasingly advanced controlsystems. This is especially true for steady-state operation in advanced modes, when severalparameters are to be simultaneously optimised: e.g. the current profile, which has been relatedto the formation of internal transport barrier, and the density, which plays a crucial role bothin the fusion power and in the plasma wall interactions. At a more technological level, goodmanagement of the power entering and leaving the plasma is required, by efficient additionalheating coupling, and with a full control of radiation and convection losses and distribution tothe first wall elements. For these goals, several feed-back mechanisms have been developed withsuccess on Tore-Supra, in the past four years. Most of them are based on software, implementedin a set of micro-computers connected through a VME network.
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133
(EXP1/12) Recent Experiment Progress on the HL-1M Toka-mak
Y. Liu1), E. Y. Wang1), X. T. Ding1), L. W. Yan1), J. F. Dong1), G. C. Guo1), Z. G. Xiao1),Y. J. Zheng1), J. Y. Cao1), Z. C. Deng1), Y. Zhou1), D. M. Xu1), M. L. Shi1), J. Rao1), H. Hua1),G. J. Lei1), L. B. Ran1), J. C. Yan1)
1) Southwestern Institute of Physics, Chengdu, China
Abstract. Experiments on the auxiliary heating, fueling of plasma and wall conditioning werecarried out on HL-1M. ECRH experiments were conducted successfully with Te increase morethan 50%. The double sawtooth in soft X-ray radiation were observed, which imply that thereversed magnetic shear could be formed during ECRH. An eight-shot pellet injector (PI) wasused for experiments. After the pellet injection, a hollow electron temperature and peakeddensity profile were obtained, accompanied with the increase of energy confinement time. Thepellet ablation process was investigated with a CCD camera and an H.. emission detector ar-ray. Obviously asymmetry in the pellet cloud was observed in both the toroidal and poloidaldirection. It is found that the velocity of pellet is slowed down obviously after the pellet entersinto the plasma. The safety-factor q-profile was estimated with the inclination angle of ablationcloud with respect to the torus. Density limit investigations have been performed at differentwall condition with three kinds of fuelling methods. It is found that higher density limit canbe achieved in following conditions: one is the strong reduction of the impurity content aftersiliconization, another is the peaked density profile with pellet injection and/or SMBI. With aNBI system of 1MW, preliminary results of NBI experiments were obtained with increase of iontemperature from 600eV to 800eV.
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134
(EXP1/13) First Results on Dense Plasma Confinement atthe Multimirror Open Trap GOL-3-II
V. S. Koidan1), A. V. Arzhannikov1), V. T. Astrelin1), A. V. Burdakov1), I. A. Ivanov1),S. A. Kuznetsov1), K. I. Mekler1), S. A. Novozhilov1), S. V. Polosatkin1), V. V. Postupesv1),A. F. Rovenskikh1), A. V. Savchkov1), S. L. Sinitsky
1) Budker Institute of Nuclear Physics, Novosibirsk, Russia
Abstract. First results of experiments on plasma confinement in multimirror open trap GOL-3-II are presented. This facility is an open trap with total length of 17 m intended for confinementof a relatively dense (1015 − 1017cm−3) plasma in axially-symmetrical magnetic system. Theplasma heating is provided by a high-power electron beam (1 MeV, 30 kA, 8 ms, 200 kJ). Newphase of the experiments is aimed to confinement of high-β thermalized plasma. Two essentialmodifications of the facility have been done. First, plasma column was separated by vacuumsections from the beam accelerator and exit beam receiver. Second, the magnetic field on partof the solenoid was reconfigured into multimirror system with Hmax/Hmin ∼ 1.5 and 22 cm celllength. Results of the experiments at modified configuration of the device indicate that theconfinement time of the plasma with ne ∼ (0, 5÷ 5) · 1015cm−3 and Te ∼ 1 keV increases morethan order of magnitude.
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135
Session EXP2 — MHD &Energetic Particles
Contents
(EXP2/01) Study on the Behaviour of High Energy Electrons in REPUTE-1ULQ Plasmas . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 137
(EXP2/02) Interdependence of Magnetic Islands, Halo Current and Run-away Electrons in T-10 Tokamak . . . . . . . . . . . . . . . . . . . . . . . 138
(EXP2/03) The Combined Effect of EPM and TAE Modes on EnergeticConfinement and Sawtooth Stabilization . . . . . . . . . . . . . . . . . . 139
(EXP2/04) Direct Measurements of Damping Rates and Stability Limitsfor Low Frequency MHD Modes and Alfven Eigenmodes in the JETTokamak . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 140
(EXP2/05) Alfven Eigenmodes Driven by Energetic Ions in JT-60U . . . . . 141
(EXP2/08) Energetic Particle Physics and MHD Stability in JET and START144
136
(EXP2/01) Study on the Behaviour of High Energy Elec-trons in REPUTE-1 ULQ Plasmas
Y. Ogawa1), H. Nihei2), J. Morikawa1), T. Nakajima2), D. Ozawa2), M. Ohno2), T. Suzuki3),H. Himura4), Z. Yoshida4), S. Morita5), Y. Shirai5)
1) High Temperature Plasma Center, the University of Tokyo, Tokyo, Japan2) School of Engineering, the University of Tokyo, Tokyo, Japan3) Japan Atomic Energy Research Institute, Naka, Ibaraki, Japan4) School of Frontier Science, the University of Tokyo, Tokyo, Japan5) National Institute for Fusion Science, Toki, Gifu, Japan
Abstract. In REPUTE-1 Ultra-Low-q (ULQ) plasmas, behaviors of high energy electrons havebeen studied through a low-Z pellet injection experiment, in addition to the measurements ofsoft-X ray PHA and Electron Energy Analyzer(EEA). The high energy tail has been measuredin the soft-X ray spectrum, and EEA signal has shown a strong anisotropy of the electron distri-bution function (i.e., the electron flux to the electron drift side is dominant). To study temporaland spatial information on these high energy electrons, a low-Z pellet injection experiment hasbeen conducted. A small piece of plastic pellet is injected from the top of the REPUTE-1 device,and the trajectory of the pellet inside the plasma is measured by CCD camera. We have observeda large deflection of the pellet trajectory to the toroidal direction opposite to the plasma current(i.e., the electron drift side). This suggests that a pellet is ablated selectively only from one sidedue to the high energy electrons with a large heat flux. We have calculated the heat flux carriedby high energy electrons. Since the repulsion force to the pellet can be calculated with the 2nd
derivative of the pellet trajectory, we have estimated the heat flux of high energy electrons tobe a few tens MW/m2 around the plasma center. Experimental data by EEA measurement andlow-Z pellet ablation show the large population of the high energy electrons at the core regionin comparison with the edge region, suggesting a MHD dynamo mechanism for the productionof the high energy electrons.
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137
(EXP2/02) Interdependence of Magnetic Islands, Halo Cur-rent and Runaway Electrons in T-10 Tokamak
N. V. Ivanov1), A. M. Kakurin1), V. A. Kochin1), P. E. Kovrov1), I. I. Orlovski1), Yu. D. Pavlov1),V. V. Volkov1)
1) Nuclear Fusion Institute, Russian Research Center “Kurchatov Institute”, Moscow, Russia
Abstract. The results of experiments on a modulation of halo current through a rail lim-iter and of x-ray emission from the limiter under the influence of rotating magnetic islands arepresented. The external part of the halo-current circuit connected the rail limiter at one sideand the discharge chamber with a circular limiter at the other side. A controllable connectorthat was switched on at a preprogrammed moment of time during the tokamak discharge wasintroduced into the circuit. In discharges with MHD activity, oscillations of the halo currentwere observed. The frequency of the oscillations was equal to the frequency of the dominantmode of the poloidal magnetic field perturbation. In some conditions the switching on of theconnector in the halo-current circuit resulted in a shift of the MHD mode frequency. This meansthat the halo current can influence the rotation velocity of the magnetic islands. In the case oflow plasma density, repetitive spikes of hard x-ray emission from the rail limiter were observed.These spikes were coherent with the MHD-activity signal and the halo-current oscillations. Itcan be concluded that besides an effect of magnetic islands resulting in halo-current and x-raymodulation with the frequency of MHD activity, an influence of halo current on the magneticisland behaviour was observed. This influence can be attributed to a coupling between the mag-netic islands and the space-resonant component of the halo-current magnetic field.
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138
(EXP2/03) The Combined Effect of EPM and TAE Modeson Energetic Confinement and Sawtooth Stabilization
S. Bernabei1), R. Budny1), E. Fredrickson1), N. N. Gorelenkov1), R. W. Harvey3), J. C. Hosea1),C. C. Petty2), C. K. Phillips1), R. I. Pinsker2), P. Smirnov4), R. White1), J. R. Wilson1)
1) Princeton Plasma Physics Laboratory, Princeton N.J. 085402) General Atomics, San Diego, California 921863) CompX, P.O. Box 2672, Del Mar, California 920144) Moscow State University, Moscow, Russia.
Abstract. It is shown in this paper for the first time, that the chirping Alfven instabilities ob-served mostly during ICRF heating have been positively identified as Energetic Particle Modes.This has been possible because of the detailed measurement of the q-profile with the MSE diag-nostic in DIII-D. The EPMs are shown to be the leading cause of the monster sawtooth crash.It is also shown that TAEs are excited either directly or indirectly by the EPMs and they causefast ion losses. A scenario for the stabilization and the crash of the monster sawtooth and forthe degradation of the ICRF heating efficiency at high power is presented.
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139
(EXP2/04) Direct Measurements of Damping Rates andStability Limits for Low Frequency MHD Modes and AlfvenEigenmodes in the JET Tokamak
A. F. Fasoli1), D. Testa1), A. Jaun2), S. Sharapov3), C. Gormezano4)
1) Plasma Science and Fusion Center, MIT, Cambridge, USA.2) Alfven Laboratory, Euratom-NFR Association, KTH, Stockholm, Sweden.3) JET Joint Undertaking, Abingdon, United Kingdom.4) Associazione EURATOM-ENEA, Frascati, Italy.
Abstract. The linear stability properties of global modes that can be driven by resonant en-ergetic particles or by the bulk plasma are studied using an external excitation method basedon the JET saddle coil antennas. Low toroidal mode number, stable plasma modes are drivenby the saddle coils and detected by magnetic probes to measure their structure, frequency anddamping rate, both in the Alfven Eigenmode (AE) frequency range and in the low frequencyMagneto-Hydro-Dynamic (MHD) range. For AEs, the dominant damping mechanisms are iden-tified for different plasma conditions of relevance for reactors. Spectra and damping rates of lowfrequency MHD modes that are localized at the foot of the internal transport barrier and canaffect the plasma performance in advanced tokamak scenarios have been directly measured forthe first time. This gives the possibility of monitoring in real time the approach to the instabilityboundary.
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140
(EXP2/05) Alfven Eigenmodes Driven by Energetic Ions inJT-60U
K. Shinohara1), Y. Kusama1), G. J. Kramer2), M. Takechi1), A. Morioka1), M. Ishikawa1),N. Oyama1), K. Tobita1), T. Ozeki1), S. Takeji1), S. Moriyama1), T. Fujita1), T. Oikawa1),T. Suzuki1), T. Nishitani1), T. Kondoh1), S. Lee1), M. Kuriyama1), N. N. Gorelenkov2), R. Nazikian2),C. Z. Cheng2), G. Y. Fu2), A. Fukuyama3) and the JT-60U Team1)
1) Japan Atomic Energy Research Institute, Naka-machi, Naka-gun, Ibaraki 311-0193, Japan2) Princeton Plasma Physics Laboratory, P.O. Box 451, Princeton, NJ 08543, USA3) Department of Nuclear Engineering, Kyoto University, 606-8501 Japan
Abstract. Instabilities with frequency chirping in the frequency range of Alfven eigenmodeshave been found in the domain of 0.1% < 〈βh〉 < 1% and vb‖/vA ∼ 1 with high energy neutralbeam injection in JT-60U. One instability appears with frequency inside the Alfven continuumspectrum and its frequency increases slowly to the Toroidicity induced Alfven eigenmodes (TAE)gap in an equilibrium change time scale of ∼ 200 ms. Another instability appears with frequencyinside the TAE gap and its frequency changes very fast by 10–20 kHz in 1–5 ms. During theoccurrence of Fast FS modes, abrupt large-amplitude events often appears with an enhanceddrop of neutron emission rate and increase in fast neutral particle fluxes. The loss of energeticions increases with the peak fluctuation amplitude of Bθ/Bθ. Energy dependence of loss ions isobserved and suggests the resonant interaction between energetic ions and a mode.
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141
(EXP2/06) Relativistic Runaway Electrons in TEXTOR-94
N. J. Lopes Cardozo1), I. Entrop1), R. Jaspers1), K. H. Finken2), H. L. M. Widdershoven1)2)
1) FOM Institute Rijnhuizen, Nieuwegein, The Netherlands2) Institut fur Plasmaphysik, Forschungszentrum Julich GmbH, Julich, Germany
Abstract. This paper reviews results concerning generation, confinement and transport of run-away electrons in the energy range 20-30 MeV in the TEXTOR tokamak. Runaway electronsabove 20 MeV emit synchrotron radiation in the (near) infrared wavelength range, which caneasily be detected by thermographic cameras. This technique is developed and exploited at theTEXTOR-94 tokamak and has resulted in some spectacular results. These include: the experi-mental evidence of the secondary (‘knock-on’) runaway generation; the discovery of the runawaysnake; the observation of disruption generated runaways; the probing of magnetic turbulence inthe core of the plasma in Ohmic and additionally heated plasmas.
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142
(EXP2/07) Dynamics of Runaways in JET
R. D. Gill, B. Alper, A. W. Edwards, L. C. Ingesson, M. F. Johnson, D. Ward
EURATOM/UKAEA Fusion Association, Culham Science Centre, Abingdon, Oxon, OX14 3EA,UK.
Abstract. Measurements are presented of the properties of the runaway beams generated inJET following disruptions. Radiation is emitted by the runaways, both when they are in flightand when they hit the vessel walls. Because radiation protected soft x-ray cameras were devel-oped for the JET DT campaign, it has been possible to make the first direct observations of therunaway beam in flight from the x-ray line radiation produced by the beam excitation of K-shellvacancies in the metallic impurities of the residual plasma. These observations give clear imagesof the runaway beam and provide detailed information on its time development, size, positionand stability. The current density and q-profile have also been determined. It has been foundthat there is a delay between the disruption and the start of runaway generation and this offers apossibility of instigating runaway control methods. Detailed determination of the runaway-wallinteraction suggests that the runaways have a braided structure.
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143
(EXP2/08) Energetic Particle Physics and MHD Stabilityin JET and START
K. G. McClements1), R. J. Akers1), L. C. Appel1), R. O. Dendy1), A. Gondhalekar1), A. A. Ko-rotkov1), M. R. O’Brien1), S. Sharapov1), R. J. Hastie2), J. P. Graves3), K. I. Hopcraft3),D. N. Borba4), M. F. F. Nave4), M. Mantsinen5), N. N. Gorelenkov6), M. V. Gorelenkova7),F. S. Zaitsev8)8)
1) EURATOM/UKAEA Fusion Association, Culham Science Centre, Abingdon, UK2) MIT, Plasma Science and Fusion Center, Cambridge, Massachusetts, USA3) School of Mathematical Sciences, Nottingham University, Nottingham, UK4) Associacao EURATOM/IST, Centro de Fusao Nuclear, Lisbon, Portugal5) Association EURATOM-TEKES, Helsinki University of Technology, Espoo, Finland6) Princeton Plasma Physics Laboratory, Princeton University, Princeton, USA7) TRINITI, Troitsk, Moscow Region, Russia8) Moscow State University, Moscow, Russia
Abstract. Data from sawtoothing JET discharges with ICRH has revealed a correlation be-tween the heated ion contribution to the m = 1 kink energy and the Shafranov shift gradient atq = 1, consistent with theory. A correlation has also been found between the total energetic par-ticle kink energy and sawtooth period in a sequence of JET pulses with different beam tritiumconcentrations. Neutral particle analyzer (NPA) measurements of alpha particles in JET, whichreveal MeV deuterons resulting from nuclear elastic scattering, can in principle yield informationon alpha particle anisotropy. The spherical tokamak geometry in START has provided a rigor-ous test of toroidal Alfven eigenmode (TAE) theory: independent codes give consistent resultsfor TAE mode frequencies and structure. A mechanism capable of explaining the excitation ofchirping fishbones by circulating beam ions in START has also been identified.
(EXP3/17) Progress in the Prediction of Disruptions in ASDEX-Upgradevia Neural and Fuzzy-Neural Techniques . . . . . . . . . . . . . . . . . . 162
145
(EXP3/01(R)) Resistive Wall Mode Dynamics and ActiveFeedback Control in DIII-D
A. M. Garofalo1), J. Bialek1), A. H. Boozer1), M. S. Chu2), E. Fredrickson3), M. Gryaznevich3),T. H. Jensen2), L. C. Johnson4), R. J. La Haye2), G. A. Navratil1), M. Okabayashi4), E. J. Strait2),J. T. Scoville2), A. D. Turnbull2), and the DIII-D Team
1) Columbia University, New York, New York, USA2) General Atomics, San Diego, California USA3) Princeton Plasma Physics Laboratory, Princeton, New Jersey, USA4) UKAEA-Culham Laboratory, Abingdon, United Kingdom
Abstract. Recent DIII-D experiments have shown that the n=1 resistive wall mode (RWM) canbe controlled by an external magnetic field applied in closed loop feedback using the six elementerror field correction coil (C-coil). The RWM constitutes the primary limitation to normalizedbeta in recent DIII-D advanced tokamak plasma experiments. The toroidal rotation of DIII-Dplasmas does not seem sufficient to completely suppress the RWM: a very slowly growing RWM(growth rate γ 1/τw) is often observed at normalized beta above the no-wall limit and thissmall RWM slows the rotation. As the rotation decreases, there is a transition to more rapidgrowth (γ ∼ 1/τw). The application of magnetic feedback is able to hold the RWM to a verysmall amplitude, prolonging the plasma duration above the no-wall limit for durations muchlonger than the growth time of the RWM. These initial experimental results are being used tocompare control algorithms, to benchmark models of the feedback stabilization process and toguide the design of an upgraded coil-sensor system for stabilization of the RWM at normalizedbeta values closer to the ideal-wall limit.
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146
(EXP3/02) Neoclassical Tearing Mode Studies in JET
T. C. Hender1), D. N. Borba2), R. J. Buttery1), D. F. Howell1), G. T. A. Huysmans3), R. J. LaHaye4), P. J. Lomas1), A. Martynov5), M. Mantsinen6), S. D. Pinches7), O. Sauter5), M. Za-biego3)
1) Euratom/UKAEA Fusion Association, Culham Science Centre, Abingdon, UK.2) EFDA-JET Close Support Unit, Culham Science Centre, Abingdon, UK.3) Association Euratom/CEA, CEA CAdarache, Saint-Paul-Lez-Durance, France4) General Atomics, San Diego, California, USA5) CRPP Euratom-Confederation Suisse, Ecole Polytechnique Federal de Lausanne, Switzerland6) Association Euratom-Tekes, Helsinki University of Technology, Finland7) Association Euratom/IPP, Max Planck Institute, Garching, Germany
Abstract. Studies of (3,2) and (2,1) Neo-classical Tearing Modes (NTMs) in JET are presented.The effects of plasma shape and edge safety factor (q95) on the β-limits set by NTMs are ad-dressed, as are results on the effect of ICRF heating. For the (2,1) NTMs initial comparisonswith the β-limit thresholds from the DIII-D tokamak are presented.
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147
(EXP3/03) Long Sustainment of Quasi-steady State HighBeta-p H-mode Discharges in JT-60U
A. Isayama1), Y. Kamada1), T. Ozeki1), S. Ide1), T. Fujita1), T. Oikawa1), T. Suzuki1), Y. Ney-atani1), N. Isei1), K. Hamamatsu1), Y. I. Ikeda1), K. Takahashi1), K. Kajiwara1) and the JT-60UTeam1)
1) Japan Atomic Energy Research Institute, Naka Fusion Research Establishment
Abstract. Quasi-steady state high βp H-mode discharges performed by suppressing neoclas-sical tearing modes (NTM) is described. Two operational scenarios have been developed forlong sustainment of the high βp H-mode discharge: NTM suppression by profile optimizationand NTM stabilization by local electron cyclotron current drive (ECCD) / electron cyclotronheating (ECH) at the magnetic island. By optimizing pressure and safety factor profiles, a highβp H-mode plasma with H89PL = 2.8, HHy2 = 1.4, βp = 2.0, βN = 2.5 has been sustained for 1.3s at small values of νe∗ and ρi∗ without destabilizing the NTMs. Characteristics of the NTMsdestabilized in the region of the central safety factor above unity are investigated. Relationbetween the beta value at the mode onset βon
N and that at the mode disappearance βoffN can be
described as βoffN = 0.05− 0.4βon
N , which shows the existence of hysteresis. The value of βN/ρi∗at the onset of an m/n=3/2 NTM has collisionality dependence, which is empirically given byβN/ρi∗ ∝ ν0.36
e∗ . However, the profile effects are equally important, such as relative shapes ofpressure and safety factor. Onset condition seems to be affected by strength of the pressuregradient at the mode rational surface. Stabilization of the NTM by local ECCD/ECH at themagnetic island has been attempted. The 3/2 NTM has been completely stabilized by EC waveinjection of 1.6 MW.
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148
(EXP3/04) Beta Limit Due to Resistive Instabilities in T-10
D. A. Kislov1), Yu. V. Esipchuk1), N. Kirneva1), I. V. Klimanov1), Yu. D. Pavlov1), A. A. Sub-botin1), V. V. Alikaev1), A. A. Borshegovskij1), Yu. V. Gott1), A. M. Kakurin1), S. V. Krilov1),T. Myalton1), I. N. Roy1), E. Trukhina1), V. V. Volkov1) and the T-10 Team
1) RRC “Kurchatov Institute”, Moscow, Russia
Abstract. Soft beta limiting phenomena have been observed in T-10 in ECRH heated plas-mas. Neoclassical tearing modes are supposed to be responsible for the beta limitation. TheMHD onset was observed at high βp but low βN values. Critical β has been found to be almostindependent of collisionality parameter ν∗e . Sawtooth stabilization by ECCD does not result inan increase of critical β . Dependence of the critical β on the q(r) profile (modified by ECCD)has been observed.
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149
(EXP3/05) Polarization Current and Neoclassical TearingMode Threshold in Tokamaks: Comparison of Experimentwith Theory
R. J. La Haye1), R. J. Buttery2), S. Gunter3), G. T. A. Huysmans4), M. Maraschek3), F. Wael-broeck5), and H. R. Wilson2)
1) General Atomics, San Diego, California USA2) Euratom/UKAEA Fusion Association, Culham Science Center, Abingdon, United Kingdom3) Max Planck Institute fur Plasmaphysik, Garching, Federal Republic of Germany4) JET Joint Undertaking, Abingdon, United Kingdom (now at CEA, Cadarache, France)5) Institute for Fusion Studies, Austin, Texas USA
Abstract. Neoclassical tearing mode islands are one of the main causes of reduced performanceat high βθ in standard ELMy sawtoothing H-mode. The leading candidate for the threshold isthe helical polarization/inertial current which arises from mode propagation at frequency ω inthe Er = 0 guiding center frame of plasma flow. A threshold island width wpol is predicted,which is proportional to the ion banana width ε1/2ρθi and also depends on ω. The polarizationcurrent is predicted to be stabilizing only for 0 ≤ ω ≤ ωi∗, the ion diamagnetic drift frequency,and yields a minimum βθ (below which the helically perturbed bootstrap current is too smallto excite NTMs) that gives critical βN scaling linearly with ρi∗. A database compiled from thetokamaks ASDEX Upgrade (AUG), DIII-D and JET shows such a βNcrit ∝ ρi∗ is indeed observedfor the m/n=3/2 NTM induced by a sawtooth crash. Typically, unstable seed island widths thatgrow are observed to be of the order wpol. Detailed measurements of mode propagation in theEr = 0 frame are also consistent with a polarization current threshold.
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150
(EXP3/06) Dependence of Edge Stability on Plasma Shapeand Local Pressure Gradients in the DIII-D and JT-60UTokamaks
L. L. Lao1), Y. Kamada2), T. Oikawa2), L. R. Baylor3), K. H. Burrell1), V. S. Chan1), M. S. Chance4),M. S. Chu1), J. R. Ferron1), T. Fukuda2), T. Hatae2), A. Isayama2), G. L. Jackson1), A. W. Leonard1),M. A. Makowski5), J. Manickam4), M. Murakami3), M. Okabayashi4), T. H. Osborne1), P. B. Sny-der1), E. J. Strait1), S. Takeji2), T. Takizuka2), T. S. Taylor1), A. D. Turnbull1), K. Tsuchiya2),and M. R. Wade3)
1) General Atomics, San Diego, California USA2) Japan Atomic Energy Research Institute, Naka-machi, Naka-gun, Ibaraki-ken, Japan3) Oak Ridge National Laboratory, Oak Ridge, Tennessee USA4) Princeton Plasma Physics Laboratory, Princeton, New Jersey USA5) Lawrence Livermore National Laboratory, Livermore, California USA
Abstract. The dependence of edge stability on plasma shape and local pressure gradients,P’, in the DIII-D and JT-60U tokamaks is studied. The stronger plasma shaping in DIII-Dallows the edge region of DIII-D discharges with Type I (“giant”) ELMs to have access to thesecond region of stability for ideal ballooning modes and larger edge P’ than JT-60U Type IELM discharges. These JT-60U discharges are near the ballooning mode first regime stabilitylimit. DIII-D results support an ideal stability based working model of Type I ELMs as low tointermediate toroidal mode number, n, MHD modes. Results from stability analysis of JT-60UType I ELM discharges indicate that predictions from this model are also consistent with JT-60U edge stability observations.
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151
(EXP3/07) Control of a Pressure Driven m=1 Mode in aLow Shear LHCD Plasma by ECH in the WT-3 Tokamak
T. Maekawa1), Y. Terumichi2), S. Yoshimura3), M. Asakawa2), H. Tanaka1)
1) Graduate School of Energy Science, Kyoto University, Kyoto 606-8502, Japan2) Department of Physics, Kyoto University, Kyoto 606-8502, Japan3) Plasma Physics Laboratory, Osaka University, Osaka 565-0871, Japan
Abstract. A large amplitude m=1 mode is excited by LHCD in the WT-3 tokamak (a = 20cm,R = 65cm, Bt ≤ 1.75T). The mode excitation is accompanied with decrease of the magneticshear near the q = 1 surface and peaking of the soft X-ray (SX) emissivity profile, suggestingthat the mode is a pressure driven mode. The mode is suppressed by ECH near the q = 1surface, while enhanced by ECH near the plasma center. The mode stabilization is accompaniedwith a local enhancement of the SX emissivity near the q=1 surface.
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152
(EXP3/08) Origin of Rapid Impurity Penetration to PlasmaCenter During Disruption in Tokamak T-11M
S. V. Mirnov1), A. M. Belov1), D. Yu. Prokhorov2), A. G. Alekseyev1), I. N. Makashin1)
1)TRINITI, Troitsk, Moscow reg. 142190 RUSSIA2) Institute of Nuclear Fusion, RSC “Kurchatov Institute”, Moscow, 123182 RUSSIA
Abstract. The results of impurity penetration studies at T-11M tokamak just before and dur-ing the disruption are presented. Two scenarios of the process are considered: (i) initiated bythe major disruption, the latter being the main source of impurity, and (ii) minor disruptionor Locked Mode (LM) provide preliminary impurity penetration into the periphery followed bydeep internal disruption. Well-known “vacuum bubble” capture model is proposed for the ex-planation of rapid impurity penetration.
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153
(EXP3/09) Experimental Results on Pellet Injection andMHD from the RTP Tokamak
A. A. M. Oomens1), J. de Kloe1), F. J. B. Salzedas1), M. R. de Baar1), C. J. Barth1), M. N. A. Beurskens1),A. J. H. Donne1), B. de Groot1), G. M. D. Hogeweij1), F. A. Karelse1), O. G. Kruijt1), J. Lok1),N. J. Lopes Cardozo1), H. J. van der Meiden1), R. Meulenbroeks1), E. Noordermeer1), T. Oye-vaar1), R. W. Polman1), F. C. Schuller1), E. Westerhof1)
1) FOM-Instituut voor Plasmafysica ‘Rijnhuizen’, Association EURATOM-FOM, Trilateral Eu-regio Cluster, Nieuwegein, The Netherlands
Abstract. The ablation of hydrogen pellets has been studied in the Rijnhuizen Tokamak ProjectRTP with a diagnostic with high spatial and temporal resolution. It has been observed that(part of the) ablation cloud drifts away from the pellet in opposite direction. These drifts oc-cur in semi-periodical bursts. A summary of a detailed analysis of this drift of the cloud andits implications for the fueling profile is presented. Stabilization of m/n = 2/1 tearing modespreceding density limit disruptions, has been studied with modulated and continuous ECRH.The results indicate that EC heating of the islands under these conditions is very inefficient.The time dependence of the growth rate of the precursor mode is first algebraic, but becomesexponential in a later phase.
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154
(EXP3/10) Shape Effects on Sawtooth/Internal Kink Sta-bility and Confinement with EC Waves in the TCV Toka-mak
A. Pochelon1), F. Hofmann1), H. Reimerdes1), C. Angioni1), R. Behn1), R. Duquerroy1), I. Furno1),T. Goodman1), P. Gomez1), M. A. Henderson1), A. Martynov1), P. Nikkola1), O. Sauter1),A. Sushkov2)
1) CRPP-EPFL, Association EURATOM-Confederation Suisse, CH-1015 Lausanne EPFL, Switzer-land2) NFI-KIAE, RRC-Kurchatov, Russia
Abstract. This paper addresses the effect of plasma shaping (triangularity and elongation)on the sawtooth stability as well as the technique of current profile broadening using off-axiselectron cyclotron heating (ECH) to enlarge the stable operational range towards higher elon-gations. The plasma shape strongly influences the sawtooth period and amplitude. This effectis emphasised by ECH, with the sawtooth period becoming shorter at low triangularity or athigh elongation; for these plasma shapes, the pressure profile inside q=1 remains essentially flat.A comparison of the sawtooth response with marginal Mercier stability shows that the criticalpressure gradient at q=1 is particularly low for plasma shapes where the increased sawtoothrepetition frequency prevents the peaking of the pressure profiles. For theses shapes, the idealinternal kink is also found unstable from stability calculations. The stability of highly elongatedplasmas depends largely on current profiles. The operational range at low current has beenextended towards higher elongation using ECH heated discharges. Far off-axis second harmonicX-mode ECH power deposition proves to be an efficient tool for current profile tailoring allowinga significant elongation increase at constant quadrupole field.
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155
(EXP3/11) Control of RFP Dynamics with Rotating Heli-cal Fields
S. Masamune1), M. Iida1), K. Ohta1), H. Oshiyama1)
1) Kyoto Institute of Technology, Kyoto, Japan
Abstract. Control of MHD mode dynamics is essential to improving the reversed field pinch(RFP) experiments as a part of the fusion research programs. The STE-2 RFP has been oper-ated only with a vacuum vessel (τw ≤ 0.15ms) to test the idea of driving MHD mode and/orplasma rotation by using resonant rotating helical f ield (RHF) which is applied from outside ofthe vessel. We report new results which indicate direct interaction between the RHF and inher-ent tearing modes. Without the RHF, the magnetic disturbance outside the vessel is dominatedby almost locked m/n=1/8 core resonant tearing modes, growing with the scale of vessel timeconstant. The RHF tends to drive the mode rotation with a transient suppression of the modegrowth. Improved RFP performance has not become clear yet. Stabilization of the externalkink modes will also be discussed.
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156
(EXP3/12) MHD Characteristics in High-Beta Regime ofthe Large Helical Device
S. Sakakibara1), H. Yamada1), K. Y. Watanabe1), Y. Narushima1), K. Toi1), S. Ohdachi1),M. Takechi1), S. Yamamoto3), K. Narihara1), K. Tanaka1), N. Ashikawa2), P. de Vries1), M. Emoto1),H. Funaba1), M. Goto1), K. Ida1), H. Idei1), K. Ikeda1), S. Inagaki1), N. Inoue1), M. Isobe1),S. Kado1), O. Kaneko1), K. Kawahata1), K. Khlopenkov1), A. Komori1), S. Kubo1), R. Ku-mazawa1), S. Masuzaki1), T. Minami1), J. Miyazawa1), T. Morisaki1), S. Morita1), S. Mu-rakami1), S. Muto1), T. Mutoh1), Y. Nagayama1), Y. Nakamura1), H. Nakanishi1), K. Nishimura1),N. Noda1), T. Notake3), T. Kobuchi2), Y. Liang2), N. Ohyabu1), Y. Oka1), M. Osakabe1),T. Ozaki1), R. O. Pavlichenko1), B. J. Peterson1), A. Sagara1), K. Saito3), R. Sakamoto1),H. Sasao2), M. Sasao1), K. Sato1), M. Sato1), T. Seki1), T. Shimozuma1), M. Shoji1), H. Suzuki1),Y. Takeiri1), N. Tamura2), T. Tokuzawa1), Y. Torii3), K. Tsumori1), I. Yamada1), S. Yam-aguchi1), M. Yokoyama1), Y. Yoshimura1), T. Watari1), N. Nakajima1), K. Ichiguchi1), H. Taka-hashi4), W. A. Cooper5), K. Yamazaki1), O. Motojima1), Y. Hamada1), M. Fujiwara1)
1) National Institute for Fusion Science, Toki, Japan2) Department of Fusion Science, Graduate Univ. for Advanced Studies, Hayama, Japan3) Department of Energy Engineering and Science, Nagoya Univ., Nagoya, Japan4) Princeton Plasma Physics Laboratory, Princeton University, Princeton, USA5) Centre de recherches en physique des plasmas, Ecole Polytechnique Federale de Lausanne,Lausanne, Switzerland
Abstract. The highest volume averaged beta value βt of 2.2% at Bt = 0.75 T (gas puff) and2.4% at Bt = 1.3 T (pellet) in all of helical devices have been achieved in the Large Helical De-vice (LHD). The βt dependence of MHD activities has been investigated in NBI plasmas. Then/m = 1/2 mode excited in the core region and χ = 1 resonant modes in peripheral region havebeen observed. Both of the fluctuation amplitudes increase with βt and the pressure gradient.The strong n/m = 1/2 mode which affects plasma profile have been observed in high-βt dis-charges and the abrupt disappearance of the mode leads to the restoration of Te profile. Violentinstabilities which terminate the plasma and degradation of global energy confinement have notbeen observed so far.
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157
(EXP3/13) Comparative Study of CT Equilibrium and Sta-bility Using Controlled Current Drive: Compact RFP, Sphero-mak and ST in a Single Device
Y. Ueda1), Y. Ono1), M. Tsuruda1), Y. Murata1), G. Yamada1), H. Hayashiya1), T. Itagaki1),M. Katsurai1)
1) Department of Electrical Engineering, University of Tokyo, Tokyo, Japan
Abstract. A comparative experiment of low-aspect ratio (A ≈ 1.5) torus plasmas: compactRFPs, spheromaks and STs has been performed in the TS-3 and 4 CT devices. Especially, acompact RFP with q0 as large as 0.3 was found to have low-n dynamo mode n = 3 like n = 2mode of spheromak with q0 ≈ 0.5. Under OH current sustainment, magnetic fluctuation andloop voltage were observed to increase inversely with its q-value. A new edge-current-drivemethod was developed by use of axial CT merging, realizing a balanced current drive togetherwith the OH current drive. The balanced current drive was found to reduce the dynamo fluctua-tions of compact RFP and spheromak by factor 3. The multiple CT merging can be a promisingedge-current drive because of its intermittent current drive and its detailed control of currentprofiles using varied size of colliding CTs.
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158
(EXP3/14) Mode Dynamics and Confinement in the Re-versed Field Pinch
P. R. Brunsell1), H. Bergsaker1), J. H. Brzozowski1), M. Cecconello1), J. R. Drake1), J.-A. Malm-berg1), J. Scheffel1), D. D. Schnack2)
1) Alfven Laboratory, Royal Institute of Technology, Sweden2) Science Application International Corp., CA, USA
Abstract. Tearing mode dynamics and toroidal plasma flow in the RFP has been experi-mentally studied in the Extrap T2 device. A toroidally localised, stationary magnetic fieldperturbation, the “slinky mode” is formed in nearly all discharges. There is a tendency of in-creased phase alignment of different toroidal Fourier modes, resulting in higher localised modeamplitudes, with higher magnetic fluctuation level. The fluctuation level increases slightly withincreasing plasma current and plasma density. The toroidal plasma flow velocity and the iontemperature has been measured with Doppler spectroscopy. Both the toroidal plasma velocityand the ion temperature clearly increase with I/N . Initial, preliminary experimental resultsobtained very recently after a complete change of the Extrap T2 front-end system (first wall,shell, TF coil), show that an operational window with mode rotation most likely exists in therebuilt device, in contrast to the earlier case discussed above. A numerical code DEBSP hasbeen developed to simulate the behaviour of RFP confinement in realistic geometry, includingessential transport physics. Resulting scaling laws are presented and compared with results fromExtrap T2 and other RFP experiments.
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159
(EXP3/15) Nonlinear Dynamics of the Reversed Field Pinch:Torques, Dynamo, and Reconnection
D. J. Den Hartog1), D. Craig1), N. A. Crocker1), G. Fiksel1), P. W. Fontana1), A. K. Hansen1),C. C. Hegna1), S. C. Prager1), J. S. Sarff1)
1) Department of Physics, University of Wisconsin-Madison, Madison, Wisconsin U.S.A.
Abstract. The magnetic field configuration of the Reversed-Field Pinch (RFP) typically ex-hibits resistive tearing modes of poloidal mode number m = 1 resonant in the plasma core andm = 0 resonant in the plasma edge. In the Madison Symmetric Torus (MST) RFP, these fluctu-ations cause electromagnetic torques which alter the flow profile, and magnetic reconnection anddynamo effects which alter the magnetic configuration and current density profile. Described inthis paper are three key physics results: 1) The discovery of internal electromagnetic torquesbetween two core modes, a three-wave interaction requiring the mediation of the m = 0, n = 1mode at the plasma edge. 2) Direct measurements of the 〈v × B〉‖ dynamo at the plasma edgeconfirm that it balances Ohm’s law and is primarily driven by the m = 0 mode. 3) Measure-ments across the reconnection layer at the q = 0 resonant surface demonstrate the dominanceof m = 0 current density fluctuations in the vicinity of this resonant surface and show a phaseflip of the radial plasma flow velocity fluctuations across the resonant surface.
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160
(EXP3/16(R)) Active Feedback Control of the Wall Stabi-lized External Kink Mode
G. A. Navratil1), J. Bialek1), A. H. Boozer1), C. Cates1), H. Dahi1), M. E. Mauel1), D. Maurer1),M. Shilov1)
1) Columbia University, New York, NY 10027, USA
Abstract. Abstract. Active feedback control has been used in the HBT-EP tokamak to controlthe growth of the n=1 resistive wall mode. These experiments were carried out using a set ofthin stainless-steel wall segments with magnetic diffusion time of ∼0.4 ms positioned near theplasma boundary. In plasmas that would normally exhibit a strong ideal n = 1 external kinkmode without a nearby conducting wall, the resistive wall slows the growth of the external kinkto the ∼1 ms time scale where it can be stabilized by active feedback control. The approachtaken in these experiments is to use a network of active feedback coils mounted on the surface ofthe stainless-steel wall segments and driven by an active feedback control system that simulatesthe electrical response of a superconducting wall and minimizes the radial flux penetration ofthe perturbed mode field through the wall. This implementation of the so called ‘smart shell’in HBT-EP has 30 independent sensor/driver feedback loops.
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161
(EXP3/17) Progress in the Prediction of Disruptions inASDEX-Upgrade via Neural and Fuzzy-Neural Techniques
M. Versaci1), F. C. Morabito1), C. Tichmann2), G. Pautasso2), the ASDEX Upgrade Team2)
1) Universita “Mediterranea” degli Studi di Reggio Calabria, Associazione EURATOM-ENEA-CREATE, Via Graziella, Feo di Vito, I-89100 Reggio Calabria, Italy.2) Max-Planck-Institut fur Plasmaphysik (IPP), D-85748 Garching bei Munchen, Germany.
Abstract. The paper addresses the problem of predicting the onset of a disruption on the basisof some known precursors possibly announcing the event. The availability in real time of a largeset of diagnostic signals allows us to collectively interpret the data in order to decide whether weare near a disruption or during a normal operation scenario. As a relevant experimental example,a database of disruptive discharges in ASDEX-Upgrade has been analysed in this work. BothNeural Networks (NN’s) and Fuzzy Inference Systems (FIS) have been investigated as suitabletools to cope with the prediction problem. The experimental database has been exploited aimingto gain information about the mechanisms which drive the plasma column to a disruption. Theproposed processor will operate by implementing a classification of the shot type, and outputinga real number that indicates the time left before the disruption will effectively take place (ttd).
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162
Session EXP4 — Current Drive,Heating, Fuelling, Divertors &Edge Physics
(EXP4/06) Radio Frequency Experiments in JFT-2M: Demonstration ofInnovative Applications of a Travelling Wave Antenna . . . . . . . . . . 170
(EXP4/07) On the Fast Electron Beam, Consequent Generation of Electro-static Fields and Fast Ion Production in Front of LH Grills: Measure-ments and Theory . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 171
(EXP4/08) High Harmonic Fast Wave Heating Experiments on NSTX . . . 172
(EXP4/09) Heating, Current Drive and MHD Control Using ECH in TCV 173
(EXP4/01) Numerical Modelling of ICRF Physics Experi-ments in the Alcator C-Mod Tokamak
P. T. Bonoli1), R. L. Boivin1), M. Brambilla2), C. Fiore1), J. A. Goetz1), R. S. Granetz1),M. J. Greenwald1), A. Hubbard1), I. H. Hutchinson1), J. Irby1), B. LaBombard1), W. DavisLee1), B. Lipschultz1), E. S. Marmar1), A. Mazurenko1), E. Nelson-Melby1), D. Mossessian1),C. K. Phillips3), M. Porkolab1), J. Rice1), G. Schilling3), J. A. Snipes1), J. L. Terry1), S. Wolfe1),S. Wukitch1)
1) MIT Plasma Science and Fusion Center, Cambridge, MA (USA)2) Max Planck Institut fur Plasmaphysik, Garching (Germany)3) Princeton Plasma Physics Laboratory, Princeton, NJ (USA)
Abstract. A full-wave spectral code (TORIC) has been used to simulate mode convertedion Bernstein wave (IBW) propagation and absorption for the first time at high poloidal modenumber (−80 < m < +80). Converged wave solutions for the mode converted wave are obtainedin this limit and the predicted electron damping of the IBW is found to be consistent withexperimental measurements from the Alcator C-Mod tokamak. The TORIC code has also beencoupled to a bounce-averaged Fokker Planck module FPPRF and the combined codes are nowrun within the transport analysis tool TRANSP. This model was used to analyze off-axis hy-drogen minority heating experiments in C-Mod where an internal transport barrier was obtained.
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165
(EXP4/02) ECRH Experiments and Developments for LongPulses in Tore Supra
G. Giruzzi1), C. Darbos1), R. Dumont1), R. Magne1), Y. Peysson1), X. L. Zou, F. Bouquey1),L. Courtois1), G. T. Hoang1), F. Imbeaux1), M. Lennholm1), X. Litaudon1), P. Moreau1),A. L. Pecquet1), J. L. Segui1), M. Zabiego1), and the TORE SUPRA Team1)
1) Association EURATOM-CEA sur la Fusion, DEPARTEMENT DE RECHERCHES SURLA FUSION CONTROLEE, CEA-Cadarache, 13108 St. Paul-lez-Durance (FRANCE)
Abstract. The ECRH system presently under construction at CEA/Cadarache for the ToreSupra tokamak is described. The system will be equipped by 6 gyrotrons (118 GHz, 400 kW,cw), manufactured by Thomson Tubes Electroniques. The results of the tests of the prototypeand of the first series gyrotron are reported and discussed. The best performance obtained wasa pulse of 102 s at 310 kW average power on dummy load, which corresponds to the new recordenergy of 32 MJ. Results of first ECRH experiments on Tore Supra with the prototype gyrotronare also reported. 350 kW have been coupled in O-mode both to Ohmic and to LHCD plasmasin continuous or modulated pulses lasting up to 2 s. During non-inductive discharges, fully sus-tained by LHCD, a significant response of the hard X-ray signals to the ECRH power has beenobserved, despite the low power ratio between the two waves (0.35 MW EC / 4.5 MW LH waves).
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166
(EXP4/03) ECRF Experiments for Local Heating and Cur-rent Drive in JT-60U
Y. I. Ikeda1), S. Ide1), T. Suzuki1), A. Kasugai1), K. Takahashi1), K. Kajiwara1), A. Isayama1),T. Oikawa1), K. Hamamatsu1), Y. Kamada1), T. Fujita1), K. Sakamoto1), S. Moriyama1),M. Seki1), R. Yoshino1), T. Imai1), K. Ushigusa1), T. Fujii1) and the JT-60U Team1)
1) Japan Atomic Energy Research Institute, Ibaraki, Japan
Abstract. An ECRF program has been started to study the local heating and current drive inJT-60U. The frequency of 110GHz was adopted to couple the fundamental O-mode from low fieldside with an oblique toroidal injection angle for current drive. Experiments were performed atthe injection power of ∼1.5 MW by using three gyrotrons, each of which has generated the outputpower up to ∼0.8 MW for 3 seconds. A strongly peaked Te profile was observed and the centralelectron temperature increased up to ∼15 keV when the O-mode was absorbed on axis. Thelocal electron heating clarified the significant difference on the heat pulse propagation betweenin the plasmas with the internal transport barrier (ITB) and without ITB. The driven currentestimated by the Motional Stark Effect (MSE) diagnostic showed that the EC driven currentwas ∼0.2MA at the local electron temperature and density of Te ∼ 6keV, ne ∼ 0.7× 1019m−3.The measured driven current near the axis was consistent with the theoretical prediction usinga Fokker Planck code. In the case of co-ECCD, the sawteeth activity in NB heated plasma wascompletely suppressed for 1.5 s with the deposition at the inversion radius, while the sawteethwas enhanced for counter-ECCD at the same deposition condition.
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167
(EXP4/04) High Power ECCD Experiments at W7-AS
H. Maassberg1), J. Geiger1), H. Laqua1), N. B. Marushchenko1), M. Rome2), C. Wendland1),W7-AS Team1)
1) Max-Planck-Institute fur Plasmaphysik, EURATOM-Association, Garching, Germany2) Dipartimento di Fisica, Universita degli Studi, Milano, Italy
Abstract. At the W7-AS stellarator, high power electron cyclotron current drive (ECCD) ex-periments are analyzed. In these net-current-free discharges, the ECCD as well as the bootstrapcurrent are feedback controlled by an inductive current. Based on measured profiles, the neo-classical predictions for the bootstrap and the inductive current densities as well as the ECCDfrom the linear adjoint approach with trapped particles included are calculated, and the currentbalance is checked. Launch-angle scans at fixed density as well as density scans at fixed launch-angle are described.
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168
(EXP4/05) Current Profile Control Experiments in the LHCDPlasma on TRIAM-1M
K. Nakamura1), S. Itoh1), H. Zushi1), M. Sakamoto1), K. Hanada1), E. Jotaki1), Y. D. Pan2),S. Kawasaki1), H. Nakashima1)
1) Advanced Fusion Research Center, Research Institute for Applied Mechanics, Kyushu Uni-versity, Japan2) On leave from Southwestern Institute of Physics, China
Abstract. Controllability of current profile in long term discharges is studied by two kindsof combinations of wave spectra. First by superposition of higher N‖-spectrum wave at 2.45GHz on a target plasma driven by lower one at 8.2 GHz, a hollow j(r) has been achieved andsustained for 20 sec. The current profile can be well controlled below a threshold power of 14kW by varying the power of 2.45 GHz. A spontaneous transition phenomenon to a peaked j(r),however, has occurred over the threshold power even after a reasonable steady state conditionhas been obtained. Second using two opposite travelling waves at 8.2 GHz the total currentis clearly reduced and the j(r) becomes peaked when the backward travelling LHWs (BW) aresuperposed at low power of 20 kW. When the BW power is increased above 27 kW, furthercurrent reduction is suppressed, and then above 34 kW the direction of the current driven byBW is completely reversed and the j(r) becomes broad. Thus, current modification by BWshows a highly nonlinear behavior with respect to the BW power.
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169
(EXP4/06) Radio Frequency Experiments in JFT-2M: Demon-stration of Innovative Applications of a Travelling WaveAntenna
T. Ogawa1), K. Hoshino1), S. Kanazawa2), M. Saigusa2), T. Ido3), H. Kawashima1), N. Ka-suya4), Y. Takase4), H. Kimura1), Y. Miura1), K. Takahashi1), C. P. Moeller5), R. I. Pinsker5),C. C. Petty5)
1) Japan Atomic Energy Research Institute, Tokai-mura Ibaraki, 319-1195 Japan.2) Ibaraki University, Hitachi, Ibaraki, 316-8511 Japan.3) National Institute for Fusion Science, Toki, Gifu, 509-52 Japan.4) The University of Tokyo, Bunkyou-ku, Tokyo, 113-0033 Japan.5) General Atomics, San Diego, California, 92186-5608 U.S.A.
Abstract. Abstract. Several innovative applications of a traveling wave (combline) antennadesigned for fast wave current drive have been demonstrated for the first time in the JFT-2Mtokamak. High energy electrons of at least 10 keV were produced in the plasma core by highlydirectional fast waves in electron-cyclotron-heated plasmas. The ponderomotive potential of thebeat wave, produced by fast waves at two different frequencies, was directly measured for thefirst time by a heavy ion beam probe. Plasma production was demonstrated using the wavefields excited by the combline antenna over a wide range of toroidal magnetic fields (0.5∼2.2 T).
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170
(EXP4/07) On the Fast Electron Beam, Consequent Gen-eration of Electrostatic Fields and Fast Ion Production inFront of LH Grills: Measurements and Theory
V. Petrzilka1), F. Zacek1), B. Kolman1), F. Kroupa1), K. Jakubka1), J. Stoeckel1), R. Klima1),L. Krlin1), P. Pavlo1), J. Gunn2), M. Goniche2), V. Fuchs2), P. Devynck2), D. Tskhakaya3),S. Kuhn3), J. A. Tataronis4), J. Lorincik5)
1) Institute of Plasma Physics, Ass. Euratom/IPP.CR, P.O.Box 17, 182 21 Praha 8, CzechRepublic2) Ass. Euratom-CEA, CEA/Cadarache, 13108 Saint Paul-lez Durance, France3) Department of Theoretical Physics, University of Innsbruck, Austria4) University of Wisconsin, 1500 Engineering DR, Madison, WI 53706, USA5) Institute of Physical Chemistry, Dolejskova 3, 182 23 Praha 8, Czech Republic
Abstract. Abstract: The paper presents measurements of radial variations of the floating po-tential at the Tore Supra (TS) tokamak ergodic divertor plate and in front of the CASTORtokamak lower hybrid (LH) grill, due to the presence in these two locations of the fast particlebeam generated in front of LH grills. The paper also presents a scanning electron microscope(SEM) and secondary ion mass spectrometric (SIMS) analysis of an eroded graphite tile fromthe TS LH grill guard limiter, performed in order to check the authors’ theoretical conclusionthat fast ions can be generated in a thin layer in front of LH grills and that they can contributeto damage of tokamak vessel components. The paper first presents theoretical conclusions thatare relevant to the experimental data and then the experimental results.
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171
(EXP4/08) High Harmonic Fast Wave Heating Experimentson NSTX
J. R. Wilson1), R. Bell1), M. Bitter1), P. T. Bonoli2), M. D. Carter3), D. Gates1), J. C. Hosea1),B. LeBlanc1), R. Majeski1), T. K. Mau4), J. Menard1), D. Mueller1), S. Paul1), C. K. Phillips1),R. I. Pinsker5), A. Rosenberg1), P. Ryan3), S. A. Sabbagh6), D. Stutman7), D. Swain3), Y. Takase8),J. Wilgen3)
1) Princeton Plasma Physics Laboratory, Princeton NJ, USA2) Massachusetts Institute of Technology, Cambridge MA, USA3) Oak Ridge National Laboratory, Oak Ridge TN, USA4) University of California at San Diego, San Diego CA, USA5) General Atomics, San Diego CA, USA6) Columbia University, New York NY, USA7) Johns Hopkins University, Baltimore MD, USA8) University of Tokyo, Tokyo, Japan
Abstract. A radio frequency (rf) system has been installed on the National Spherical TorusExperiment (NSTX) with the aim of heating the plasma and driving plasma current. The sys-tem consists of six rf transmitters, a twelve element antenna and associated transmission linecomponents to distribute and couple the power from the transmitters to the antenna elementsin a fashion to allow control of the antenna toroidal wavenumber spectrum. To date, powerlevels up to 3.85 MW have been applied to the NSTX plasmas. The frequency and spectrum ofthe rf waves has been selected to heat electrons via Landau damping and transit time magneticpumping. The electron temperature has been observed to increase from 400 to 900 eV with littlechange in plasma density resulting in a plasma stored energy of 59 kJ , a toroidal beta, βT =10% and a normalized beta, βn = 2.7.
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172
(EXP4/09) Heating, Current Drive and MHD Control Us-ing ECH in TCV
T. Goodman1), TCV Team
1) CRPP/EPFL, Lausanne, Switzerland
Abstract. The 6 beam 2nd harmonic X-mode (X2), 3MW, ECH/ECCD system of the TCVtokamak allows a fine tailoring of the deposition profiles in the plasma. The sensitivity of thesawtooth period to the deposition location is used to increase the equilibria reconstruction andray-tracing accuracy. Off-axis ECH, followed by on-axis counter-ECCD produces improved cen-tral confinement regimes in which τEe exceeds RLW scaling by a factor of 3.5. The PRETORtransport code (incorporating an RLW local transport model but constrained by the experi-mental density profiles) predicts an extreme sensitivity of τEe to the deposition location of thecounter-ECCD. This is confirmed by experiments. Sawtooth simulations using PRETOR, in-cluding the effects of current drive with inputs from the TORAY ray-tracing code, are in goodagreement with experimental results. These results are an initial benchmark for the package ofanalysis codes, LIUQE / TORAY / PRETOR used during ECH/ECCD experiments on TCV.
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173
(EXP4/10) Behaviour of Compact Toroid Injected into theExternal Magnetic Field
M. Nagata1), N. Fukumoto1), H. Ogawa2), T. Ogawa2), K. Uehara2), H. Niimi3), T. Shibata2),Y. Suzuki4), Y. Miura2), T. Uyama1), H. Kimura2), JFT-2M Group2)
1) Faculty of Engineering, Himeji Institute of Technology, Himeji, Hyogo, Japan2) Department of Fusion Plasma Research, Japan Atomic Energy Research Institute, Naka,Ibaraki, Japan3) Center for Advanced Research of Energy Technology, Hokkaido University, Sapporo, Hokkaido,Japan4) Plasma Theory Laboratory, Japan Atomic Energy Research Institute, Naka, Ibaraki, Japan
Abstract. Interaction of a compact toroid (CT) plasma with an external field and with a toka-mak plasma has been studied experimentally on the JFT-2M and FACT devices. Fast framingcamera and SX emission profile measurements indicate shift and/or reflection motions of theCT plasma. New electrostatic probe measurements indicate that the CT plasma reaches atleast up to the separatrix for discharges with the toroidal field strengths of 1.0–1.4 T, and thatthere exists a trailing plasma behind the CT. We have observed a large amplitude fluctuationon ion-saturation current and magnetic coil signals. Power spectrum analysis suggests that thisfluctuation is related to magnetic reconnection between the CT plasmoid and the toroidal field.The low density trailing plasma may be able to move across the external magnetic field moreeasily in the injector owing to the Hall effect.
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174
(EXP4/11) Quasi-Steady-State Operation around Opera-tional Limit in HT-7
J. Li1), J. K. Xie1), B. N. Wan1), J. R. Luo1), X. Gao1), Y. Zhao1), Y. Yang1), G. L. Kuang1),Y. Bao1), B. J. Ding1), Y. X. Wan1), and HT-7 Team
1) Institute of Plasma Physics, Academia Sinica, China
Abstract. Efforts have been made on HT-7 tokamak for extending the stable operation bound-aries. Extensive RF boronization and siliconization have been used and wider operational Hugilldiagram was obtained. Transit density reached 1.3 time of Greenwald density limit in ohmicdischarges. Stationary high performance discharge with qa = 2.1 has been obtained after sili-conization. Confinement improvement was obtained due to the significant reduction of electronthermal diffusivity χe in the out region of the plasma. Improved confinement phase was alsoobserved by LHCD under the density range 70% ∼ 120% of Greenwald density limit. The weakhollow current density profile was attribute to off-axis LHW power deposition. Code simulationsand measurements showed a good agreement of off-axis LHW deposition. Supersonic molecularbeam injection has been successfully used to get stable high-density operation in the range ofGreeenwald density limit.
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175
(EXP4/12) Study of LHW and IBW Synergy Experimenton the HT-7 Superconducting Tokamak
X. Gao1), HT-7 Team1)
1) Institute of Plasma Physics, Academia Sinica, Hefei, China
Abstract. A successful experiment on lower hybrid wave (LHW) and ion Bernstein wave(IBW) synergy has been carried out in the HT-7 superconducting tokamak. With 500 kW ofLHW heating power and 200 kW of injected IBW power, it is observed that the ion temperatureincreases from 500 eV to about 850 eV, the electron temperature increases from 800 eV to 1.2keV, and the averaged electron density increases from 0.9 × 1019m−3 to 2.6 × 1019m−3. Theplasma parameters were obviously enhanced by means of the LHW and IBW heating and theirsynergy. The charge-exchange spectra of the neutral particle analysis (NPA) diagnostics dataclearly showed that the high-energy ion tail which was produced by the LHW was decreased bythe synergy with the IBW, and the bulk ion temperature was increased. The mechanism of theLHW and IBW synergy effect is discussed.
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176
(EXP4/13) Hydrogen Cluster-like Behaviour during Super-sonic Molecular Beam Injection on the HL-1M Tokamak
L. Yao1), Y. Zhou1), J. Y. Cao1), B. B. Feng1), Z. Feng1), J. L. Luo1), J. F. Dong1), L. W. Yan1),W. Y. Hong1), K. H. Li1), Y. Liu1), E. Y. Wang1) and HL-1M Team1)
1) Southwestern Institute of Physics, Chengdu, China
Abstract. Pulsed supersonic molecular beam injection (SMBI) has been developed success-fully and used in the HL-1M tokamak. It is an attempt to enhance the penetration depth andfuelling efficiency. With a penetration depth of hydrogen particles beyond 8 cm, the rising rateof electron density, dne/dt, was up to 7.6 × 1020m−3s−1 without disruption, and reached thehighest plasma density ne = 8.2× 1019m−3 on HL-1M. With SMBI the plasma energy confine-ment time, τE , measured by diamagnetism is 10–30 % longer than that with gas puffing whenother discharge conditions are kept the same. The fuelling method of SMBI has recently beenimproved to make a survey of the cluster effects within the beam. A series of new phenomenashow the interaction of the beam (including clusters) with the toroidal plasma. Hydrogen clus-ters may be produced in the beam according to the Hagena empirical scaling law of clusteringonset, Γ∗ = (kd0.85P0)/T 2.29
0 ). If Γ∗ > 100, clusters will form. In the present experiment Γ∗ isabout 127.
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177
(EXP4/14) Recent Efficiency Improvements and Experi-mental Results on the Repetitive Compact Toroid Accel-erator at UC Davis
K. L. Baker1), D. Q. Hwang1), R. D. Horton1), R. W. Evans1), H. S. McLean2), S. D. Terry3)
1) Department of Applied Science, University of California at Davis, Livermore, CA2) Present Address: Lawrence Livermore National Laboratory, Livermore, CA3) Present Address: University of Wisconsin, Madison, Wisconsin
Abstract. This article reports on the acceleration dynamics of a compact toroid plasma configu-ration in a background hydrogen gas. The acceleration dynamics are investigated experimentallyusing magnetic probes and chordal interferometry. These measurements are then compared totwo-dimensional simulations that show good agreement with experiments. The experimentalmeasurements indicate that the velocity and field strength initially increase as a function of ac-celerator voltage, however, at higher voltages the compact toroid velocity ceases to increase withincreased accelerator voltage. In our investigation, we examine the “blowby” effect along withadditional processes such as mass accumulation and charge exchange as potential mechanismscontributing to the stagnation of the compact toroid’s velocity.
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178
(EXP4/15) Influence of Anisotropic Ion Heating on Con-finement of the Tandem Mirror
K. Ishii1), T. Goto1), A. Itakura1), I. Katanuma1), Y. Katsuki1), T. Saito1), T. Tamano1),A. Mase2)
1) Plasma Research Center, University of Tsukuba, Tsukuba, Japan2) Advanced Science and Technology Center for Cooperative Research, Kyushu University, Ka-suga, Japan
Abstract. Ion transport in the velocity space due to the Alfven ion-cyclotron (AIC) fluctua-tion was investigated using a developed diagnostic device ELECA (end-loss energy componentanalyzer) in the tandem mirror. The AIC waves are excited spontaneously in the central cellby carrying out anisotropic ion heating with injection of the ion-cyclotron range of frequency(ICRF) wave. Hump structure was observed on the energy distribution function of the end-lossions, and was correlated with the excitation of the AIC wave. A remarkable feature is thattrapped ions with characteristic energy flow selectively into the loss region so as to cause thehump structure which appears in the energy region (1.0–5.0) keV. The AIC waves enhance theion diffusion from the trapped region to the loss region in the velocity space, and decrease theconfinement. However, it was found that the hump structure decreased in the high potentialmode and the high confining potential more than 3 kV could confine a large mount of loss energyflux due to the AIC fluctuations because of the existence of the upper limit of the hump energy,even in strong excitation of the AIC wave caused by high anisotropy.
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179
(EXP4/16) Current Drive Experiments in the HIT-II Spher-ical Tokamak
T. R. Jarboe1), P. Gu1), V. A. Izzo1), P. E. Jewell1), K. J. McCollam1), B. A. Nelson1), R. Ra-mon1), A. J. Redd1), P. E. Sieck1), R. J. Smith1), M. Nagata2), T. Uyama2)
1) University of Washington, Seattle, Washington, United States of America2) Himeji Institute of Technology, Himeji, Japan
Abstract. The Helicity Injected Torus (HIT) program has made progress in understandingrelaxation and helicity injection current drive. Helicity-conserving MHD activity during theinductive (Ohmic) current ramp demonstrates the profile flattening needed for coaxial helicityinjection (CHI). Results from cathode and anode central column (CC) CHI pulses are consis-tent with the electron locking model of current drive from a pure n = 1 mode. Finally, lowdensity CHI, compatible with Ohmic operation, has been achieved. Some enhancement of CHIdischarges with the application of Ohmic is shown.
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180
(EXP4/17) Additional Heating Experiments of FRC Plasma
S. Okada1), T. Asai1), F. Kodera1), K. Kitano1), T. Suzuki1), K. Yamanaka1), T. Kanki1), M. In-omoto1), S. Yoshimura1), M. Okubo1), S. Sugimoto1), S. Ohi1), S. Goto1)
1) Plasma Physics Laboratory, Graduate School of Engineering, Osaka University, Osaka, Japan
Abstract. Additional heating experiments of neutral beam (NB) injection and application oflow frequency wave on a plasma with extremely high averaged beta value of about 90% - a fieldreversed configuration (FRC) plasma - are carried out on the FRC Injection experiment (FIX)apparatus. These experiments are made possible by translating the FRC plasma produced in aformation region of a theta pinch to a confinement region in order to secure better accessibilityto heating facilities and to control plasma density. By appropriate choice of injection geometryand the mirror ratio of the confinement region, the NB with the energy of 14keV and the currentof 23A is enabled to be injected into the FRC in the solenoidal confining field of only 0.04–0.05T.Confinement is improved by this experiment. Ion heating is observed by the application of lowfrequency (80kHz ; about 1/4 of the ion gyro frequency) compressional wave. A shear wave,probably mode converted from the compressional wave, is detected to propagate axially.
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181
(EXP4/18) Impact of Pellet Injection on Extension of Op-erational Region in LHD
R. Sakamoto1), H. Yamada1), K. Tanaka1), K. Narihara1), S. Morita1), S. Sakakibara1), S. Ma-suzaki1), L. R. Baylor2), P. W. Fisher2), S. K. Combs2), M. J. Gouge2), S. Kato1), A. Komori1),O. Kaneko1), N. Ashikawa3), P. de Vries1), M. Emoto1), H. Funaba1), M. Goto1), K. Ida1),H. Idei1), K. Ikeda1), S. Inagaki1), M. Isobe1), S. Kado1), K. Kawahata1), K. Khlopenkov1),S. Kubo1), R. Kumazawa1), T. Minami1), J. Miyazawa1), T. Morisaki1), S. Murakami1), S. Muto1),T. Mutoh1), Y. Nagayama1), Y. Nakamura1), H. Nakanishi1), K. Nishimura1), N. Noda1),T. Notake4), T. Kobuchi3), Y. Liang3), S. Ohdachi1), N. Ohyabu1), Y. Oka1), M. Osakabe1),T. Ozaki1), R. O. Pavlichenko1), B. J. Peterson1), A. Sagara1), K. Saito4), H. Sasao3), M. Sasao1),K. Sato1), M. Sato1), T. Seki1), T. Shimozuma1), M. Shoji1), S. Sudo1), H. Suzuki1), M. Takechi1),Y. Takeiri1), N. Tamura3), K. Toi1), T. Tokuzawa1), Y. Torii4), K. Tsumori1), I. Yamada1), S. Ya-maguchi1), S. Yamamoto4), Y. Yoshimura1), K. Y. Watanabe1), T. Watari1), K. Yamazaki1),Y. Hamada1), O. Motojima1), M. Fujiwara1)
1) National Institute for Fusion Science, Toki, Gifu 509-5292, Japan2) Oak Ridge National Laboratory, PO Box 2009, Oak Ridge, TN 37831-8071, USA3) Department of Fusion Science, School of Mathematical and Physical Science, Graduate Uni-versity for Advanced Studies, Hayama, 240-0193, Japan4) Department of Energy Engineering and Science, Nagoya University, 464-8603, Japan
Abstract. Pellet injection has been used as a primary fueling scheme in Large Helical Device(LHD). Pellet injection has extended an operational region of NBI plasmas to higher densitieswith maintaining preferable dependence of energy confinement on density, and achieved severalimportant data, such as plasma stored energy (0.88 MJ), energy confinement time (0.3 s), β (2.4% at 1.3 T) and density (1.1× 1020m−3). These parameters cannot be attained by gas puffing.Ablation and subsequent behavior of plasma has been investigated. Measured pellet penetrationdepth that is estimated by duration of the Hα emission is shallower than predicted penetrationdepth from the simple neutral gas shielding (NGS) model. The penetration depth can be ex-plained by NGS model with fast ion effect on the ablation. Just after ablation, redistributionof ablated pellet mass was observed in short time (∼ 400ms). The redistribution causes shallowdeposition and low fueling efficiency.
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182
(EXP4/19) On the Neutral Beam Heating of MAST
R. J. Akers1), L. C. Appel1), E. Arends1), C. Byrom2), M. Cox1), M. Tournianski1), P. G. Car-olan1), N. J. Conway1), S. Gee1), M. Nightingale1), M. Gryaznevich1), K. G. McClements1),M. Walsh3), A. Sykes1), P. Helander1), H. R. Wilson1), S. Medley4), L. Roquemore4)
1) EURATOM/UKAEA Fusion Association, Culham Science Centre, Abingdon, UK2) University of Manchester, Institute of Science and Technology, Manchester, UK3) Walsh Scientific Ltd, Culham Science Centre, Abingdon, UK4) Princeton Plasma Physics Laboratory, Princeton, New Jersey 08543, USA
Abstract. We present a brief synopsis of the phenomenology associated with Neutral Beamheating (NBI) of the MAST Spherical Tokamak. Results, although at an early stage of analysis,are highly encouraging for the future auxiliary heating capabilities of the device. In particular,only 500-800kW of NBI power (30keV H injection, ∼10% of design total) is sufficient to signif-icantly increase Te and to double the thermal electron pressure. Further, preliminary resultsindicate that the ion temperature increases by a factor of ∼3, qualitatively consistent withinsystematic errors with that expected, given tolerable fast ion loss and ITER confinement scalingIPB98(y,1). NBI data recorded so far exhibit suprathermal ion tail formation and ohmic dis-charges a phenomenologically similar tail following Internal Reconnection Events. The injectedfast ions, corresponding to 10-50% of the total stored energy, are responsible for driving a widevariety of sporadic, high frequency MHD modes, the low magnetic fields (and hence low Alfvenspeeds) inherent to the Spherical Tokamak geometry providing an ideal testing ground for fast-particle MHD theory.
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183
(EXP4/20) Recent Developments in Tokamak Edge Physicsat Garching
D. P. Coster1), X. Bonnin2), K. Borrass1), H.-S. Bosch1), B. Braams3), H. Buerbaumer4),A. Kallenbach1), M. Kaufmann1), J.-W. Kim1), E. Kovaltsova5), E. Mazzoli6), J. Neuhauser1),D. Reiter7), V. Rozhansky5), R. Schneider2), W. Ullrich1), S. Voskoboynikov5), P. Xantopoulos8)
and the ASDEX Upgrade Team1)
1) Max-Planck-Institut fur Plasmaphysik, EURATOM Association, 85748 Garching, Germany2) Max-Planck-Institut fur Plasmaphysik, EURATOM Association, 17489 Greifswald, Germany3) Courant Institute, NYU, New York, USA4) Institut fur Allgemeine Physik, TU Wien, Austria, Association EURATOM-OEAW5) St. Petersburg State Technical University, St. Petersburg, Russia6) Politecnico di Torino, Italy7) Dusseldorf University, Dusseldorf, Germany8) N.C.S.R. ‘Demokritos’, GR-15310 A.G. Paraskevi, Greece
Abstract. The use of the SOLPS edge plasma and neutral codes is discussed. First SOLPS4.0B2-Eirene is used to understand the role of C radiation in lowering the divertor target power loadand the effects of power, density and divertor geometry on helium compression and enrichment.Then the newer B2-SOLPS5.0 is used in an interpretive mode to derive edge transport coeffi-cients from experimental data, and then, with drift effects included, to model scrape-off layercurrents. Finally some analytic and semi-analytic results are mentioned. The understandingof divertor power loading, helium compression and enrichment, edge transport coefficients anddrifts are all important for current machines, but vital for extrapolation to future machines.
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184
(EXP4/21) Experimental Study of Tokamak Plasma Inter-action with Lithium Capillary-Pore Systems
V. A. Evtikhin1), I. E. Lyublinski1), A. V. Vertkov1), E. A. Azizov2), S. V. Mirnov2), V. B. Lazarev2),S. M. Sotnikov2)
Abstract. Behavior and influence of two lithium capillary-pore system - based rail limiter mod-els on the discharge parameters at heat fluxes of about 10MW/m2 and discharge duration of 0.1s have been experimentally investigated in a T-11M to substantiate the lithium capillary-poresystem (CPS) application as a plasma facing material. Limiter thermal loads, plasma-limiterthermal balance, physics of lithium erosion and lithium accumulation in tokamak plasma havebeen studied. Reradiation effects in the lithium vapor at the plasma periphery has been de-tected and studied. The stability of lithium CPS under tokamak plasma conditions has beeninvestigated and confirmed.
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185
(EXP4/22) Impurity Radiation Efficiency and Retention inTore Supra Ergodic Divertor Experiments
R. Guirlet1), J. Hogan2), P. Monier-Garbet1), Y. Corre1), C. De Michelis1), P. Ghendrih1), R. Gi-annella1), C. Grisolia1), A. Grosman1), J. Gunn1), B. Schunke1)
1) Association Euratom-CEA sur la Fusion, C.E. Cadarache, 13108 St-Paul-les-Durance, France2) Oak Ridge National Laboratory, Fusion Energy Division, PO Box 2009, Oak Ridge, TN37831-8072, USA
Abstract. The screening effect of Ne and C in ergodic divertor plasmas is studied. Spectro-scopic measurements show that the screening mechanism is not the same for the two impurities.A 2D model explains this difference by the longer penetration length of neutral Ne. 3D mod-elling of the plasma edge with the BBQ code confirms the brightness profile shape dependenceon the edge Te. The 1D impurity transport code MIST coupled to BBQ interprets the screeningeffect as posssibly due to strong impurity outfluxes coming out of the ergodic region.
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186
(EXP4/23) Detachment in Variable Divertor Geometry onTCV
R. A. Pitts1), A. Loarte2), B. P. Duval1), J.-M. Moret1), J. A. Boedo3), D. P. Coster4), J. Ho-racek5), A. S. Kukushkin6) and the TCV Team1)
1) Centre de Recherches en Physique des Plasmas, Association EURATOM-Confederation Su-isse, Ecole Polytechnique Federale de Lausanne, CH-1015 Lausanne, Switzerland2) EFDA-CSU, Max Planck Institut fur Plasmaphysik, D-85748 Garching, Germany3) Fusion Energy Program, University of California, San Diego, CA 92093-0417, USA4) Max-Plank-Insitut fur Plasmaphysik, Boltzmannstr. 2, D-85748, Garching, Germany.5) IPP, Academy of Sciences of the Czech Republic, Za Slovankou 3, POB 17, 182 21 Praha6) ITER Joint Central Team, Garching Joint Working Site, Boltzmannstr. 2, D-85748, Garching,Germany.
Abstract. Although the requirement of shape flexibility in TCV precludes the use of fixed baffleor optimised divertor target structures, it does allow for the investigation of diverted equilibrianot possible in more conventional tokamaks. One such single null configuration is simultaneouslycharacterised by a very short inboard poloidal depth from X-point to strike point on a verticaltarget and an extremely long poloidal depth to a horizontal target on the outboard side. Densityramp discharges leave the inboard target plasma attached even at the highest densities, whilstclear partial detachment is observed at the outboard target. Modeling of this configuration usingthe B2-Eirene code package shows that the outboard divertor achieves high recycling at verylow densities, with the rollover to detachment occurring near the outer strike point very soonafter the density ramp begins. An important result of the modeling effort is that, due to the lowapparent densities in the TCV outboard divertor, the code cannot quantitatively reproduce theabsolute level of observed detachment without artificially increasing fivefold the charged particlesink due to three-body and radiative recombination.
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187
(EXP4/24) Operation of ASDEX Upgrade with High-Z WallCoatings
V. Rohde, R. Neu, K. Krieger, H. Maier, A. Geier, A. Tabasso, A. Kallenbach, R. Pugno,D. Bolshukhin, M. Zarrabian and the ASDEX Upgrade Team
Abstract. The material for plasma facing components of a future fusion device is still notdecided. At present most experiments use graphite, because of its good thermo mechanicalproperties and the low radiation potential of carbon. Due to the high erosion yield and, espe-cially, due to the codeposition with tritium, its use in a fusion reactor is still questionable. Basedon the good experience using tungsten as divertor material in ASDEX Upgrade, which demon-strated that a divertor tokamak can be operated with a tungsten divertor without reduction ofthe performance, a step by step strategy was followed. Main sources of the carbon are predictedat the inner heat shield, which covers the central column. Tungsten test tiles confirm the ero-sion at this position due to charge exchange neutral, but also a non negligible ion sputteringcomponent. A first step was done by siliconisation. In ASDEX Upgrade the maximal siliconconcentration was 0.002. Consequently the performance of the experiment was not influencedby silicon radiation. A second step was done by tungsten coating of 1.2m2 of the inner heatshield. Experiments are done without subsequent wall coating, which would cover the tungsten.Spectroscopically measured central tungsten densities are always below ≈ 5 ∗ 10−6 and mostlybelow the detection limit. Again no influence on the plasma performance parameters are found.Extrapolation to ITER conditions yields concentrations, which will not prohibit successful op-eration. The next step in ASDEX Upgrade will be a mostly tungsten covered inner heat shieldat the next experimental campaign.
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188
(EXP4/25) The Control of Divertor Carbon Erosion/Redepositionin the DIII-D Tokamak
D. G. Whyte1), W. P. West2), C. P. C. Wong2), R. Bastasz3), J. N. Brooks4), W. R. Wampler3),N. H. Brooks2), J. W. Davis5), R. Doerner1), A. A. Haasz5), R. C. Isler6), G. L. Jackson2),R. G. Macaulay-Newcombe5), M. R. Wade6)
1) Fusion Energy Research Program, University of California, San Diego, La Jolla, Califor-nia USA2) General Atomics, P.O. Box 85608, San Diego, California 92186-5608 USA3) Sandia National Laboratories, Albuquerque, New Mexico & Livermore, California USA4) Argonne National Laboratory, Argonne, Illinois USA5) University of Toronto Institute for Aerospace Studies, Toronto Canada6) Oak Ridge National Laboratory, Oak Ridge, Tennessee USA
Abstract. The DIII-D tokamak has demonstrated an operational scenario where the graphite-covered divertor is free of net erosion. Reduction of divertor carbon erosion is accomplishedusing a low temperature (detached) divertor plasma that eliminates physical sputtering. Like-wise, the carbon source rate arising from chemical erosion is found to be very low in the detacheddivertor. Near strikepoint regions, the rate of carbon deposition is ∼ 3 cm/burn-year, with acorresponding hydrogenic codeposition rate > 1kg/m2/burn-year; rates both problematic forsteady-state fusion reactors. The carbon net deposition rate in the divertor is consistent withcarbon arriving from the core plasma region. Carbon influx from the main wall is measured tobe relatively large in the high-density detached regime and is of sufficient magnitude to accountfor the deposition rate in the divertor. Divertor redeposition is therefore determined by non-divertor erosion and transport. Despite the success in reducing divertor erosion on DIII-D withdetachment, no significant reduction is found in the core plasma carbon density, illustrating theimportance of non-divertor erosion and the complex coupling between erosion/redeposition andimpurity plasma transport.
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189
(EXP4/26) Recycling and Wall Pumping in Long DurationDischarges on TRIAM-1M
M. Sakamoto1), S. Itoh1), K. Nakamura1), H. Zushi1), K. Hanada1), E. Jotaki1), Y. D. Pan2),S. Kawasaki1), H. Nakashima1)
1) Advanced Fusion Research Center, Research Institute for Applied Mechanics, Kyushu Uni-versity, Fukuoka, Japan2) Southwestern Institute of Physics, China
Abstract. Recycling and wall pumping have been studied comparing low (∼ 1018m−3) andhigh (∼ 1019m−3) density long duration plasmas in TRIAM-1M. The recycling coefficient ofeach plasma increases with time. There exist two time constants in the temporal evolution ofthe recycling coefficient. One is a few seconds and the other is about 30 s. They may relate withcharacteristic times during which the physical adsorption and absorption due to the CX neutralsreach the equilibrium state, respectively. The wall pumping rates of low and high density plas-mas are evaluated to be ∼ 1.5×1016 atoms m−2s−1 and ∼ 4×1017 atoms m−2s−1, respectively.The difference is caused by the difference of the total amount of the CX neutral flux with theenergy of < 0.7keV. In the ultra-long discharge (∼ 70 min), the recycling coefficient becomesunity or more and again decreases below unity, i.e. the wall repeats a process of being saturatedand refreshed. This refreshment of the wall seems to be caused by the co-deposition of Mo, whichis a material of the limiter and divertor plates. In the high power and high density experiments,the wall saturation phenomenon has been observed. The discharge duration limited by the wallsaturation decreases with increase in the density.
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190
(EXP4/27) Characteristic Behaviors of Divertor Scrape-offPlasma in the TPE-2M Reversed Field Pinch
K. Hayase1), Y. Sato1), S. Kiyama1), Y. Maejima1), H. Koguchi1), K. Sugisaki1), M. Watanabe2),M. Maeyama3)
Abstract. The divertor discharge of reversed field pinch (RFP) has been studied in TPE-2M(R/r = 0.87m/0.27m). The characteristic behaviors of divertor plasma and its effects on thecore plasma are investigated by visible spectroscopy and probe measurements. The observed iondensity profile at the divertor plate surface is rather smoothed out to a single hump possiblydue to the particle scattering by a large amplitude fluctuation (around 15 %) of magnetic fieldin the open shell (divertor) region. An anomalous particle loss through the X-point region issuggested. The discharge depends on the position of X-point; when it locates near the plasmasurface, the plasma may be less stable and the discharge terminates earlier presumably becausethe shell proximity of the core plasma surface is deteriorated. In this case, a large-amplitude,sometimes burst-like, fluctuation, is seen. The edge behaviors of core plasma in the shell regionare less sensitive to the divertor field.
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191
(EXP4/28) Island Divertors: Concepts and Status of Ex-perimental and Modelling Results
F. Sardei1), Y. Feng1), J. Kisslinger1), P. Grigull1), R. Koenig1), K. McCormick1), W7-AS Team1)
Abstract. Basic features of the island divertor concepts for low-shear advanced stellarators likeW7-AS and W7-X (intrinsic island divertors) and for high-shear heliotrons like CHS and LHD(local island divertors, LID) and first results for both concepts are shortly reviewed. The divert-ing fields of island divertors are either intrinsic (low-shear case) or externally imposed (high-shearcase). The associated field perturbations are very small compared to tokamak poloidal field di-vertors, which explains the high flexibility of island divertor configurations, but sufficiently largeto generate divertor-viable islands. Although the physics of island divertors is expected to besimilar to that of tokamak poloidal field divertors, leading geometrical parameters significantlydiffer from those of comparable-size tokamaks. Furthermore, strong three-dimensional effectsarise from toroidally discontinuous target plates. For both the intrinsic and local island divertorconfigurations, the island structure and the plasma diversion have been verified experimentally.For W7-AS, stable high recycling conditions could be demonstrated in a stellarator for a sim-ple island divertor geometry. In CHS with the LID field switched on, a strong increase of thepumping efficiency was measured, resulting in a 50% reduction of the core density for the samegas-puff rate as in the non-LID case. 3D numerical transport studies with the EMC3/EIRENEcode predict, for typical W7-AS values of the upstream density and power into the SOL, partialdetachment with 80% power loss by carbon radiation in the SOL and significant momentumlosses associated with the island divertor geometry.
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192
(EXP4/29) Static and Dynamic Behaviors of Plasma De-tachment in Divertor Simulator NAGDIS-II
S. Takamura1), N. Ohno1), Y. Uesugi1), D. Nishijima1), M. Motoyama1), N. Hattori1), H. Arakawa1),N. Ezumi2), S. I. Krasheninnikov3), A. Pigarov3), U. Wenzel4)
1) Department of Energy Engineering and Science, Graduate School of Engineering, Nagoya,University, Nagoya 464-8603, Japan2) Nagano National College of Technology, Nagano 381-8550, Japan3) University of California at San Diego, La Jolla, California 92093-0411, USA4) Max-Planck-Institut fur Plasmaphysik, EURATOM Association, D-1011 Berlin, Germany
Abstract. Abstract. We have performed comprehensive investigation on the static and dynamicbehaviors in detached recombining plasmas in the linear divertor plasma simulator, NAGDIS-II.For the stationary plasma detachment, the transition from electron-ion recombination (EIR) tomolecular activated recombination (MAR) has been observed by injecting hydrogen gas to highdensity He plasmas. The particle loss rate due to MAR is found to be comparable to that ofEIR. We have also performed experiments on injection of a plasma heat pulse produced by rfheating to the detached recombining He plasma to demonstrate the dynamic behavior of thevolumetric plasma recombination. Negative spikes in Balmer series line emissions were observedsimilar to the so called negative ELM observed in tokamak divertors, which were analyzed withcollisional-radiative model in detail. Rapid increase of the ion flux to the target plate was ob-served associated with the re-ionization of the highly excited atoms generated by EIR.
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193
(EXP4/30) Ion Bernstein Wave Heating Experiments inHT-7 Superconducting Tokamak
Y. Zhao1), J. Li1), HT-7 Team1)
1) Institute of Plasma Physics, Chinese Academy of Sciences, P.R. of China
Abstract. Ion Bernstein Wave (IBW) heating has been investigated in HT-7 superconductingtokamak. The electron heating mode was concentrated on a deuterium plasma with an injec-tion power up to 320kW. Direct electron heating via electron Landau damping from the IBWwas observed. The bulk electron temperature showed a significant rise with a heating factor,∆Te × ne/PRF, up to 7.4 (eV × 1013cm−3/kW) and maximum increment of electron tempera-ture was above 400 eV. IBW heating was combined with pellet injection, which also showed itsunique merit.
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194
(EXP4/31) Mechanism of H-mode Triggering by CT Injec-tion in the STOR-M Tokamak
C. Xiao1), D. R. McColl1), A. Hirose1), S. Sen2)
1) Plasma Physics Laboratory, University of Saskatchewan, Saskatoon, Canada2) School of Mathematical and Computational Sciences, University of St. Andrews, UK; also atCentre of Plasma Physics, Sapta Swahid Path, India
Abstract. The H-mode like discharges induced by tangential compact torus (CT) injectionin the STOR-M tokamak and by other techniques, such as by a short current pulse and elec-trode/limiter biasing, are commonly characterized by an increase in the electron density, sig-nificant reduction in the Hα radiation level, and steepening of the edge density profile. TheH-modes induced by negative biasing and edge heating via a short current pulse result in largenegative electric potential biasing, formation of a strong poloidal velocity shear, and slowdownof the toroidal flow. In contrast, in the H-modes induced by positive biasing and CT injection,only moderate positive electric potential biasing occurs. Formation of a strong poloidal velocityshear is absent, but the toroidal flow velocity increases. A plausible mechanism that does notrequire strong velocity shear itself to suppress the long wavelength turbulence is attributed tothe stabilizing role of the curvature in the toroidal velocity profile.
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195
Session EXP5 — Transport
Contents
(EXP5/01) Dynamic Behaviour of Transport in Normal and Reversed ShearPlasmas with Internal Barriers in JT-60U . . . . . . . . . . . . . . . . . 198
(EXP5/03) Transport of Particles and Impurities in DIII-D Discharges withInternal Regions of Enhanced Confinement . . . . . . . . . . . . . . . . . 199
(EXP5/04) Improved Fueling and Transport Barrier Formation with PelletInjection from Different Locations on DIII-D . . . . . . . . . . . . . . . 200
(EXP5/05) Role of Magnetic Configuration and Heating Power in ITB For-mation in JET . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 201
(EXP5/31) Transition to H-mode Regime in JET with and without PumpedDivertors . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 225
(EXP5/32) Measurement of Impurity Transport Coefficients in the Con-fined Plasma of ASDEX Upgrade . . . . . . . . . . . . . . . . . . . . . . . 226
(EXP5/33) Propagation of Cold Pulses and Heat Pulses in ASDEX Upgrade227
(EXP5/34) Development of Fast Helium Beam Emission Spectroscopy forPlasma Density and Temperature Diagnostics . . . . . . . . . . . . . . . 228
197
(EXP5/01) Dynamic Behaviour of Transport in Normal andReversed Shear Plasmas with Internal Barriers in JT-60U
S. V. Neudatchin1), T. Takizuka2), Yu. Dnestrovskij1), H. Shirai2), T. Fujita2), A. Isayama2),Y. Kamada2), Y. Koide2)
1) Nuclear Fusion Institute, RRC “Kurchatov Institute”, Moscow, Russia2) Japan Atomic Energy Research Institute, Naka Fusion Research Establishment, Naka-machi,Naka-gun, Ibaraki-ken 311-0193, Japan
Abstract. Transport evolution in reverse shear (RS) and normal shear (NrS) JT-60U tokamakplasmas with internal transport barrier (ITB) is described as a combination of various fast andslow time scales processes. Abrupt in time and wide in space (∼0.3 of minor radius) variations ofelectron and ion heat diffusivities δχe, i (seen as “spontaneous-like” simultaneous rise and decayof Te,i in two spatial zones) are found for weak ITBs in both RS and NrS plasmas. Profiles ofδχe, i in RS plasmas with strong ITB are usually localized near ITB “foot” inside smaller spaceregion. The maximum of the heat flux variation is located near position of the minimum ofsafety factor q in various RS plasmas, and variation is extended in positive shear region. Inwardand outward heat pulse propagations created by δχe, i and sawtooth-like crashes are analyzed,small values of χe, i and absence of heat pinch are found in ITB region. Another source of abruptδχe, i inside most of plasma volume, including significant part of weak ITB in RS plasmas, isseen as ELM-induced H-L transitions.
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198
(EXP5/03) Transport of Particles and Impurities in DIII-DDischarges with Internal Regions of Enhanced Confinement
D. R. Baker1), M. R. Wade2), L. R. Baylor2), J. C. DeBoo1), C. M. Greenfield1), W. A. Houl-berg2), and B. W. Stallard3)
1) General Atomics, San Diego, California USA2) Oak Ridge National Laboratory, Oak Ridge, Tennessee USA3) Lawrence Livermore National Laboratory, Livermore, California USA
Abstract. In DIII-D discharges with centrally enhanced confinement and peaked density pro-files there is an accumulation and peaking of impurities in the core, as is predicted by neoclassicaltheory. In VH-mode discharges with an outer region of enhanced confinement and a broad den-sity profile, the neoclassical thermal screening effect causes the impurities to accumulate nearthe edge. In discharges with anomalously high transport, the measured particle diffusivitiesare close to the measured thermal diffusivities and the density profiles are proportional to afractional power of 1/q. When the plasma has regions of enhanced confinement, where theion thermal transport is near neoclassical, then the particle transport also exhibits neoclassicaltransport behavior such as low diffusivities, neoclassical pinch effects and ion temperature gra-dient screening. Here we present experimental results of both impurity and electron transport inDIII-D discharges with large internal regions of enhanced confinement and discuss the relevanceof these results to future tokamak designs.
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199
(EXP5/04) Improved Fueling and Transport Barrier For-mation with Pellet Injection from Different Locations onDIII-D
L. R. Baylor1), T. C. Jernigan1), P. Gohil2), G. L. Schmidt3), K. H. Burrell2), S. K. Combs1),D. R. Ernst3), C. M. Greenfield2), R. J. Groebner2), W. A. Houlberg1), C. Hsieh2), M. Mu-rakami1), P. B. Parks2), M. Porkolab4), W. D. Sessions5), G. M. Staebler2), E. Synakowski3) andthe DIII-D Team
1) Oak Ridge National Laboratory, Oak Ridge, TN, USA2) General Atomics, San Diego, CA, USA3) Princeton Plasma Physics Laboratory, Princeton, NJ, USA4) Massachusetts Institute of Technology, Cambridge, MA, USA5) Tennessee Technological University, Cookeville, TN, USA
Abstract. Pellet injection has been employed on DIII-D from different injection locations tooptimize the mass deposition for density profile control and internal transport barrier formation.Transport barriers have been formed deep in the plasma core with central mass deposition fromhigh field side (HFS) injected pellets and in the edge with pellets that trigger L-mode to H-modetransitions. Pellets injected from all locations can trigger the H-mode transition, which dependson the edge density gradient created and not on the radial extent of the pellet deposition. Pel-lets injected from inside the magnetic axis from the inner wall or vertical port lead to strongercentral mass deposition than pellets injected from the low field side (LFS) and thus yield deepermore efficient fueling.
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200
(EXP5/05) Role of Magnetic Configuration and HeatingPower in ITB Formation in JET
Abstract. An extensive database of more than 250 OS plasmas with different power level,heating scheme and current profile control has been formed and used to determine the con-ditions for ITB formation. The results of the database analysis are presented which confirmthe crucial role of excess power and other key plasma parameters. JET experiments have alsoshown that the magnetic configuration, particularly the q-profile and magnetic shear, also playan important role in the ITB formation. Different scenarios, which have been used on JET totailor the q-profile, will be discussed. Results of the predictive modelling of a series of JET OSplasmas are presented alongside experimental results.
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201
(EXP5/06) Transport Analysis of ITB Discharges
A. G. Peeters1), A. Bergmann1), G. Conway1), H. Meister1), G. V. Pereverzev1), G. Tardini1),O. Gruber1), M. Manso2), F. Serra2), R. C. Wolf1)
1) Max Planck Institut fuer Plasmaphysik, IPP Garching2) CNF, IST Lisboa
Abstract. Detailed studies of the confinement physics in advanced scenarios on ASDEX Up-grade are presented. The confinement of the improved H-mode can largely be unified with thatof the standard H-mode. The increase in the H-factor is partly due to density peaking, partlydue to the density dependence in the scaling law, and partly due to the non-thermal popula-tion if a non-corrected scaling law is used. The confinement region of these discharges can besimulated well with the ITG/TEM models. Clear transport barriers are observed to be formedin discharges with reversed shear. Turbulence suppression is observed in these discharges, andwithin the error bars this is in agreement with the E×B paradigm. Monte Carlo simulations ofthe neoclassical transport in discharges with reversed shear show a reduction compared to thestandard theory, and agree reasonably well with the experimental ion heat conductivity over alarge inner region.
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202
(EXP5/07) On the Role of Rational Surfaces on Transportin Fusion Plasmas
L. Garcia1), M. A. Pedrosa2), C. Hidalgo2), B. A. Carreras3), R. Balbın2), A. Lopez-Fraguas2),V. E. Lynch3), J. A. Jimenez2), B. van Milligen2), E. Sanchez2)
1) Universidad Carlos III, 28911 Leganes, Madrid, Spain2) Asociacion EURATOM-CIEMAT, 28040 Madrid, Spain3) Oak Ridge National Laboratory, Oak Ridge, TN 37831, U.S.A.
Abstract. Experimental evidence of ExB sheared flows linked to rational surfaces has beenobtained in the plasma edge region of the TJ-II stellarator. A possible explanation of the flowstructure near the rational surface is the nonlinear beating of the magnetic field component ofa vacuum field island with a plasma instability. To simulate the main characteristics of theexperimental results, we use a resistive interchange model with the rotational transform profiledetermined by the vacuum magnetic field calculations.
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203
(EXP5/08) Reversed Magnetic Shear Experiment in theHL-1M Tokamak
E. Y. Wang1), Y. Liu1), J. Y. Cao1), J. F. Dong1), X. T. Ding1), Z. C. Deng1), G. C. Guo1),X. D. Li1), X. Hua1), W. Y. Hong1), J. Rao1), L. B. Ran1), W. B. Xu1), D. M. Xu1), L. W. Yan1),L. Yao1), Y. Zhou1), N. M. Zhang1) and HL-1M Team1)
1) Southwestern Institute of Physics, Chengdu, Sichuan 610041, China
Abstract. An ohmic shear reversed configuration has been obtained in HL-1M by combinedcontrol of plasma current rise and supersonic Molecular Beam Injection (MBI). An equilibriumreconstruction code has been applied to the HL-1M database, which stores data from 32 mag-netic probes to derive the current density profile and safety q profile. The current density profilederived from the code is found to be hollow in core plasma and q is nonmonotonic. They arecharacterized by the peaked density profile and hollow electron temperature profiles. The hol-low temperature profile leads to a hollow current density profile and reversed magnetic shear.Current profile control by LHCD was conducted in HL-1M. The sawtooth and m=1 mode insta-bilities were observed to be suppressed by LHCD. The current profile and q profile reconstructedby the equilibrium code show a hollow current profile during LHCD and they are related to thenonmonotonic q profile. In ECRH off-axis experiments the occurrence of the compound saw-tooth observed by the soft X-ray diode array implies that the hollow current density profileswere formed.
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204
(EXP5/09) Transport Mechanisms and Enhanced Confine-ment Studies in RFX
M. Valisa1), V. Antoni1), L. Apolloni1), M. Bagatin1), W. Baker1), O. Barana1), R. Bartiromo1),P. Bettini1), A. Boboc1), T. Bolzonella1), A. Buffa1), A. Canton1), S. Cappello1), L. Carraro1),R. Cavazzana1), G. Chitarin1), S. Costa1), F. D’Angelo1), S. Dal Bello1), A. De Lorenzi1),D. Desideri1), D. F. Escande2), L. Fattorini1), P. Fiorentin1), P. Franz1), E. Gaio1), L. Garzotti1),L. Giudicotti1), F. Gnesotto1), L. Grando1), S. C. Guo1), P. Innocente1), A. Intravaia1), R. Loren-zini1), A. Luchetta1), G. Malesani1), G. Manduchi1), G. Marchiori1), L. Marrelli1), P. Martin1),E. Martines1), S. Martini1), A. Maschio1), A. Masiello1), F. Milani1), M. Moresco1), A. Murari1),P. Nielsen1), M. O’Gorman1), S. Ortolani1), R. Paccagnella1), R. Pasqualotto1), B. Pegurie1),S. Peruzzo1), R. Piovan1), N. Pomaro1), A. Ponno1), G. Preti1), M. E. Puiatti1), G. Ros-tagni1), F. Sattin1), P. Scarin1), G. Serianni1), P. Sonato1), E. Spada1), G. Spizzo1), M. Spo-laore1), C. Taliercio1), G. Telesca1), D. Terranova1), V. Toigo1), L. Tramontin1), N. Vianello1),M. Viterbo1), L. Zabeo1), P. Zaccaria1), P. Zanca1), B. Zaniol1), L. Zanotto1), E. Zilli1), G. Zollino1)
1) Consorzio RFX, Associazione Euratom – ENEA sulla Fusione, Corso Stati Uniti 4, 35127Padova, Italy2) UMR 6633 CNRS-Universite de Provence, Avenue Normandie-Niemen, 13397 Marseille Cedex20, France
Abstract. The results of an extensive study on transport mechanisms and on improved con-finement scenarios in RFX are reported. The scaling of the thermal conductivity in the core withthe Lundquist number indicates that the magnetic field in this region is not fully stochastic, asproved by the existence of thermal barriers observed in Single Helicity configurations. The elec-trostatic transport at the edge has been proved to depend on the highly sheared ExB flow whichhas been interpreted by fluid and Monte Carlo models. Regimes of improved confinement havebeen obtained in the core by Poloidal Current Drive techniques and the electrostatic transporthas been reduced at the edge by biasing experiments. A radiation mantle by impurity seedinghas been found to successfully reduce the local plasma wall interaction without significantlydeteriorating the plasma performance.
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205
(EXP5/10) Summarized General Scaling Laws Covering overthe Representative Tandem-Mirror Operations in Gamma10
T. Cho1), M. Hirata1), H. Hojo1), M. Ichimura1), K. Ishii1), A. Itakura1), I. Katanuma1), J. Koh-agura1), Y. Nakashima1), T. Saito1), S. Tanaka1), Y. Tatematsu1), M. Yoshikawa1), T. Tamano1),K. Yatsu1), S. Miyoshi1)
1) Plasma Research Centre, University of Tsukuba, Ibaraki, Japan
Abstract. Generalized scaling laws covering over the representative tandem-mirror operationalmodes from 1979 to 2000 in GAMMA 10 [i.e., characterized in terms of (i) a high-potentialmode having kV-order ion-confining potentials (φc) and thermal-barrier potentials (φb), and (ii)a hot-ion mode yielding fusion neutrons with 10-20-keV bulk-ion temperatures] are found andsummarized for providing the possibilities of combining the characteristics from each mode intonovel extended operational modes. The physics scalings of the formation of the plasma confiningpotentials as well as the associated effects of the potentials on plasma-parameter improvementsare systematically investigated as follows: (i) The potential-formation scalings in the two repre-sentative modes are consolidated and generalized on the basis of the consistency with the novelfindings of wider validity of Cohen’s strong electron-cyclotron heating (ECH) theory coveringover both modes. (ii) The produced potentials, in turn, provide a favorable novel scaling of theincrease in the central-cell electron temperatures Te with increasing φb, limited by the availableECH powers. The scaling of Te with φb is well interpreted by the Pastukhov theory of plasmapotential confinement, as a similarly reported scaling of potential-confined ion temperatureswith φc. (iii) A scaling of φc [kV] with ECH powers in a plug region, PECH [kW], and thecentral-cell densities, nc [1018m−3], is summarized as φc = 1.73× 10−4P 1.73
ECH exp(−0.33nc). (iv)Consequently, under the assumption of the validity of the extension of these theoretically wellinterpreted scaling data, the formation of Pastukhov’s predicted φc of 30 kV for confining Q = 1plasmas is scaled to require 5-MW PECH .
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206
(EXP5/11) Turbulence and Transport Studies in the EdgePlasma of the HT-7 Tokamak
W. Baonian1), M. Song1), B. L. Ling1), G. S. Xu1), Y. Zhao1), J. R. Luo1), J. Li1), HT-7 Team1)
1) Institute of Plasma Physics, Chinese Academy of Sciences, P.O.Box 1126, Hefei, China 230031
Abstract. The edge fluctuations and transport in the HT-7 tokamak are investigated using aLangmuir probe in ohmic and IBW heating discharges. The normalized fluctuation levels are inthe range of 35-50% and 10-25% for electron density and temperature respectively and have anon-Boltzmann relation in the SOL. In IBW heated plasma, the particle confinement is greatlyimproved, the poloidal velocity shear in the SOL is strongly modified, equivalent to an addi-tional poloidal velocity in electron diamagnetic direction. A de-correlation in the fluctuationsand suppression in turbulent transport by the effect of the poloidal shear are found to be acommon mechanism. The electrostatic driven turbulent transport can account for a significantpart of global particle loss in both ohmic and IBW heated plasmas.
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207
(EXP5/12) Pellet Injection during RF Heating on HT-7Tokamak
Y. Yang1), Y. Zhao1), X. Gao1), C. Y. Xia1), J. Li1), B. N. Wan1), J. R. Luo1), X. M. Gu1),L. Q. Hu1), S. Y. Zhang1), Y. X. Jie1), J. K. Liu1), X. D. Tong1), Y. F. Chen1) and HT-7 Team
1) Institute of Plasma Physics, Chinese Academy of Sciences (ASIPP), Hefei, P. R. China
Abstract. One of the main focuses of the plasma research on the HT-7 superconducting toka-mak is the exploration of advanced steady-state operation. RF heating and pellet injectionexperiments are carried out. D and H pellet injection has been utilized successfully for extend-ing the operational density and modification of plasma parameters. Effective heating has alsobeen observed in the 1999 and 2000 experimental campaigns. Recently, experiments on HT-7illustrate a good effect on the off-axis RF heating by H and D pellet injection. The heatingeffect is observed clearly, the reason for which may lie in the steep electron pressure gradientcreated by the pellet injection. Combined with the steep local ne gradient, good off-axis heatingcould provide another condition for profile and local transport studies. In addition, shortly afterthe injection of the pellet, the coupling efficiency of the wave is enhanced and confinement isimproved. In the paper, the experiments are introduced and discussed.
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208
(EXP5/13) Energy Confinement and Sawtooth Stabiliza-tion by ECRH at High Electron Density in FTU Tokamak
C. Sozzi1), S. Cirant1), A. Airoldi1), G. Bracco2), A. Bruschi1), P. Buratti2), F. Gandini1),G. Granucci1), A. Jacchia1), H. Kroegler2), E. Lazzaro1), S. Nowak1), G. Ramponi1), O. Tud-isco2), B. Angelini2), M. L. Apicella2), G. Apruzzese2), E. Barbato2), L. Bertalot2), A. Bertoc-chi2), G. Buceti2), A. Cardinali2), C. Centioli2), R. Cesario2), S. Ciattaglia2), V. Cocilovo2),F. Crisanti2), R. De Angelis2), M. De Benedetti2), B. Esposito2), D. Frigione2), L. Gabellieri2),G. Gatti2), E. Giovannozzi2), C. Gormezano2), M. Grolli2), M. Leigheb2), G. Maddaluno2),M. Marinucci2), G. Mazzitelli2), P. Micozzi2), F. P. Orsitto2), D. Pacella2), L. Panaccione2),M. Panella2), V. Pericoli-Ridolfini2), L. Pieroni2), S. Podda2), G. Pucella2), G. B. Righetti2),F. Romanelli2), S. E. Segre3), A. Simonetto1), P. Smeulders2), E. Sternini2), N. Tartoni2),A. A. Tuccillo2), V. Vitale2), G. Vlad2), V. Zanza2), M. Zerbini2), F. Zonca2)
1) Associazione EURATOM-ENEA-CNR, Istituto di Fisica del Plasma, Milano, Italy2) Associazione EURATOM-ENEA sulla Fusione, C.R. Frascati, Frascati, Roma, Italy3) INFM and Universita di Roma “Tor Vergata”, Roma, Italy
Abstract. The dependence of the local effective electron thermal diffusivity and of the globalenergy confinement on the intensity and the distribution of the heat source is explored in ECRHexperiments performed on FTU tokamak. Energy transport is analysed by applying a dominantEC heating with Pecrh ≥ 0.8MW and Poh ≈ 0.2MW during ECRH, at the frequency of 140GHz correspondent to the fundamental electron cyclotron resonance at Btor = 5T. In order todiscriminate the underlying diffusive transport against internal disruptions events, ECRH is alsoused to stabilize sawteeth. In these conditions, the predominance of ECRH over OH allows agood estimate of heat fluxes. Ion heating in the order of Ti/Ti ≈ 30% is observed in experimentswith central density up to ≈ 1020m−3.
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209
(EXP5/14) Results from Transient Transport Experimentsin Rijnhuizen Tokamak Project: Heat Convection, Trans-port Barriers and ‘Non-local’ Effects
P. Mantica1), G. Gorini1)2), G. M. D. Hogeweij3), J. de Kloe3), N. J. Lopes Cardozo3), A. M. R. Schil-ham3)
1) Istituto di Fisica del Plasma ‘P.Caldirola’, Associazione Euratom-ENEA-CNR, Milano, Italy2) INFM and Dipartimento di Fisica ‘G.Occhialini’, Universita degli Studi di Milano-Bicocca,Milano, Italy3) FOM Instituut voor Plasmafysica ‘Rijnhuizen’, Associatie Euratom-FOM, Trilateral EuregioCluster, 3430 BE Nieuwegein, The Netherlands
Abstract. An overview of experimental transport studies performed on the Rijnhuizen Toka-mak Project (RTP) using transient transport techniques in both Ohmic and ECH dominatedplasmas is presented. Modulated Electron Cyclotron Heating (ECH) and oblique pellet injection(OPI) have been used to induce electron temperature (Te) perturbations at different radial loca-tions. These were used to probe the electron transport barriers observed near low order rationalmagnetic surfaces in ECH dominated steady-state RTP plasmas. Layers of inward electron heatconvection in off-axis ECH plasmas were detected with modulated ECH. This suggests thatRTP electron transport barriers consist of heat pinch layers rather than layers of low thermaldiffusivity. In a different set of experiments, OPI triggered a transient rise of the core Te dueto an increase of the Te gradient in the 1 < q < 2 region. These transient transport barrierswere probed with modulated ECH and found to be due to a transient drop of the electron heatdiffusivity, except for off-axis ECH plasmas, where a transient inward pinch is also observed.Transient transport studies in RTP could not solve this puzzling interplay between heat diffu-sion and convection in determining an electron transport barrier. They nevertheless providedchallenging experimental evidence both for theoretical modelling and for future experiments.
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210
(EXP5/15) Mid-high Z Impurity Behaviour in High Tem-perature Tokamak Plasmas
D. Pacella1), M. Leigheb1), L. Gabellieri1), L. Panaccione1), M. Zerbini1), M. Mattioli4), M. May3),M. Finkenthal3), K. B. Fournier2), W. H. Goldstein2)
1) Associazione EURATOM-ENEA sulla Fusione, Frascati, Rome (Italy)2) Lawrence Livermore Laboratories, Livermore,CA (USA)3) Johns Hopkins University, Baltimore,MD (USA)4) ENEA guest
Abstract. Impurity transport and temperature effects have been studied in FTU high temper-ature ECRH heated plasmas, by means of X-VUV emissions from mid-high Z elements (Fe, Ge,Mo, W). Experiments have been performed in collaboration with the Johns Hopkins University(JHU) and the Lawrence Livermore National Laboratory (LLNL). Medium-high Z elements havethe advantage to not be fully ionized even at very high electron temperature (tens of keV) andto exhibit a large variety of soft X and VUV emissions very sensitive to the local plasma prop-erties like temperatures, non thermal effects, turbulence, changes of transport properties and soon. For example X-ray emissions (L-shell) of intrinsic Mo in the plasma core, heated by ECRHpower at about 8 keV during the current ramp up with a magnetic shear still negative or zero,revealed a negligible impurity transport and a central impurity peaking. Impurity transport inECRH heated plasmas have been also studied by means of Ge and W, injected by laser ablation.
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211
(EXP5/16) H-mode Features under ECRH on T-10
Yu. V. Esipchuk1), V. V. Alikaev1), A. A. Borshegovskij1), V. Chistyakov1), Yu. Dnestrovskij1),M. Dremin1), S. Grashin1), L. Eliseev1), L. Khimchenko1), G. Kirnev1), N. Kirneva1), A. Kislov1),D. A. Kislov1), S. V. Krilov1), A. Melnikov1), T. Myalton1), G. Notkin1), Yu. D. Pavlov1),V. Poznyak1), I. Roy1), D. Shelukhin1), S. Soldatov1), A. Sushkov1), K. Tarasyan1), V. Trukhin1),E. Trukhina1), V. Vershkov1)
Abstract. The main feature of the H-mode obtained on T-10 under ECRH (with HF powerabsorbed in the plasma Pab = 0.8MW) is the spontaneous density ne increase at Dα intensitydecrease (up to 4 times). This ne rise is the result of electron transport barrier formation ina narrow region (∆rH
∼= 3cm) near the limiter. Confinement enhancement in H-mode up toIL = τH
E /τLE∼= 1.6− 1.7 in general is the result of τ
A∼ ne dependence. The thermal transport
barrier contribution to total confinement enhancement is small. Threshold power PLHth does not
contradict ITER scaling. Maximal IL values are achieved at low qL ∼ 2.2, i.e. the dependencePLH
th (qL) exists under T-10 conditions. The results of the radial electric field and turbulencemeasurements are discussed.
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212
(EXP5/17) Regimes with Sharp Density Profile and En-hanced Energy Confinement under Deuterium Pellet Injec-tion and ECRH in T-10
Yu. D. Pavlov, Yu. Dnestrovskij, A. A. Borshegovskij, V. V. Chistiakov, A. V. Gorshkov,Yu. V. Gott, N. V. Ivanov, A. M. Kakurin, L. Khimchenko, A. V. Khramenkov, N. Kirneva,S. V. Krylov, V. A. Krupin, V. V. Matveev, T. B. Mialton, A. Yu. Novikov, V. V. Piterskij,S. V. Popovichev, V. V. Prut, I. N. Roy, M. B. Safonova, V. V. Sannikov, S. A. Shibaev,A. Sushkov, V. Trukhin, V. V. Volkov, V. Zaveriaev
Nuclear Fusion Institute, Russian Research Center “Kurchatov Institute”, Moscow, Russia
Abstract. Experiments with deuterium pellet injection have been carried out on T-10. Themain plasma parameters are as follows: Ip ≈ 250− 300kA, ne ≈ (3− 4)× 1013cm−3, PECRH ≈400− 700kW. A deuterium pellet was injected into the plasma at the stage of auxiliary ECRH.The main results are as follows: (a) a rapid sharpening of the plasma density profile with theincrease of ne(0) after pellet injection and the sustainment of this profile during the whole ECRHpulse while the averaged chord density remained constant; (b) the lowering of Zeff(0) and theredistribution of Zeff(r); (c) the growth of ion temperature ; (d) the increase of plasma energycontent and of energy confinement time τE by a factor of ∼ 1.5 as compared to τE before pelletinjection. The experimental results obtained are close to the RI-mode in TEXTOR after impu-rity feeding and can be described by the Canonical Profiles Transport Model.
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213
(EXP5/18) Confinement Bifurcation by Current DensityProfile Perturbation in TUMAN-3M Tokamak
S. V. Lebedev1), M. V. Andreiko1), L. G. Askinazi1), V. V. Bulanin2), M. I. Vildjunas1),V. E. Golant1), N. A. Zhubr1), V. A. Kornev1), S. V. Krikunov1), L. S. Levin1), A. I. Smirnov1),V. S. Roitershtein2), V. Rozhansky2), V. V. Rozhdestvensky1), M. Tendler3), A. S. Tukachin-sky1), S. Voskoboynikov2)
1) Ioffe Physico-Technical Institute, RAS, St.-Petersburg, Russia2) St.-Petersburg State Technical University, St.-Petersburg, Russia3) Alfven Laboratory, EURATOM-NFR Association, Stockholm, Sweden
Abstract. In the recent experiments performed on TUMAN-3M the possibility to switch on/offthe H-mode by current density profile perturbations has been shown. The j(r) perturbationswere created by fast Current Ramp Up/Down or by Magnetic Compression produced by a fastincrease of the toroidal magnetic field. It was found that the Current Ramp Up (CRU) andMagnetic Compression (MC) are useful means for H-mode triggering. The Current Ramp Down(CRD) triggers H-L transition. The difference in the j(r) behavior in these experiments sug-gests the peripheral current density may not be the critical parameter controlling L-H and H-Ltransitions. Confinement bifurcation in the above experiments could be explained by the unifiedmechanism: variation of a turbulent transport resulting from radial electric field emerging nearthe edge in the conditions of alternating toroidal electric field Ej and different electron and ioncollisionalities. According to the suggested model the toroidal field Eϕ arising in the peripheryduring the CRU and MC processes amplifies Ware drift, which mainly influences electron com-ponent. As a result the favorable for the transition negative (inward directed) Er emerges. In theCRD scenario, when Eϕ is opposite to the total plasma current direction, the mechanism shouldgenerate positive Er, which is thought to be unfavorable for the H-mode. The experimental dataon L-H and H-L transitions in various scenarios and the results of the modeling of Er emergingin the CRU experiment are presented in the paper.
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214
(EXP5/19) Local Physics Basis of Confinement Degrada-tion in JET ELMy H-mode Plasmas and Implications forTokamak Reactors
Abstract. ELMy H-mode plasmas form the basis of conservative performance predictions fortokamak reactors of the size of ITER. Relatively high performance for long durations has beenachieved and the scaling is favorable. It will be necessary to sustain low Zeff and high density forhigh fusion yield. This paper studies the degradation in confinement and increase in the anoma-lous heat transport observed in two JET plasmas: one in which the degradation occurs withan intense gas puff, and the other with a spontanous transition at the heating power thresholdfrom Type I to III ELMs. Linear gyrokinetic analysis gives the growth rate, γlin of the fastestgrowing mode. Our results indicate that the flow-shearing rate ωExB and γlin are large near thetop of the pedestal. Their ratio decreases approximately when the confinement degrades andthe transport increases. This suggests that tokamak reactors may require intense toroidal orpoloidal torque input to maintain sufficiently high |ωE×B|/γlin near the top of the pedestal forhigh confinement.
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215
(EXP5/20) Correlation Between Core and Pedestal Tem-peratures in JT-60U
D. R. Mikkelsen1), H. Shirai2), N. Asakura2), T. Fujita2), T. Fukuda2), T. Hatae2), S. Ide2),A. Isayama2), Y. Kamada2), Y. Kawano2), Y. Koide2), O. Naito2), Y. Sakamoto2), T. Tak-izuka2), H. Urano3)
1) Princeton Plasmas Physics Laboratory2) JAERI Naka Fusion Research Establishment3) Hokkaido University
Abstract. The ‘stiffness’ of thermal transport in ELMy H-modes is explored in a series ofcarefully chosen JT-60U plasmas and with temperature predictions based on several transportmodels. Four scans of pedestal temperature, Tped, with constant heating power and one heat-ing power scan with constant Tped are presented. We find that 30–80% increases in Tped areassociated with 10–70% increases in core temperature even though the total heating power isconstant. Increasing the heating power by 45% gives almost the same core temperatures (anda 12% density increase) in a group of five plasmas with the same pedestal temperature. Theresults can be characterized as having relatively ‘soft’ transport in the plasma periphery andrelatively ‘stiff’ transport in the core. Another series of experiments varied the heating in thedeep core by employing different groups of neutral beams that deposit their power on-axis andoff-axis. In these plasmas on-axis heating produces systematically more peaked temperatureprofiles; the rise from the periphery to the central region is ∼20% higher in plasmas that have60% more heating power inside r=a/2. Transport models are tested by solving the power bal-ance equations to predict temperatures, which are then compared to the measurements. TheRLWB and IFS/PPPL models’ predictions generally agree with the measured temperatures, butthe Multimode model uniformly predicts temperatures that are too high except for the centralsawtoothing region.
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216
(EXP5/21) Progress in Quantifying the Edge Physics of theH-mode Regime in DIII-D
R. J. Groebner1), D. R. Baker1), J. A. Boedo2), K. H. Burrell1), T. N. Carlstrom1), R. D. Dera-nian3), E. J. Doyle4), J. R. Ferron1), P. Gohil1), G. R. McKee5), R. A. Moyer2), C. L. Rettig4),T. L. Rhodes4), D. M. Thomas1), T. H. Osborne1), and W. P. West1)
1) General Atomics, San Diego, California USA2) University of California-San Diego, La Jolla, California USA3) Cardiff University, Wales4) University of California-Los Angeles, Los Angeles, California USA5) University of Wisconsin-Madison, Madison, Wisconsin USA
Abstract. Edge conditions in DIII-D are being quantified in order to provide insight into thephysics of the H-mode regime. Electron temperature is not the key parameter that controls theL-H transition. Gradients of edge temperature and pressure are much more promising candidatesfor such parameters. The quality of H-mode confinement is strongly correlated with the heightof the H-mode pedestal for the pressure. The gradient of the pressure appears to be controlledby MHD modes, in particular by kink-ballooning modes with finite mode number n. For a widevariety of discharges, the width of the barrier is well described with a relationship that is pro-portional to (βped
p )1/2. An attractive regime of confinement has been discovered which providessteady-state operation with no ELMs, low impurity content and normal H-mode confinement. Acoherent edge MHD-mode evidently provides adequate particle transport to control the plasmadensity and impurity content while permitting the pressure pedestal to remain almost identicalto that observed in ELMing discharges.
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217
(EXP5/22) Enhanced Particle Confinement and TurbulenceReduction Due to ExB Shear in the TEXTOR Tokamak
J. A. Boedo1), D. S. Gray1), P. W. Terry2), S. Jachmich3), G. Tynan1), R. W. Conn1), and theTEXTOR Team4)
1) University of California, San Diego, U.S.A.2) University of Winsconsin, Madison, U.S.A.3) ERM/KMS, Brussels, Belgium4) IPP FZ-Julich, Julich, Germany
Abstract. The scaling of turbulence, turbulent particle flux and cross-phase with shear ismeasured and compared with various analytical theories. It is found that the scaling can beexpressed as a second-order polynomial and that the cross-phase plays a key role in the suppres-sion of the particle flux. The variable rate of shear, kept below the value required to produce aLow-to-High particle confinement transition, was obtained by changing, in a shot to shot basis,the voltage applied to an electrode introduced 4 cm into the plasma.
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218
(EXP5/25) New Transition Phenomena in a Long Dischargeon TRIAM-1M
K. Hanada1), S. Itoh1), K. Nakamura1), H. Zushi1), M. Sakamoto1), E. Jotaki1), Y. D. Pan2),M. Hasegawa1), S. Kawasaki1), H. Nakashima1)
1) Advanced Fusion Research Center, Research Institute for Applied Mechanics, Kyushu Uni-versity, Japan2) Southwestern Institute of Physics, China
Abstract. Abstract Enhancement of current drive (ECD) efficiency mode, which is character-ized by the spontaneous increase of current drive efficiency, ηCD, from 0.3−0.4×1019A/Wm−2 to0.7−1.0×1019A/Wm−2, is observed in long pure lower hybrid current drive (LHCD) plasmas onTRIAM-1M. The energy confinement time is also improved due to the increase of line averagedelectron density, ion and electron temperatures. The current drive efficiency is proportional tothe electron density. The transition to ECD mode occurs at a critical density, which slightlydepends on the refractive index to the toroidal direction, N‖ of the injected wave.
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219
(EXP5/26) Substantial Reduction of H-mode TransitionThreshold Power in JT-60U
K. Tsuchiya1), T. Fukuda1), H. Takenaga1), N. Asakura1), Y. Kamada1), T. Takizuka1), K. Itami1),T. Fujita1) and the JT-60U Team1)
1) JAERI Naka Fusion Research Establishment, Ibaraki, Japan
Abstract. Substantial reduction of the H-mode transition threshold power was observed underthe W-shaped divertor in JT-60U, in comparison with the open divertor. Radiation power inmain region was also decreased in this case. These difference of radiation power from the mainplasma between open and W-shaped divertor was not large enough to account for the appar-ent reduction of threshold power. In this study, the edge plasma parameters which relate toL-H transition were put emphasis on. It was found that edge ion temperature at which L-Htransition occurred in the open divertor case was established by lower input power since edgedensity became smaller in the W-shaped divertor case under the condition of the same line-averaged density. After the modification of divertor geometry, ion edge collisionality just beforeL-H transition became lower and it was established with lower auxiliary power input. It wassuggested that reduction of edge ion collisionality arose from neutral particles near the X-pointby the analysis of poloidal profile of neutral density. Therefore, after the modification of divertorgeometry, threshold power decreased though edge neutral density became larger.
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220
(EXP5/27) Impurity Transport Induced Oscillations in LHD
B. J. Peterson1), Y. Nakamura1), K. Yamazaki1), N. Noda1), J. Rice2), Y. Takeiri1), M. Goto1),K. Narihara1), K. Tanaka1), K. Sato1), S. Masuzaki1), S. Sakakibara1), K. Ida1), H. Fun-aba1), M. Shoji1), M. Osakabe1), M. Sato1), Y. Xu1), T. Kobuchi3), N. Ashikawa3), P. deVries1), M. Emoto1), H. Idei1), K. Ikeda1), S. Inagaki1), N. Inoue1), M. Isobe1), S. Kado1),K. Khlopenkov1), S. Kubo1), R. Kumazawa1), T. Minami1), J. Miyazawa1), T. Morisaki1), S. Mu-rakami1), S. Muto1), T. Mutoh1), Y. Nagayama1), H. Nakanishi1), K. Nishimura1), T. Notake4),Y. Liang3), S. Ohdachi1), Y. Oka1), T. Ozaki1), R. O. Pavlichenko1), A. Sagara1), K. Saito4),R. Sakamoto1), H. Sasao3), M. Sasao1), T. Seki1), T. Shimozuma1), H. Suzuki1), M. Takechi1),N. Tamura3), K. Toi1), T. Tokuzawa1), Y. Torii4), K. Tsumori1), I. Yamada1), S. Yamaguchi1),S. Yamamoto4), M. Yokoyama1), Y. Yoshimura1), K. Y. Watanabe1), T. Watari1), K. Kawa-hata1), O. Kaneko1), N. Ohyabu1), H. Yamada1), A. Komori1), S. Sudo1), O. Motojima1)
1) National Institute for Fusion Science, Toki-shi 509-5292, JAPAN2) Plasma Science & Fusion Center, Mass. Inst. of Tech., Cambridge, MA 02139 , USA3) Dept. of Fusion Science, School of Math. and Phys. Science, Grad. Univ. for Adv. Studies,Hayama 240-0193, Japan4) Dept.of Energy Engineering and Science, Nagoya Univ., Nagoya 464-8603, Japan
Abstract. In LHD during experiments using stainless steel divertor plates a slow (∼ 1 second)cyclic oscillation in the plasma parameters known as ‘breathing’ plasma was observed duringNBI-heated long pulse discharges. The core iron density, calculated from measured parametersusing the cooling rate from a corona-equilibrium average-ion model, oscillates out of phase withthe electron density. The correlation of the iron impurity concentration with the change inelectron temperature and with the local power balance between radiation and beam depositionindicates that when radiation from the iron impurity dominates the local power balance thecore plasma is cooled. The increase in the calculated iron density during the phase of the os-cillation when the divertor electron temperature exceeds the sputtering threshold suggests thatsputtering of the stainless steel divertor plate may be the source of the iron impurity. Howeverthe shortness of the delay time raises questions about the causality between these two signalsand points to the need for a closer examination of the role of impurity transport in this oscillation.
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221
(EXP5/28) Shape Invariance and Evolution of the ElectronTemperature Profile in LHD
K. Narihara1), I. Yamada1), N. Ohyabu1), K. Y. Watanabe1), N. Ashikawa2), P. de Vries1),M. Emoto1), H. Funaba1), M. Goto1), K. Ichiguchi1), K. Ida1), H. Idei1), K. Ikeda1), S. In-agaki1), N. Inoue1), M. Isobe1), S. Kado1), O. Kanako1), K. Kawahata1), K. Khlopenkov1),T. Kobuchi2), A. Komori1), S. Kubo1), R. Kumazawa1), Y. Liang2), T. Masuzaki1), T. Mi-nami1), J. Miyazawa1), T. Morisaki1), S. Morita1), S. Murakami1), S. Muto1), T. Mutoh1),Y. Nagayama1), Y. Nakamura1), H. Nakanishi1), Y. Narushima1), K. Nishimura1), N. Noda1),T. Notake3), S. Ohdachi1), Y. Oka1), M. Osakabe1), S. Ozaki1), R. O. Pavlichenko1), B. J. Peter-son1), A. Sagara1), K. Saito3), S. Sakakibara1), R. Sakamoto1), H. Sasao2), M. Sasao1), K. Sato1),M. Sato1), T. Seki1), T. Shimozuma1), C. Shoji1), H. Suzuki1), A. Takayama1), M. Takechi3),Y. Takeiri1), N. Tamura2), K. Tanaka1), K. Toi1), N. Tokuzawa1), Y. Torii3), K. Tsumori1),T. Watari1), H. Yamada1), S. Yamaguchi1), S. Yamamoto3), M. Yokoyama1), Y. Yoshimura1),S. Satow1), K. Itoh1), K. Ohkubo1), K. Yamazaki1), S. Sudo1), O. Motojima1), Y. Hamada1),M. Fujiwara1)
1) National Institute for Fusion Science, Toki, 592-5292, Japan2) Graduate University for Advanced Studies, Hayama, 240-0193, Japan3) Department of Energy Engineering and Science, Nagoya University 464-8603, Japan
Abstract. With a multi-point (200) repetitive (50–200 Hz) Thomson scattering system, westudied the shape and evolution of the electron temperature (Te) profiles of the plasma confinedin LHD. We first survey various shapes of the observed Te profiles and then describe two notablefindings in some detail: (1) A pedestal often appeared on Te profiles around the ι/2π = 1 surface,but its correlation with confinement was weak; (2) A magnetic island generated by an externalerror field changed its size in plasma. Normally the island shrank in plasma, but grew upon ahydrogen pellet injection.
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222
(EXP5/29) Transition Dynamics and Confinement Scalingin COMPASS-D H-mode Plasmas
S. J. Fielding1), P. G. Carolan1), J. W. Connor1), N. J. Conway1), A. R. Field1), P. Helander1),Yu. Igitkhanov2), B. Lloyd1), H. Meyer1), A. W. Morris1), O. Pogutse1), M. Valovic1), H. R. Wil-son1), COMPASS-D Team1), COMPASS-ECRH Team1)
1) EURATOM/UKAEA Fusion Association, Culham Science Centre, Abingdon, Oxfordshire,OX14 3DB, UK2) ITER Joint Central Team, Joint Work-Site, D-85748 Garching, Germany
Abstract. This paper describes experimental analysis of different phases of ECRH H-modedischarges on COMPASS-D, from L-H transition trigger to steady state. Comprehensive high-resolution measurements of the transport barrier region have enabled significant progress to bemade in assessing H-mode trigger mechanisms, as well as clarifying the evolution of the localelectric field and its shear. Stationary ELMy H-modes are achieved at high ρ∗ with good con-finement, and dimensionless scaling over a range of ν∗ is presented.
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223
(EXP5/30) ELMing H-mode Accessibility in Shaped TCVPlasmas
Y. R. Martin1), TCV Team1)
1) CRPP / EPFL, CH-1015 Lausanne, Switzerland
Abstract. The H-mode regime is easily reached with ohmic heating only in a wide range ofTCV plasma parameters. However, the plasma usually enters an ELM free H-mode phase afterthe LH transition, leading to a high density disruption. Therefore, the access to a stable ELMyregime requires an LH transition directly leading to an ELMy phase. A “gateway” to the ELMyregime was found in TCV ohmic discharges. Although small in terms of plasma control param-eter ranges, this “gateway” is well defined and robust against changes in wall conditioning forinstance. Once in the ELMing regime, the plasma was successfully driven in a much wider rangeof plasma parameters, whilst remaining in this better confinement mode.
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224
(EXP5/31) Transition to H-mode Regime in JET with andwithout Pumped Divertors
E. Righi1), the JET Team1)
1) see paper OV1/2
Abstract. In the present paper the JET threshold database from 1990 to 1999 is analysed.Sources of data scatter, which introduce considerable uncertainty in the estimate of the powerneeded for the H-mode transition in ITER, are identified and eliminated. The influence of diver-tor geometry and plasma configuration on the power threshold is analysed in detail. In particular,the height of the X-point over the divertor plate (or septum) and the plasma-first wall distanceare considered. The data show consistency throughout the ten years of JET operation, andconfirm that the power threshold is minimised with decreasing X-point height, while it increasessharply if the plasma is pushed too near the outer wall. Increased threshold due to operation onthe vertical target plates can be also explained in terms of X-point height. Although geometryinfluences the scatter in the database obtained with single divertors, it is found to be negligible ifthe whole database is considered. More important seems to be the trend towards stronger den-sity dependence with increasing divertor closure. If this trend is taken into account, a significantreduction in the database scatter is obtained and data from different divertors become consistent.
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225
(EXP5/32) Measurement of Impurity Transport Coefficientsin the Confined Plasma of ASDEX Upgrade
R. Dux1), A. G. Peeters1), A. Gude1), A. Kallenbach1), R. Neu1), ASDEX Upgrade Team1)
Abstract. For ASDEX Upgrade H-Mode discharges, the core plasma impurity transport hasbeen investigated in the quiet phase between sawtooth crashes. For the elements Ne, Ar, Krand Xe the diffusion coefficient in the center is D ≤ 6 × 10−2m2/s and rises with the radialdistance from the center. With increasing Z number the transport becomes strongly convectivewith inwardly directed drift velocities that produce very peaked impurity densities for high Z.The calculated neoclassical diffusion coefficient and drift velocity are close to the experimentalvalues for the lower Z elements Ne and Ar. The calculated drift velocity is too small by a factorof 10 for Kr and Xe. For these elements, toroidal rotation of the plasma leads to an increasedimpurity density on the outboard side of the flux surfaces which is not taken into account bythe neoclassical calculations. The outboard/inboard ratio for Kr is ≈ 1.5 and the toroidal Machnumber Mtor ≈ 2. Investigations of Si and Ne transport in the edge region, i.e. the region of thetemperature pedestal and the steep temperature gradient zone, yield an inwardly directed driftvelocity in the quiet phase between ELMs. The inward pinch is observed in the radial range ofthe steep temperature gradient. The transport induced by an ELM is best described by a veryhigh radial diffusion coefficient rather than by an outwardly directed drift velocity.
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226
(EXP5/33) Propagation of Cold Pulses and Heat Pulses inASDEX Upgrade
R. Neu1), F. Ryter1), R. Dux1), H.-U. Fahrbach1), A. Jacchia2), J. E. Kinsey3), F. Leuterer1),F. De Luca4), G. V. Pereverzev1), J. Stober1), W. Suttrop1), ASDEX Upgrade Team1)
1) Max-Planck-Institut fur Plasmaphysik, EURATOM Association, Garching, Germany2) Istituto di Fisica del Plasma, Associazione EURATOM-ENEA-CNR, Milan, Italy3) General Atomics, San Diego, California, United States of America4) INFM and Dipartimento di Fisica, Universita degli Studi di Milano, Milan, Italy
Abstract. Experiments on electron heat transport were performed in the tokamak ASDEXUpgrade, mainly in ohmically heated plasmas, applying either edge cooling by impurity injec-tion or edge heat pulses with ECH. Repetitive pulses within one plasma discharge were madeallowing Fourier transformation of the temperature perturbation. This yields a good signal tonoise ratio up to high harmonics and allows a detailed investigation of the pulse propagation.For densities lower than 1.8 × 1019m−3, an increase of the central electron temperature wasfound as the response to the edge cooling via impurity injection similar to observations made inother tokamaks. The inversion does not appear instantaneously, but with a time delay roughlycompatible with diffusion. Modeling of the propagation of the cold pulses in the frameworkof the IFS-PPPL model yields qualitative agreement. However the predicted increase of theion temperature is not observed experimentally on the fast time scale. The response to ECHheat pulses is not perfectly symmetrical to cold pulse experiments, but the similarities suggesta common underlying physical mechanism. No inversion of the heat pulse is found, instead theinitial pulse from the edge is associated with a second, much slower heat pulse in the centrewhich is similar (and not symmetrical) to that of the cold pulses. It is found that the central in-crease is related to the arrival of the pulse close to the inversion radius and not to the initial pulse.
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227
(EXP5/34) Development of Fast Helium Beam EmissionSpectroscopy for Plasma Density and Temperature Diag-nostics
M. Proschek1), S. Menhart1), H. D. Falter1), H. Anderson2), H. P. Summers2), A. Stabler3),P. Franzen3), H. Meister3), J. Schweinzer5), T. T. C. Jones4), S. Cox4), N. Hawkes4), F. Au-mayr1), H. P. Winter1)
1) Institut f. Allgemeine Physik, TU-Wien, A-1040 Wien, Wiedner Hauptstrasse 8-10 (Asso-ciation EURATOM-OEAW)2) Dept. of Physics and Applied Physics, Univ. of Strathclyde, Glasgow, UK3) Max-Planck-Institut f. Plasmaphysik, D-85748 Garching4) UKAEA-Euratom Association, Culham Science Centre, Abingdon, OXON, OX14 3EA, UK5) EFDA-JET CSU, Building K1, Culham Science Centre, Abingdon, OXON, OX14 3EA, UK.
Abstract. For developing a novel electron density and -temperature diagnostics based on fastHe beam emission spectroscopy, experiments have been performed at the ASDEX Upgrade toka-mak (AUG) in Garching and the JET tokamak in Culham. The measured He I emission profileswere compared with model calculations which are based on a collisional-radiative model devel-oped by the ADAS group. For exploratory measurements at AUG one of the heating beamsources has been operated with pure helium. The beam emission profiles show satisfactoryagreement with the profiles modelled using density and temperature profiles from other diag-nostics. At JET and recently at AUG a small amount of helium was added to one standarddeuterium ion source in order to produce a “doped” helium/deuterium beam. The respectivemeasurements were performed using groups of identical pulses. In total, 11 different He I lineswere investigated at JET with respect to their dependence on plasma density and -temperature.Seven lines were found to have sufficient intensity but the beam emission profile suffers fromlimited bandwidth of the spectrometer used. Good beam emission profiles could be obtainedfrom recent AUG measurements showing a scatter of 9%.
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228
Session ITERP — ITER
Contents
(ITERP/01) Neutral Beam Heating and Current Drive System and its Rolein ITER-FEAT Operation Scenarios . . . . . . . . . . . . . . . . . . . . . 230
(ITERP/02) ITER-FEAT Magnetic Configuration and Plasma Position/ShapeControl in the Nominal PF Scenario . . . . . . . . . . . . . . . . . . . . . 231
(ITERP/01) Neutral Beam Heating and Current Drive Sys-tem and its Role in ITER-FEAT Operation Scenarios
N. Fujisawa1), T. Inoue1), E. Di Pietro1), P. L. Mondino1), Y. Murakami1), M. Shimada1)
1) ITER Joint Central Team
Abstract. The NB H&CD system, providing 33 MW in deuterium beams at 1 MeV fromtwo injectors, in addition to 40 MW RF power, contributes to heating a plasma to sub-ignitionthrough the L-H mode transition followed by finite-Q driven-burn (Q ≥ 10), and achievementof a hybrid operation with an extended-duration (∼ 1000 s) or steady-state operation with Q≥ 5. To achieve such operations, the NB provides non-inductive current drive by injecting thebeams tangentially into the plasma with the capability of on- and off-axis current drive. Thepresent engineering design is under the constraints of the beam envelope, vacuum confinement,neutron shielding, tolerances, and clearances required with the toroidal field coils. The on- andoff-axis current drive is to be achieved by tilting the beam axis vertically. Each beam axis ofthe NB injectors can be tilted independently, providing flexibility in the control of heating andthe driven current profile.
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230
(ITERP/02) ITER-FEAT Magnetic Configuration and PlasmaPosition/Shape Control in the Nominal PF Scenario
Y. V. Gribov1), R. Albanese2), G. Ambrosino2), M. Ariola2), R. H. Bulmer3), M. Cavinato1),E. Coccorese2), H. Fujieda4), A. Kavin1), R. Khayrutdinov5), K. Lobanov6), V. Lukash7),L. Makarova6), A. Mineev6), P. L. Mondino1), A. Pironti2), A. Portone8), E. Rumyantsev6),I. Senda4), T. Shoji4), V. Vasiliev6)
1) ITER Joint Central Team, Naka Joint Work Site, Naka-machi, Ibaraki-ken, Japan2) Association EURATOM/ENEA/CREATE, Reggio Calabria, Italy3) LLNL, Livermore, California, USA4) JAERI Naka Fusion Research Establishment, Naka-machi, Ibaraki-ken, Japan5) TRINITI, Troitsk, Moscow Region, Russia6) Efremov Institute, St. Petersburg, Russia7) Kurchatov Institute, Moscow, Russia8) EFDA, CSU, Max Planck Institute for Plasma Physics, Garching, Germany
Abstract. The capability of the ITER-FEAT poloidal field system to support the four “de-sign” scenarios and the high current “assessed” scenario have been studied. To operate withhighly elongated plasma, the system has segmentation of the central solenoid and a separate fastfeedback loop for plasma vertical stabilisation. Within the limits imposed on the coil currents,voltages and power, the poloidal field system provides the required plasma scenario and controlcapabilities. The separatrix deviation from the required position, in scenarios with minor dis-ruptions is within less than about 100 mm.
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231
(ITERP/03) Understanding of the H-mode Pedestal Char-acteristics Using the Multi-machine Pedestal Database
T. Hatae1), M. Sugihara2), A. Hubbard3), Yu. Igitkhanov2), Y. Kamada1), G. Janeschitz2),L. D. Horton4), N. Ohyabu5), T. H. Osborne6), M. Osipenko7), W. Suttrop4), H. Urano8),H. Weisen9)
1) Japan Atomic Energy Research Institute, Naka-machi, Ibaraki-ken, Japan2) ITER Joint Central Team, Garching, Germany3) MIT Plasma Science and Fusion Center, Cambridge, MA, USA4) Max Planck Institut fur Plasmaphysik, Garching, Germany5) National Institute for Fusion Science, Gifu-ken, Japan6) General Atomics, San Diego, CA, USA7) Kurchatov Institute, Moscow, Russia8) Hokkaido University, Hokkaido, Japan9) Centre de Recherches en Physique des Plasmas, Ecole Polytechnique Federale de Lausanne,Switzerland
Abstract. With the use of a multi-machine pedestal database, essential issues for each regime ofELM types are investigated. They include (i) understanding and prediction of pedestal pressureduring Type I-ELMs which is a reference operation mode of a future tokamak reactor, (ii) iden-tification of the operation regime of Type-II ELMs which have small ELM amplitude with goodconfinement characteristics, (iii) identification of upper stability boundary of Type-III ELMs foraccess to the higher confinement regimes with Type-I or -II ELMs, (iv) relation between coreconfinement and pedestal temperature in conjunction with the confinement degradation in highdensity discharges. Scaling and model-based approaches for expressing pedestal pressure areshown to roughly scale the experimental data equally well and initial predictions for a futurereactor case could be performed by them. It is identified that q and δ are important parametersto obtain the Type-II ELM regime. A theoretical model of Type-III ELMs is shown to reproducethe upper stability boundary reasonably well. It is shown that there exists a critical pedestaltemperature, below which the core confinement starts to degrade.
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232
(ITERP/04) Next Step Tokamak Physics: Confinement-oriented Global Database Analysis
O. J. Kardaun1), A. Kallenbach1), P. McCarthy1)2), F. Ryter1), A. Stabler1), J. Stober1),W. Suttrop1), M. Valovic3), S. J. Fielding3), G. Bracco4), J. Cordey5), D. C. McDonald5),D. Hogeweij6), A. Sykes3), A. Dnestrovskij3)7), M. Valovic3), M. Walsh3), Y. R. Martin8), J. On-gena9), G. T. Hoang10), T. Aniel10), K. Thomsen5)11), S. V. Lebedev12), V. A. Kornev12),Yu. V. Esipchuk7), V. M. Leonov7), V. Vershkov7), M. J. Greenwald13), A. Hubbard13), J. C. De-Boo14), A. Cote15), G. Pacher15), C. Bush16)17), S. Kaye17), T. Takizuka18), T. Fukuda18),Y. Kamada18), K. Tsuchiya18), Y. Miura19), N. Isei19), K. Tsuzuki19), A. Chudnovskij20)7),V. Mukhovatov20)7)
1) MPI fur Plasmaphysik, Garching, Germany2) University College Cork, Cork, Ireland3) UKAEA, Culham, United Kingdom4) ENEA, Frascati, Italy5) JET Joint Undertaking, Abingdon, United Kingdom6) FOM Instituut voor Plasmafysica, Jutphaas, The Netherlands7) Kurchatov Institute, Moscow, Russia8) CRPP, Lausanne, Switzerland9) IPP Forschungszentrum, Juelich, Germany10) CEA, Cadarache, France11) EFDA Close Support Unit, Garching, Germany12) Ioffe Institute, St. Petersburg, Russia13) MIT, Cambridge, USA14) General Atomics, San Diego, USA15) CCFM Hydro-Quebec, Varennes, Canada16) ORNL, Oak Ridge, USA17) PPPL, Princeton, USA18) JAERI, Naka Fusion Research Establishment, Naka, Japan19) JAERI, Tokai Research Establishment, Naka, Japan20) ITER Joint Central Team, Naka site, Japan
Abstract. We describe and analyse an international multi-tokamak confinement database, bothmotivated by physics and with a view toward prediction of next-step burning-plasma experimentssuch as ITER. Significant additional ohmic and L-mode data have been assembled from severaltokamaks, which has resulted in the ‘ITERL.DB2’ dataset. Simple density-roll-over scalings arepresented for ohmic confinement. For H-mode, the confinement time in the essentially enlargeddata set ITERH.DB3 is compared with the ITERH-98P(y,2) reference scaling. A distinction ismade between discharges with and without heavy gaspuff. Beyond a standard power-law scaling,the empirical ‘influence’ on confinement of q95/qcyl, directly related to triangularity, and of theglobal density peaking factor (for L- and H-mode) is quantified. A log-linear quadratic formulais given which describes physically more precisely than ITERH-98P(y,2) the relation betweenthe isotope effect and the heating power degradation of confinement, while predicting a similarthermal confinement time for ITER (τE,th ' 3.5s). Based on a recently provided plasma edgedataset, ‘E.1’, separate scalings of the plasma core and pedestal energy are derived. Finally,a class of nonlinear scalings is discussed which are suitable, in contrast to offset (non-)linearmodels, to fit roll-over dependence, and, simultaneously, the scaling of L-mode and H-modeconfinement.
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233
(ITERP/05) Physics Basis of ITER-FEAT
M. Shimada1), D. J. Campbell2), M. Wakatani3), H. Ninomiya4), N. V. Ivanov5), V. Mukhova-tov1) and the ITER Joint Central Team and Home Teams
1) ITER Joint Central Team, Naka Joint Work Site, Naka-machi, Ibaraki-ken, Japan2) EFDA Close Support Unit, Garching, Germany3) Kyoto University, Kyoto, Japan4) Japan Atomic Energy Research Institute, Naka-machi, Ibaraki-ken, Japan5) Kurchatov Institute, Moscow, Russian Federation
Abstract. This paper reviews Physics R&D results obtained since the publication of the ITERPhysics Basis document. The heating power required for the LH transition has been re-assessed,including recent results from C-Mod and JT-60U and it has been found that the predicted poweris a factor of two lower than the previous projection. For predicting ITER-FEAT performance, aconservative scaling IPB98(y,2) has been adopted for the energy confinement, producing confine-ment times ∼ 20% lower than those derived from the IPB98(y,1) law. While energy confinementdegradation at high density remains a serious issue, recent experiments suggest that good con-finement is achievable in ITER at n/nG ∼ 0.85 with high triangularity. The estimated runawayelectron energy has been reduced to ∼ 20MJ, since recent experiments show that runaway elec-trons disappear for q95leq2.
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234
(ITERP/06) Performance Assessment of ITER-FEAT
Y. Murakami1), I. Senda1), H. Matsumoto2), M. Shimada2), A. Chudnovskij2), A. Polevoi2),G. Vayakis2), O. J. Kardaun3), V. M. Leonov4)
1) Naka Fusion Research Establishment, JAERI, Naka, Ibaraki, Japan2) ITER Joint Central Team, Naka Joint Work Site, Naka, Ibaraki, Japan3) IPP Garching, Max-Planck-Institut fur Plasmaphysik, Germany4) RRC, Kurchatov Institute, Moscow, Russia
Abstract. A performance assessment for ELMy H-mode operation of ITER-FEAT mainly atthe nominal plasma current of 15 MA is made by using 1.5D transport codes PRETOR and AS-TRA. Operation domain analysis is performed for various transport assumptions. Sensitivitiesto density profile, the ratio of ion thermal diffusivity to electron thermal diffusivity χi/χe and theion heating fraction are investigated. It is shown that, under rather conservative assumptions,400 MW operation with fusion gain Q = 10 should be achievable. Operations with lower andhigher fusion power are explored and an operation range of 200 ∼ 600MW is obtained. A proba-bilistic performance assessment is also done by using 0D modeling. The “maximized conditionalprobability (MCP)” to reach Q larger than a specified lower bound is estimated considering thebeta limit βN ≤ 2.5, L-H transition threshold power and density limit ne/nGR ≤ 0.85. The MCPof achieving Q ≥ 10 is about 70%, and the MCP of Q ≥ 50 is about 30% when the HH factordistribution is a Gaussian with σ = 20%. By increasing the plasma current to 17 MA, the MCPsof achieving Q ≥ 10 and Q ≥ 50 increase to 85% and 50%, respectively.
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235
(ITERP/07) Theory of Neoclassical Tearing Modes and itsApplication to ITER-FEAT
V. D. Pustovitov1), A. B. Mikhailovskii1), N. Kobayashi2), S. V. Konovalov1), V. S. Mukhova-tov3), A. V. Zvonkov1)
1) Russian Research Centre “Kurchatov Institute”, Moscow, Russia2) ITER Garching Joint Work Site, Garching, Germany3) ITER Naka Joint Work Site, Naka-machi, Naka-gun, Ibaraki-ken, Japan
Abstract. Neoclassical tearing modes (NTM) can be responsible for beta limitation in long-pulse ITER discharges. The excitation and growth of NTM are governed by the competingbootstrap current, polarization current and so-called ∆′ effects. Also, the magnetic well andElectron Cyclotron Current Drive (ECCD) can stabilize the NTM. We study analytically andnumerically all the effects with a particular emphasis on the polarization current in the analyt-ical part of our study. We show that the polarization current description requires a generalizedtransport theory including the hyperviscosity, electron pressure gradient and, as well, the finiteion Larmor radius effects in the perpendicular current. The profile function nonstationaritymust be taken into account for calculation of the island rotation frequency. Results of numericalsimulation of NTM suppression by modulated ECCD in ITER are presented.
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236
(ITERP/08) High-n Ideal TAE Stability of ITER
G. Vlad1), F. Zonca
1) Associazione EURATOM-ENEA sulla Fusione, Frascati, Rome, Italy
Abstract. The ideal MHD stability analysis of high-n (n is the toroidal mode number) ToroidalAlfven Eigenmodes (TAE’s) is presented, using realistic and completely general ITER equilibriawith shaped, up-down asymmetric, magnetic flux surfaces. An approach has been used, basedon analytical-theoretical methods, which can give interesting results and allows us to analyzethe conditions for enhanced TAE damping (although preventing us from computing the exci-tation thresholds). The frequency spectrum of TAE modes is found by solving the fully twodimensional problem using a two spatial-scale WKB formalism. The phase space integration isextended to a complete periodic orbit (at fixed frequency ω) in the (r, θk) phase-space (r is herea general flux coordinate and θk is the WKB eikonal entering in the expression of the radialenvelope of the mode). The equilibria, analyzed here, are characterized by ideal TAE’s localizedin the half outer part of the plasma column, where the α-particle drive is expected to be smalland modes are likely affected by continuum damping.
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237
(ITERP/09) Measurement Requirements and DiagnosticSystem Designs for ITER-FEAT
A. J. H. Donne1), A. E. Costley2), K. Ebisawa2), G. Janeschitz3), S. Kasai4), A. Kislyakov5),A. V. Krasilnikov6), Y. Kusama4), A. Malaquias3), P. Nielsen7), F. P. Orsitto8), V. Strelkov9),T. Sugie4), G. Vayakis2), C. Walker3), S. Yamamoto3), V. Zaveriaev9), H. Zushi10)
1) FOM-Instituut voor Plasmafysica ‘Rijnhuizen’, Nieuwegein, The Netherlands2) ITER JCT, Naka Joint Working Site, Japan3) ITER JCT, Garching Joint Working Site, Germany4) JAERI, Naka, Japan5) Ioffe Institute, St. Petersburg, Russia6) TRINITY, Troitsk, Russia7) Consorzio RFX, Padova, Italy8) JET-EFDA/CSU, Abingdon, England9) Kurchatov Institute, Moscow, Russia10) Kyushu University, Fukuoka, Japan
Abstract. The requirements for plasma measurements necessary to support the differentplanned operating scenarios of ITER-FEAT are presented. It is found that as the plasmaperformance becomes more enhanced the requirements for plasma measurements become moredemanding. The measurements will be made with a comprehensive diagnostic system and thisis briefly described.
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238
(ITERP/10(R)) Basic Divertor Operation in ITER-FEAT
A. S. Kukushkin1), G. Janeschitz1), A. Loarte2), H. D. Pacher3), D. P. Coster4), G. Matthews5),D. Reiter6), R. Schneider7), V. Zhogolev8)
1) 1) ITER JCT, Boltzmannstr. 2, 85748 Garching, Germany2) EFDA, Close Support Unit, Garching, Germany3) INRS-Energie et Materiaux, Varennes, Quebec, Canada4) Max-Planck-Institut fur Plasmaphysik, Garching, Germany5) JET Joint Undertaking, Abingdon, United Kingdom6) FZ Julich, Julich, Germany7) Max-Planck-Institut fur Plasmaphysik, Greifswald, Germany8) RSC “Kurchatov Institute”, Moscow, Russia
Abstract. This paper summarises the modelling studies of steady-state divertor operation be-ing performed for the ITER-FEAT design. Optimisation of the divertor geometry reveals theimportance of the proper target shape for a reduction of the peak power loads. A high gasconductance between the divertor legs is also essential for maintaining acceptable conditions inthe outer divertor which receives higher power loading than the inner. Impurity seeding, whichwould be necessary if tritium co-deposition concerns preclude the use of carbon as plasma-facingmaterial, can ensure the required high radiation level at acceptable Zeff , and the divertor per-formance is not very sensitive to the choice of the radiating impurity.
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239
(ITERP/11(R)) Predicted ELM Energy Loss and PowerLoading in ITER-FEAT
A. Loarte1), G. Saibene1), R. Sartori1), G. Janeschitz2), Yu. Igitkhanov2), A. S. Kukushkin2),M. Sugihara2), D. P. Coster3), A. Herrmann3), L. D. Horton3), J. Stober3), N. Asakura4),K. Itami4), H. Tamai4), G. Matthews5), R. Schneider6), D. Reiter7), A. W. Leonard8), G. D. Porter9)
1) EFDA CSU-Garching, MPI fur Plasmaphysik, D-85748 Garching bei Munchen, Germany2) ITER Joint Central Team, MPI fur Plasmaphysik, D-85748 Garching bei Munchen, Germany3) MPI fur Plasmaphysik, D-85748 Garching bei Munchen, Germany4) JAERI, Naka Research Establishment, Naka-machi, Naka-gun, Ibaraki-ken 311-0193, Japan5) Joint European Torus, UKAEA Culham Ass., Abingdon OX14 3EA, United Kingdom6) MPI fur Plasmaphysik, D-17489 Greifswald, Germany7) Institut fur Laser und Plasmaphysik, Heinrich-Heine-Universitat Dusseldorf, D-40225 Dusseldorf,Germany8) General Atomics, P.O. Box 85608, San Diego, CA 92186-5608, USA9) Lawrence Livermore National Laboratory, P.O. Box 808, Livermore, CA 94550, USA
Abstract. Scaling of Type I ELM energy losses from existing experiments to the ITER-FEATreference inductive scenario shows that the ELM energy loss is ∼ 12 MJ, marginal from divertorlifetime considerations. B2-Eirene modelling of the transient radiative losses induced by theELMs shows a factor of 2 reduction of the divertor load for small ELMs but not for the expected∼ 12 MJ ELMs. Regimes with reduced ELM sizes (Type II) compatible with the ITER-FEATreference performance would be required to achieve a long lifetime of the divertor target.
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240
(ITERP/12) ITER-FEAT Fuel Cycle
P. Ladd1), A. Antipenkov1), A. Busigin2), M. Glugla3), S. Konishi4), D. K. Murdoch5), M. Nishi4),H. Yoshida6)
1) ITER, Max-Planck-Institut fur Plasmaphysik, Boltzmannstr. 2, 85748 Garching, Germany2) NITEK Corporation, 38 Longview CRT, London ON N6K 4J1, Canada3) Forschungszentrum Karlsruhe, Postfach 3640, D-76021 Karlsruhe, Germany4) Japan Atomic Energy Research Institute, Tokai, Naka-gun, Ibaraki, 319-1195, Japan5) EFDA-CSU, Max-Planck-Institut fur Plasmaphysik, Boltzmannstr. 2, 85748 Garching, Ger-many6) ITER, 801-1 Mukouyama, Naka-machi, Naka-gun, Ibaraki-ken, 311-0193 Japan
Abstract. The Fuel Cycle, which includes plasma fuelling and exhaust, as well as exhaustprocessing and isotope separation, is one of the key elements on which the successful operationof ITER will depend. This paper provides an overview of this system, reviewing requirements,operational scenarios, and the integration of the various subsystems using the ITER fuel cycledynamic simulation program CFTSIM. The requirements to provide a plasma fuelling rate of200 Pam3s−1, with a flat-top burn of ∼400s and a repetition rate of two pulses per hour havethe greatest influence on the design. However, while a flat-top burn of ∼400s is the initial de-sign basis, the capability to extend the pulse to 3,000s in the longer term is essential from anoperational perspective.
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241
(ITERP/13) ITER-FEAT Vacuum Pumping and FuellingR&D Programmes
D. K. Murdoch1), A. Antipenkov2), J.-C. Boissin1), C. Day3), H. Haas3), P. Ladd2), A. Mack3),S. Pimanikhin4), G. Saksagansky5), I. Viniar6)
1) EU-HT, EFDA-CSU, Boltzmannstraße 2, D-85748 Garching, Germany2) ITER JCT, Garching, Boltzmannstraße 2, D-85748 Garching, Germany3) EU-HT, Forschungszentrum Karlsruhe, Postfach 3640, D-76021 Karlsruhe, Germany4) RF-TH, All-Russian Scientific Research Institute of Experimental Physics, Sarov, 607190,Russia5) RF, D.V. Efremov Scientific Research Institute of Electrophysical Apparatus, St. Petersburg,189631, Russia6) RF-HT, PELIN Laboratory, Ltd., 2, Admiral Makarov Str., Moscow, 125212, Russia
Abstract. The design of the ITER-FEAT vacuum pumping and fuelling systems is supportedby two key R&D programs, the first directed towards the development of a steady state tritiumcompatible pellet injector, and the second towards the development of a supercritical heliumcooled cryogenic pump for torus exhaust. While the pellet injector programme for ITER-FEATis new, that for the cryopump has evolved from a programme that originally supported the1998 ITER design. As the plasma exhaust parameters have remained essentially unchangedbetween these two machines, the R&D conducted to date remains valid. Initial test results onthe prototype injector, TPI-1, which included continuous injection of 3 mm hydrogen pellets at500 m/s and at 1 to 2 Hz for periods up to, are reported. A model of the cryopump has nowbeen installed in a new dedicated test bed at the Karlsruhe Research Centre where acceptancetests have been completed and preliminary results from pumping tests obtained. An extensivetest campaign to fully characterise pump performance and identify any mechanical details whichrequire modification has started.
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242
(ITERP/14) Design of ITER-FEAT RF Heating and Cur-rent Drive Systems
G. Bosia1), N. Kobayashi1), K. Ioki1), P. Bibet2), R. Koch3), R. Chavan4), M. Q. Tran4), K. Taka-hashi5), S. Kuzikov6), V. Vdovin7)
1) ITER Joint Central Team, Boltzmannstrasse 2, 85748 Garching,2) DRFC/STID/FCI, CE Cadarache BP n1, 13108 St Paul Lez Durance,3) Plasma Physics Laboratory, ERM/KMS, 30 Av. de la Renaissance, 1040 Brussels,4) CRPP, Ass. Euratom – Conf. Suisse, PPB, EPFL 1015 Lausanne,5) RF Heating Laboratory, JAERI Naka-machi, Naka-gun, Ibaraki-ken 311-01 JAPAN,6) Institut of Applied Physics, Nizhny Novograd,7) Kurchatov Institute for Atomic Energy, Moscow
Abstract. Three radio frequency (RF) heating and current drive (H & CD) systems are beingdesigned for ITER-FEAT: an electron cyclotron (EC), an ion cyclotron (IC) and a lower hybrid(LH) System. The launchers of the RF systems use four ITER equatorial ports and are fullyinterchangeable. They feature equal power outputs (20 MW/port), similar neutron shieldingperformance, and identical interfaces with the other machine components. An outline of thedesign is given in the paper.
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243
(ITERP/15) Adaptation of the ITER Facility Design to aCanadian Site
S. Smith1)
1) ITER Canada
Abstract. This paper presents the status of Canadian efforts to adapt the newly revised ITERfacility design to suit the specific characteristics of the proposed Canadian site located in Clar-ington, west of Toronto, Ontario. ITER Canada formed a site-specific design team in 1999,comprising participants from three Canadian consulting companies to undertake this work. Thetechnical aspects of this design activity includes: construction planning, geotechnical investiga-tions, plant layout, heat sink design, electrical system interface, site-specific modifications andtie-ins, seismic design, and radwaste management. These areas are each addressed in this paper.
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244
(ITERP/16) ITER-FEAT Safety
C. W. Gordon1), H.-W. Bartels1), T. Honda1), M. Iseli2), J. Raeder1), L. Topilski1), K. Moshonas3),N. Taylor4), W. Gulden5), B. Kolbasov6), T. Inabe7), E. Tada7)
1) ITER Joint Central Team, Garching, Germany2) ITER Joint Central Team, Naka, Japan3) ITER Canada, Toronto, Canada4) UKAEA, Culham, United Kingdom5) EFDA, Garching, Germany6) Kurchatov Institute, Moscow, Russian Federation7) JAERI, Naka, Japan
Abstract. Safety has been an integral part of the design process for ITER since the ConceptualDesign Activities of the project. The safety approach adopted in the ITER-FEAT design andthe complementary assessments underway, to be documented in the Generic Site Safety Report(GSSR), are expected to help demonstrate the attractiveness of fusion and thereby set a goodprecedent for future fusion power reactors. The assessments address ITER’s radiological hazardstaking into account fusion’s favourable safety characteristics. The expectation that ITER willneed regulatory approval has influenced the entire safety design and assessment approach. Thispaper summarises the ITER-FEAT safety approach and assessments underway.
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245
(ITERP/17) Progress and Achievements on the R&D Ac-tivities for ITER Vacuum Vessel
M. Nakahira1), K. Koizumi1), H. Takahashi1), M. Onozuka2), K. Ioki2), E. Kuzumin3), V. Krylov3),J. Maslakowski4), B. E. Nelson4), L. Jones5), W. Danner5),D. Maisonnier5)
1) JA Home Team, Japan Atomic Energy Research Institute (JAERI), Ibaraki, Japan2) ITER Joint Central Team, Garching Joint Work Site, Garching, Germany3) RF Home Team, Efremov Institute, St. Petersburg, Russian Federation4) Oak Ridge National Laboratory, USA5) EU Home Team, EFDA, Max-Planck Institute, Garching, Germany
Abstract. The ITER vacuum vessel (VV) is designed to be large double-walled structure witha D-shaped cross-section. The achievable fabrication tolerance of this structure was unknowndue to the size and complexity of shape. The Full-scale Sector Model of ITER Vacuum Vessel,which was 15m in height, was fabricated and tested to obtain the fabrication and assembly toler-ances. The model was fabricated within the target tolerance of ±5mm and welding deformationduring assembly operation was obtained. The port structure was also connected using remotizedwelding tools to demonstrate the basic maintenance activity. In parallel, the tests of advancedwelding, cutting and inspection system were performed to improve the efficiency of fabricationand maintenance of the Vacuum Vessel. These activities show the feasibility of ITER VacuumVessel as feasible in a realistic way. This paper describes the major progress, achievement andlatest status of the R&D activities on the ITER vacuum vessel.
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246
(ITERP/18) Progress of and Future Plans for the L-4 Blan-ket Project
W. Daenner1), K. Ioki2), A. Cardella2), Y. Ohara3), Y. Strebkov4), L. Jones1), P. Lorenzetto1),D. Maisonnier1), M. Merola1), A. Peacock1), E. Rodgers1), M. Enoeda3), S. Satoh3), T. Hatano3),K. Furuya3), T. Kuroda3), H. Kawamura3), M. Nakamichi3), K. Tsuchiya3), H. Yamada3),B. Rodchenkov4), K. Skladnov4), G. Sysoev4)
1) EU Home Team, EFDA CSU, Max-Planck Institute, Garching, Germany2) ITER Joint Central Team, Garching Joint Work Site, Garching, Germany3) JA Home Team, Japan Atomic Energy Research Institute (JAERI), Japan4) RF Home Team, Research and Development Institute of Power Engineering (RDIPE), Moscow,Russian Federation
Abstract. The ITER L-4 Blanket Project has achieved substantial progress over the lasttwo years. The qualification of materials so far considered as reference for the shield modulefabrication has been completed, as well as the developments for joining the triplex First Wallstructure. Several Primary Wall, baffle, and limiter mock-ups have been manufactured andtested showing comfortable margins against the loads expected in ITER. Shield prototypes havebeen manufactured by conventional and advanced technology, which have finally demonstratedthe manufacturing feasibility. More recently, activities for the qualification of the module at-tachment system have been started, and first results from materials and mock-up tests havebecome available. Several test campaigns are still to be finished to complete the data base forthe design. In the meantime, further activities have been initiated to adapt the R&D programmeto the ITER-FEAT design features, with the aim to further reduce the cost.
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247
(ITERP/20) Dielectric Properties of the ITER TFMC In-sulation after Low Temperature Reactor Irradiation
K. Humer1), H. W. Weber1), R. Hastik2), H. Hauser2), H. Gerstenberg3)
1) Atominstitut der osterreichischen Universitaten, A-1020 Wien, Austria2) Institut fur Werkstoffe der Elektrotechnik, TU Wien, A-1040 Wien, Austria3) TU Munchen, FRM Reaktorstation, D-85748 Garching, Germany
Abstract. The insulation system for the Toroidal Field Model Coil of ITER is a fiber reinforcedplastic (FRP) laminate, which consists of a combined Kapton/R-glass-fiber reinforcement tape,vacuum-impregnated with an epoxy DGEBA system. Pure disk shaped laminates, disk shapedFRP/stainless-steel sandwiches, and conductor insulation prototypes were irradiated at 5 K in afission reactor up to a fast neutron fluence of 1022m−2 (E > 0.1MeV) to investigate the radiationinduced degradation of the dielectric strength of the insulation system. After warm-up to roomtemperature, swelling, weight loss, and the breakdown strength were measured at 77 K. Thesandwich swells by 4% at a fluence of 5× 1021m−2 and by 9% at 1× 1022m−2. The weight lossof the FRP is 2% at 1 × 1022m−2. The dielectric strength remained unchanged over the wholedose range.
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248
(ITERP/21) Research and Development for the ITER ToroidalField Coils
E. Salpietro1), on behalf of the ITER Joint Central Team and the EU Home Team
1) EFDA Close Support Unit, Boltzmannstr. 2, D-85748 Garching, Germany
Abstract. The ITER Toroidal Field (TF) coils are made up of a winding pack enclosed in acase. In the central region the noses of the coils are wedge shaped and fit together to form acircular vault. On the outside an intercoil support structure joins the coil above and below theequator. The goal of the ITER project L2 is to verify the design principles, design procedures,design criteria, operating margins, analysis methods and manufacturing process, including Qual-ity Assurance (QA) capable of application to the ITER TF coils. The project is divided intotwo subprojects: TF Model Coil (TFMC) construction and testing and TF coil case fabricationdemonstration. The conceptual design of the ITER TFMC has been carried out by the ITEREU HT, the engineering design and construction by European Industries. The testing of theTFMC is foreseen in the TOSKA facility at FZK Karlsruhe starting in the first quarter of 2001.The feasibility demonstration of the TF coil case is being carried out also by European industryby: Forging trapezoidal tubes with variable wall thickness, casting new modified 316LN typematerial for the intercoil structure and the parts of the case subject to lower stresses, qualifyingthe welding and NDT methods to be applied to the heavy thickness (∼250mm) to be joinedtogether to form the casing.
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249
(ITERP/22) Neutronics Experiments for ITER at JAERI/FNS
C. Konno1), F. Maekawa1), Y. Kasugai1), Y. Uno1), J. Kaneko1), T. Nishitani1), M. Wada2),Y. I. Ikeda1), H. Takeuchi1)
1) Japan Atomic Energy Research Institute (JAERI), Ibaraki-ken, Japan2) Startcom Co., Ltd., Tokyo, Japan
Abstract. A series of fusion neutronics experiments has been performed at the Fusion Neu-tronics Source (FNS) facility at JAERI as ITER/EDA R&D Tasks in order to deal with variousnuclear problems originating from 14-MeV neutrons in ITER. Recently three experiments werecarried out; 1) straight duct streaming experiments, 2) decay heat experiments and 3) devel-opment of a fusion power monitor utilizing activation of water. The straight duct streamingexperiments suggest that the calculation accuracy for straight duct streaming analyses in ITERnuclear designs is ±40 %. The decay heat experiments show that the accuracy of the decay heatcalculation is within 10 % for copper and type 316 stainless steel, while it is ∼ 30 % for tungsten.It is demonstrated that a fusion power monitor utilizing activation of water is applicable to ITER.
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250
(ITERP/23) Status of the Extended Performance Tests forBlanket Remote Maintenance in ITER L6 Project
S. Kakudate1), K. Oka1), T. Yoshimi1), M. Hiyama1), K. Taguchi1), K. Shibanuma1), K. Koizumi1),Y. Matsumoto2), T. Honda2) and R. Haange2)
1) 1) Japan Home Team JAERI , Ibaraki-ken, Japan2) ITER Joint Center Team, Naka Joint Work Site, Ibaraki-ken, Japan
Abstract. Mechanically attached blanket module insertion tests were carried out consideringthe misalignment between module and back plate. Through the insertion tests, the module wassuccessfully inserted up to the misalignment of ±10 mm under the clearance of ±0.16 ∼ ±0.18mm between key and groove. This was achieved by the passive compliance due to the flexibilityof the manipulator through the assistance of the chamfer configuration of the key for smoothinsertion. In addition, the “correlation coefficient” based on the results obtained by the straingages located at the end-effector was found to be useful in order to estimate the forces of thecomplicated end-effector during module insertion for the development of the sensor based control.
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251
(ITERP/24) ITER-FEAT Divertor Remote Maintenance
D. Maisonnier1), G. Cerdan2), S. Chiocchio3), C. Damiani4), J.-P. Friconneau2), R. Haange5),M. Irving4), E. Martin3), J. Palmer1), A. Poggianti6), M. Siuko7), E. Tada8), N. Takeda8),A. Tesini5), R. Tivey3), A. Turner9)
1) EFDA, Boltzmannstrasse 2, 85748 Garching, Germany2) CEA, DTA/DPSA/STR, BP6, 92265 Fontenay aux Roses Cedex, France3) ITER Joint Work Site, Boltzmannstr. 2, 85748 Garching, Germany4) ENEA, CR Brasimone, CP 1, I-40032 Camugnano (Bo), Italy5) ITER Joint Work Site, Naka, Naka-machi, Ibaraki-ken, 311-01 Japan6) ENEA, via Martiri di Montesole 4, 40129 Bologna, ITALY7) IHA, Korkeakoulunkatu 2, Box 589, FIN-33101 Tampere, Finland8) JAERI, 2-4 Shirane, Shirakata, Tokai-mura, Naka-Gun, Ibaraki 319-11 Japan9) NNC Limited, Booths Hall, Chelford Road, Knutsford, Cheshire, WA16 8QZ, England
Abstract. Remote replacement of the ITER divertor will be required several times during thelife of the machine. To facilitate its regular exchange the divertor is assembled in the vacuumvessel in 54 cassettes, each being introduced into the vessel through one of three equispacedhandling ports. The remote replacement of plasma facing components in the hot-cell allowsthe cassette bodies to be re-used and to minimise the amount of activated waste. An R&Dproject was conceived during the ITER EDA to demonstrate the feasibility of divertor remotemaintenance operations. Two test platforms have been set up and are being used to evaluateequipment and procedures. Following a short description of the test facilities set up at ENEABrasimone, Italy, this paper reports the test results which confirm the overall feasibility of theproposed maintenance and refurbishment schemes.
(IF/5) The Wire Array Z Pinch Programme at Imperial College . . . . . . . 257
(IF/7) Studies in the Evolution of Hydrodynamic Instabilities and theirRole in Inertial Confinement Fusion . . . . . . . . . . . . . . . . . . . . . 258
253
(IF/1) Research on Imploded Plasma Heating by ShortPulse Laser for Fast Ignition
R. Kodama1), K. Mima1), Y. Kitagawa1), K. Fujita1), N. Miyanaga1), H. Nishimura1), N. Izumi1),H. Habara1), A. Sunahara1), Y. Sentoku2), M. Heya3), H. Fujita1), M. Mori1), H. Yoshida1),T. Jitsuno1), Y. Izawa1), M. Murakami1), K. Nishihara1), T. Yamanaka1)
1) Institute of Laser Engineering of Osaka University, Osaka, Japan2) Institute of Laser Technology, Osaka, Japan3) Institute of Free Electron Laser, Osaka, Japan
Abstract. Since the peta watt module (PWM) laser was constructed in 1995, investigatedare heating processes of imploded plasmas by intense short pulse lasers. In order to heat thedense plasma locally, a heating laser pulse should be guided into compressed plasmas as deeplyas possible. Since the last IAEA Fusion Conference, the feasibility of fast ignition has beeninvestigated by using the short pulse GEKKO MII glass laser and the PWM laser with GEKKOXII laser. We found that relativistic electrons are generated efficiently in a preformed plasma toheat dense plasmas. The coupling efficiency of short pulse laser energy to a solid density plasmais 40% when no plasmas are pre-formed, and 20% when a large scale plasma is formed by a longpulse laser pre-irradiation. The experimental results are confirmed by numerical simulations us-ing the simulation code “MONET” which stands for the Monte-Carlo Electron Transport codedeveloped at Osaka. In the GEKKO XII and PWM laser experiments, intense heating pulsesare injected into imploded plasmas. As a result of the injection of heating pulse, it is found thathigh energy electrons and ions could penetrate into imploded core plasmas to enhance neutronyield by factor 3∼5.
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254
(IF/2) Target Design Activities for Inertial Fusion Energyat Lawrence Livermore National Laboratory
M. Tabak1), M. Herrmann1), D. Callahan-Miller1), S. Hatchett1), L. J. Perkins1), J. Lindl1)
1) Lawrence Livermore National Laboratory,Livermore, CA 94550 USA
Abstract. We studied a variety of targets to be driven by ion beams or lasers in the pastyear. In order to relax target fabrication requirements, expand the allowed beam phase spacevolume and meet some radiological safety requirements, we continued to extend the set of thedistributed radiator target designs for heavy ion beams. The hydrodynamic stability of a highgain directly driven laser target recently proposed at the Naval Research Laboratory has beenstudied. Because target chambers are sensitive to the x-ray spectrum as well as the compositionand energy of the capsule debris we also present these for this target. A novel implosion schemefor the Fast Ignitor fusion scenario that minimizes the amount of coronal plasma that the ig-niting laser beam must penetrate is described. We describe recently derived scaling laws thatrelate the minimum value of the incoming fuel kinetic energy to the peak drive pressure, the fueladiabat and the implosion velocity for capsules that use the kinetic energy of the implosion toheat the hotspot to ignition temperatures.
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255
(IF/3) The National Ignition Facility
W. J. Hogan1), E. Moses1), B. Warner1), M. Sorem2), J. M. Soures3)
1) Lawrence Livermore National Laboratory2) Los Alamos National Laboratory3) University of Rochester
Abstract. The National Ignition Facility (NIF) is the largest construction project ever under-taken at Lawrence Livermore National Laboratory (LLNL). NIF consists of 192 forty-centimeter-square laser beams and a 10-m-diameter target chamber. NIF is being designed and built byan LLNL-led team from Los Alamos National Laboratory, Sandia National Laboratories, theUniversity of Rochester, and LLNL. Physical construction began in 1997. The Laser and TargetArea Building and the Optics Assembly Building were the first major construction activities,and despite several unforeseen obstacles, the buildings are now 92% complete and have beendone on time and within cost. Prototype component development and testing has proceededin parallel. Optics vendors have installed full-scale production lines and have done prototypeproduction runs. The assembly and integration of the beampath infrastructure has been recon-sidered and a new approach has been developed. This paper will discuss the status of the NIFproject and the plans for completion.
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256
(IF/5) The Wire Array Z Pinch Programme at ImperialCollege
M. G. Haines1), S. V. Lebedev1), J. P. Chittenden1), S. N. Bland1), F. N. Beg1), A. E. Dangor1),S. A. Pikuz2), T. A. Shelkovenko2)
1) The Blackett Laboratory, Imperial College, London SW7 2BZ, U. K.2) P.N.Lebedev Physics Institute RAS, Moscow, 117924, Russia
Abstract. Plasma formation and implosion dynamics of wire array z-pinches have been studiedexperimentally using the MAGPIE generator (1.4MA, 240ns) at Imperial College. Simulationsand theory verify much of the data. Both laser probing and x-ray radiography show after aninitial volumetric heating of the wires the presence of dense wire cores surrounded by low den-sity coronal plasma. Radiography shows development of perturbations on the dense core of eachwire, while laser probing shows inward jetting of the coronal plasma caused by the global JxBforce, and these plasma streams are axially non-uniform on the same spatial scale as later seenin the wire cores. The spatial scale of these perturbations (∼ 0.5mm for Al, ∼ 0.25mm forW) increases with the size of the wire cores (∼ 0.25mm for Al, ∼ 0.1mm for W). The inwardflow of the coronal plasma is usually field free and leads to formation on the array axis of astraight plasma column, the dynamics of which is strongly affected by radiation cooling. Imagesobtained by optical streak camera show that the wire cores start their inward motion late andthe implosion trajectory deviates significantly from the expected from 0-D analysis. An increaseof the number of wires (decrease of inter-wire gap) resulted in a transition to 0-D trajectoryfor aluminium wire arrays, but not for tungsten. In experiments with nested wire arrays twomodes of behaviour are observed; in the first the inner array is transparent to the implodingouter array, but the current transfers to it, leading to a fast implosion. The second mode occurswhen a significant fraction of current is flowing in the inner array and the two arrays apparentlyimplode simultaneously. In both modes the x-ray pulse is significantly sharpened in comparisonwith that generated in implosion of a single wire array.
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257
(IF/7) Studies in the Evolution of Hydrodynamic Instabil-ities and their Role in Inertial Confinement Fusion
D. Shvarts1), O. Sadot1)2), D. Oron1), R. Kishony1)4), Y. Srebro1)2), A. Rikanati1)2), D. Kar-toon1)2), Y. Yedvab1)2), Y. Elbaz1)2), A. Yosef-Hai2), U. Alon3), L. A. Levin1), E. Sarid1),L. Arazi4) and G. Ben-Dor2)
1) Nuclear Research Center - Negev, Israel.2) Ben-Gurion University of the Negev, Israel.3) Weizmann Institute, Rehovot, Israel.4) Tel-Aviv University, Tel-Aviv, Israel.
Abstract. Hydrodynamic instabilities, such as the Rayleigh-Taylor and Richtmyer-Meshkovinstabilities, have a central role when trying to achieve net thermonuclear fusion energy via themethod of Inertial Confinement Fusion. We shall review recent theoretical, numerical and exper-imental work that describes the evolution of two- and three- dimensional perturbations. Finally,the effects of these perturbation on the ignition conditions, using new self-similar solutions forperturbed burn wave propagation will be discussed.
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258
Session IFP — Inertial Fusion
Contents
(IFP/01) High Rep-Rate KrF Laser Development and Intense Pulse Inter-action Experiments for IFE . . . . . . . . . . . . . . . . . . . . . . . . . . 260
(IFP/01) High Rep-Rate KrF Laser Development and In-tense Pulse Interaction Experiments for IFE
Y. Owadano1), E. Takahashi1), I. Okuda1), I. Matsushima1), Y. Matsumoto1), S. Kato1), K. Kuwa-hara2), E. Miura1), H. Yashiro1)
1) Electrotechnical Laboratory, Tsukuba, Ibaraki, Japan2) Science University of Tokyo, Noda, Chiba, Japan
Abstract. A high repetition-rate e-beam pumped Krypton fluoride (KrF) laser has been de-veloped as a prototype of future IFE driver. A combination of power supply with high voltagemagnetic switches and e-beam diode cooled by water and self heat-radiation is generating e-beam at 0.5 Hz. Laser oscillation at 1 Hz with output energy over 20 J is expected in nearfuture. Intense short pulse (> 10J, ∼ 3ps) has been generated by using a SBS (stimulated Bril-louin scattering) short pulse generator and one of the Super-ASHURA’s beam line. Interactionexperiments with a power density ∼ 1018W/cm2 were performed with long scale-length plasmaproduced by 20 ns pulse. X-ray images show that the short pulse reached to the point close tothe original target surface.
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260
(IFP/03) Laser Diode Pumped Nd:Glass Slab Laser for In-ertial Fusion Energy
M. Yamanaka1), T. Kanabe1), H. Matsui1), R. Kandasamy1), Y. Tamaoki1), T. Kuroda1), T. Ku-rita1), M. Nakatsuka1), Y. Izawa1), S. Nakai1), T. Kawashima2), Y. Okada2), T. Kanzaki2),H. Miyajima2), M. Miyamoto2), H. Kan2)
1) Institute of Laser Engineering, Osaka University, Osaka, Japan2) Hamamatsu Photonics K.K., Shizuoka, Japan
Abstract. As a first step of a driver development for the inertial fusion energy, we are de-veloping a laser-diode-pumped zig-zag Nd:glass slab laser amplifier system HALNA 10 (HighAverage-power Laser for Nuclear-fusion Application) which can generate an output of 10 J perpulse at 1053 nm in 10 Hz operation. The water-cooled zig-zag Nd:glass slab is pumped fromboth sides by 803-nm AlGaAs laser-diode(LD) module; each LD module has an emitting areaof 420 mm x 10 mm and two LD modules generated in total 218 (max.) kW peak power with2.6kW/cm2 peak intensity at 10 Hz repetition rate. We have obtained in a preliminary experi-ment a 8.5 J output energy at 0.5 Hz with beam quality of 2 times diffraction limited far-fieldpattern, which nearly confirmed our conceptual design.
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261
(IFP/04) Probable Approaches to Develop Particle BeamEnergy Drivers and to Calculate Wall Material Ablationwith X Ray Radiation from Imploded Targets
K. Kasuya1), M. Funatsu1), S. Saitoh1), T. Kamiya1), T. Yamanaka2), S. Nakai2), K. Mima2),H. Nishimura2), T. Renk3), B. Turman3), C. Olson3), J. Quintenz3)
1) Tokyo Institute of Technology, Yokohama, Japan2) Osaka University, Suita, Japan3) Sandia National Laboratories, Albuquerque, USA
Abstract. The first subject was the development of future ion beam driver with medium-massion specie. This may enable us to develop a compromised driver from the point of view ofthe micro-divergence angle and the cost. We produced nitrogen ion beams, and measured themicro-divergence angle on the anode surface. The measured value was 5-6mrad for the abovebeam with 300-400keV energy, 300A peak current and 50ns duration. This value was enoughsmall and tolerable for the future energy driver. The corresponding value for the proton beamwith higher peak current was 20-30mrad, which was too large. So that, the scale-up experimentwith the above kind of medium-mass ion beam must be realized urgently to clarify the beamcharacteristics in more details. The reactor wall ablation with the implosion X-ray was alsocalculated as the second subject in this paper.
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262
(IFP/05) Study of Imploding Plasmas on the S-300 Machine
V. P. Smirnov1), A. Chernenko1), K. Chukbar1), S. Danko1), Yu. Kalinin1), A. Kingsep1), V. Ko-rolev1), E. Grabovsky2), M. Fedulov2), I. Frolov2), S. Nedoseev2), G. Oleinik2), A. Samokhin2),P. Sasorov2), G. Volkov2), M. Zurin2)
1) Russian Research Center “Kurchatov Institute”, 123182, Moscow, Russia2) SSC Troitsk Institute for Innovation and Fusion Research (TRINITI),142092 Troitsk, Moscowreg., Russia.
Abstract. Experimental results of the micro-liner implosion study on S-300 and “Angara-5”machines have been presented as well as some theoretical results. Our investigations were carriedout in a broad range of liner constructions, masses, and chemical compositions. On the S-300machine with a current value up to 3.5 MA, 1.5 TW X-ray pulse was obtained as a result ofmulti-wire liner implosion. The measurements of X-ray spectra have been presented. Theoreticalinvestigations and 2.5-d simulations were devoted to some critical points. Double liner implosionapproach for X-ray production and distinction between gas-puff and wire array implosions on“Angara-5” have been described. The test-bed for multi MJ X-ray generator “Baikal” namedas “MOL” is presented.
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263
(IFP/07) Development of High Quality Plastic Fuel Shellsfor Laser Fusion Energy
T. Norimatsu1), K. Nagai, T. Yamanaka
1) ILE Osaka University, Osaka, Japan
Abstract. An overview of the emulsion process to make fuel capsules for a laser fusion powerplant is presented, emphasizing the fact that high quality shells, of which sphericity is close tothe extrapolated NIF standard, were successfully fabricated. A simulation model for the center-ing process, by which uniformly thick shells were formed, was compared with the experiment,and showed good agreement. It was indicated that the water core approaches the center byrepeated instantaneous deformations.
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264
(IFP/08) Hot Superdense Plasmas and Fast Electron Jetsfrom Intense Femtosecond Laser Pulses
K. J. Witte1), U. Andiel1), K. Eidmann1), E. Fill1), C. Gahn1), I. Golovski2), D. Habs3),R. Mancini2), J. Meyer-ter-Vehn1), G. Pretzler1), A. Pukhov1), R. Rix1), T. Schlegel1), andG. Tsakiris1)
1) Max-Planck-Institut fur Quantenoptik, Hans-Kopfermann-Straße 1, D-85748 Garching2) Department of Physics, University of Nevada, Reno NV 89557-0058, USA3) Sektion Physik, LMU Munchen, Am Coulombwall 1, D-85748 Garching
Abstract. With high-contrast light pulses of ≤200-fs duration focused to intensities of ≥1018W/cm2, strongly coupled plasmas at solid state density and temperatures of a few 100eVcan be generated. These plasmas are similar to those currently achievable only in indirectly im-ploded gas-filled microspheres during stagnation. The table-top size and high-repetition rate ofthe lasers producing these ultra-short pulses provided in our case by the 2-TW facility ATLASallow to systematically investigate the basic features of these plasmas. First the mechanismof the dense-plasma generation is discussed. This is based on hydrocode and PIC simulationsas well as on measurements of light absorption and the energy transport into the target. Wethen present spectrally and time-resolved measurements of the K-shell emission from aluminumtargets. By fitting synthetic spectra obtained from code simulations to the experimental ones,it can be inferred that the plasma has an electron density of ∼ 1024/cm3 and a temperatureof ∼ 300eV. In the fast-ignitor concept, a laser-generated electron jet is expected to ignite thecentral spot of the compressed target. Using ATLAS pulses focused to on-target intensities of1018 − 1019W/cm2, we investigate the e-beam generation mechanism in preformed under-denseplasmas. A new fast-electron acceleration mechanism is identified.
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265
(IFP/09) Super-penetration of Ultra-intense Laser Light inLong Scale Length Plasmas Relevant to Fast Ignitor
R. Kodama1), K. Mima1), K. A. Tanaka2), Y. Kitagawa1), H. Fujita1), T. Norimatsu1), H. Yoshida1),Y. Sentoku1), T. Kawamura1), H. Habara1), Y. Tohyama1), M. Tampo1), T. Miyakoshi1),Y. Izawa1), T. Yamanaka1)
1) Institute of Laser Engineering, Osaka University, Osaka, Japan2) The Faculty of Engineering, Osaka University, Osaka, Japan
Abstract. The scheme of fast ignition (FI) in inertial fusion energy (IFE) would have the advan-tage to obtain higher pellet gains with a smaller energy driver as compared with the conventionalmethod or self-ignition scheme making a central hot spark. Ultra-intense laser interactions withlong scale length plasmas are especially essential for the FI research. We have studied the inter-actions with long scale-length plasmas using our 50-100 TW laser systems and particle-in-cell(PIC) simulation codes. We focused on the generation and transport of high energy densityparticles in the interactions as well as the laser beam behavior in long scale-length plasmas. Asuper penetration mode of ultra-intense laser light has been observed to penetrate in the longscale plasma to the solid density surface. The penetrated laser pulse has been efficiently con-verted to energetic electrons and ions. These results are hopeful for the progress of FI researchin IFE.
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266
(IFP/10) Update on Fast Ignition Experiments at NovaPetawatt
R. B. Stephens1), T. E. Cowan2)∗, R. R. Freeman3), S. Hatchett2), M. H. Key2), J. A. Koch2),R. W. Lee2), A. MacKinnon2), D. Pennington2), R. Snavely2), M. Tabak2), and K. Yasuike2)
1) General Atomics, San Diego, California USA2) Lawrence Livermore National Laboratory, Livermore, California USA3) University of California, Davis, Livermore, California USA∗ Present Address: General Atomics
Abstract. The physics of fast ignition was studied on the PetaWatt laser facility at LLNLfor ∼ 3 years, to May 1999. The previous report to this conference described experiments thatdemonstrated the efficient transfer of laser energy to relativistic electrons that penetrated intothe target and heated to temperatures ∼ 1keV. Since then, we have looked at energy transferand propagation in dense plasmas in considerably more detail. Measurements show that therelativistic electrons penetrate > 100µm into a CH foil in a collimated beam with a complexannular structure. Production of an energetic (up to 55 MeV) proton beam was also discovered.The protons are tightly bunched (< 40o C spread) and are emitted normal to the back targetsurface, so can be accurately directed. This gives another promising possibility for delivery ofthe ignition pulse.
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267
(IFP/11) Cluster Induced Ignition – A New Approach toInertial Fusion Energy
T. Desai1), J. T. Mendonca2), D. Batani3), A. Bernardinello3)
1) National Research Institute for Applied Mathematics, 492/G, 7th Block, Jaya Nagar, Bangalore-560082, India2) GOLP, Instituto Superior Technico, 1096-Lisboa codex, Portugal3) Dipartimento di Fisica “G. Occhialini”, Universita degli Studi di Milano-Bicocca and INFM,Via Emanueli 15 - 20126 Milano, Italy
Abstract. An ultra intense laser interaction with clusters produce energetic ions and electronsin MeV range due to cluster explosion. Here we discuss the possibility of harnessing these par-ticle energies to heat a part of the pre compressed DT fuel to ignition condition. In this articlewe are striving to present the principle concept and the preliminary results are discussed.
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268
(IFP/13) Recent Developments in Ignition Target Designfor the National Ignition Facility
L. J. Suter1), S. Haan1), J. Lindl1), T. Dittrich1), B. A. Hammel1), D. Hinkel1), O. Jones1),S. Pollaine1), J. Rothenberg1)
1) Lawrence Livermore National Laboratory, Livermore, California 94551
Abstract. Work on design of ignition targets for the National Ignition Facility (NIF) has pro-gressed in three areas. First, hohlraums have been re-optimized taking advantage of improve-ments in efficiency in several areas: use of high albedo material mixtures in the hohlraum wall;optimizing the laser entrance hole; optimizing the case-to-capsule ratio; and taking advantageof increased efficiency of longer pulses. These changes, in combination, allow for the possibilityof quite high yields (∼ 100MJ), gains (> 40) and significantly more margin for ignition on NIF.Second, work has continued on specifications for target fabrication. Third, detailed design andanalysis has been done on targets for the commissioning phase of NIF, when only 96 beams areavailable. We find excellent hydrodynamic similarity is possible with sub-scale cryogenic targets.These targets can be used to test all of the physics of full-scale ignition targets in detail except,perhaps, for ignition itself.
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269
(IFP/14(R)) The GEKKO XII-HIPER (High Intensity PlasmaExperimental Research) System Relevant to Ignition Tar-gets
N. Miyanaga1), M. Nakatsuka1), H. Azechi1), H. Shiraga1), T. Kanabe1), H. Asahara1), H. Daido1),H. Fujita1), K. Fujita1), Y. Izawa1), T. Jitsuno1), T. Kawasaki1), H. Kitamura1), S. Matsuo1),K. Mima1), N. Morio1), M. Nakai1), S. Nakai1), K. Nishihara1), H. Nishimura1), T. Sakamoto1),K. Shigemori1), K. Sueda1), K. Suzuki1), K. Tsubakimoto1), H. Takabe1), S. Urushihara1),H. Yoshida1), T. Yamanaka1), C. Yamanaka2)
1) Institute of Laser Engineering, Osaka University, 2-6 Yamada-oka, Suita, Osaka 565-0871,Japan2) Institute for Laser Technology, 1-8-4 Utsubohonmachi, Nishi-ku, Osaka 550-0004, Japan
Abstract. To test high gain targets surrogated in the planar geometry, we have constructeda new experimental system (HIPER) which provides the high ablation pressure with a uniformirradiance profile. These performances were achieved by bundling twelve beams of the existingGEKKO XII into a F/3 focus cone. The partially coherent light is introduced for the beamsmoothing of a green foot pulse consisting of three beams, and the three-directional smoothingby spectral dispersion is utilized for residual nine beams delivering a blue main drive pulse. Thedetail of design concept and results of initial activation of this system are reported.
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270
(IFP/15(R)) Experimental Studies on Hydrodynamic In-stability of Direct-Drive Laser Fusion on GEKKO XII
M. Nakai1), H. Azechi1), N. Izumi1), R. Kodama1), O. Maegawa1), M. Matsuoka1), K. Mima1),N. Miyanaga1), T. Nagaya1), S. Nakai1), M. Nakatsuka1), K. Nishihara1), M. Nishikino1), H. Nishimura1),T. Norimatsu1), K. Shigemori1), H. Shiraga1), A. Sunahara1), H. Takabe1), T. Yamanaka1),C. Yamanaka2)
1) Institute of Laser Engineering, Osaka University, Osaka, JAPAN2) Institute for Laser Technology, Osaka, JAPAN
Abstract. A series of elementary experiments on the hydrodynamic instability has been con-ducted at the GEKKO XII laser system to investigate hydrodynamic instability growth in planarfoils directly irradiated by 0.53-µm laser light. The dynamic property of laser initial imprint wassuccessfully obtained by the improved two-wavelength Young’s interference method. Two typesof the “indirect/direct hybrid” schemes were also examined to quantify the effective mitigationof laser initial imprint. Especially, a multi-density foam layer was utilized for the first timeto demonstrate the feasibility of the density varying target. Rayleigh-Taylor instability growthrate on the directly driven target around its maximum was also measured with the perturbationwavelengths less than 10 µm by a newly developed technique of x-ray moire interferometry.
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271
(IFP/16(R)) Nonlinear Theory of Laser Imprint, Richtmyer-Meshkov and Rayleigh-Taylor Instabilities
K. Nishihara1), M. Murakami1), C. Matsuoka2), N. Ohnishi1), T. Ikegawa1), Y. Fukuda1),A. Sunahara1), H. Nagatomo1), H. Takabe1), K. Mima1)
1) Institute of Laser Engineering, Osaka University, Osaka, Japan2) Faculty of Science, Ehime University, Matsuyama, Japan
Abstract. Implosion process in laser fusion can be divided into three phases: start-up, ac-celeration and stagnation phases and various hydrodynamic instabilities appear in each phase.Analytical models are developed to study nonlinear evolutions of the hydrodynamic instabilitiesin these phases, and compared with multi-dimensional simulations. We mainly discuss the for-mation of double spiral structures caused by the singularity of vorticity in the RM spikes, theeffect of the ablative stabilization in nonlinear growth of the RT instability with a finite band-width and the effect of the radiation cooling on the stability in the stagnation phase. Varioussmoothing effect, such as radiation smoothing and hydrodynamic smoothing, are also studied inthe start-up phase.
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272
(IFP/17) Two-Dimensional Analysis of Energy and Mo-mentum Deposition Effects of Alpha Particles in ICF Plas-mas
T. Johzaki1), Y. Kuroki1), Y. Nakao1)
1) Department of Applied Quantum Physics and Nuclear Engineering, Kyushu University, Fukuoka,JAPAN
Abstract. Two types of two-dimensional codes, i.e. a transport code and a diffusion one, havebeen developed for analysis of alpha-particle effects in ICF plasmas. On the basis of 2-D cou-pled transport/hydrodynamic simulations, we investigate the energy and momentum depositioneffects of alpha-particles on the Rayleigh-Taylor instability in a stagnating DT planar plasma.After the accuracy validation of the diffusion code, the sensitivity of fuel gain to the perturbationamplitude is also discussed by carrying out 2-D coupled diffusion / hydrodynamic simulationsfor a DT spherical target with mode number of l = 2− 12.
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273
(IFP/18) Simulations of ICF Hohlraum and Radiation Hy-drodynamics in a Plastic Foil
N. K. Gupta1), T. C. Kaushik1), B. K. Godwal1)
1) High Pressure Physics Division, Bhabha Atomic Research Centre, Trombay, Mumbai–400085 (INDIA)
Abstract. In this paper we study the effects of various parameters on the estimation of radi-ation temperature inside indirect drive ICF hohlraum and compare results with the publishedwork. A multi group, 1-D, cylindrical geometry, radiation hydrodynamic code is used for thisstudy. The radiation temperature inside a cylindrical hohlraum is seen to be strongly dependenton the number of frequency groups used. It is also seen that erroneous results can be obtainedif the space mesh in the hohlraum wall is not fine enough. The spectrum of the radiation isalso seen to be different from Planck, especially in the high-energy range. Hydrodynamics of aCHBr foil driven by the hohlraum radiation brings about the need of a proper equation of statemodel. A three term equation of state is used for these studies and data for CHBr is generatedby scaling density, atomic number and weight of the high-density polysterene (CH)n to that ofCHBr.
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274
(IFP/19) Implosion Physics, Alternative Targets Design andNeutron Effects on Inertial Fusion Systems
G. Velarde1), J. M. Perlado1), J. M. Martınez-Val1), E. Mınguez1), M. Piera1), J. Sanz1), P. Ve-larde1), E. Alonso2), E. Domınguez1), J. G. Rubiano1), J. M. Gil1), J. G. del Rio1), D. Lodi1),L. Malerba1), J. Marian3), P. Martel1), F. Ogando1), J. Prieto1), S. Reyes3), M. Salvador1),P. Sauvan1), M. Velarde1)
1) Instituto de Fusion Nuclear (DENIM), ETSII, Universidad Politecnica de Madrid, Spain2) present address: Polytechnique University Zurich, Switzerland3) present address: Lawrence Livermore National Laboratory (LLNL), USA
Abstract. A new radiation transport code has been coupled with an existing multimaterialfluidynamics code using Adaptive Mesh Refinement (AMR) and its testing is presented, solv-ing ray effect and shadow problems in SN classical methods. Important advances in atomicphysics, opacity calculations and NLTE calculations, participating in significant experiments(LULI/France), have been obtained. Our new 1D target simulation model allows consideringthe effect of inverse Compton scattering in DTx targets (x < 3%) working in a catalytic regime,showing the effectiveness of such tritium-less targets. Neutron activation of all natural ele-ments in IFE reactors for waste management and that of target debris in NIF-type facilitieshave been completed. Pulse activation in structural walls is presented with a new modeling.Tritium atmospheric dispersion results indicate large uncertainties in environmental responsesand needs to treat the two chemical forms. We recognise recombination barriers (metastabledefects) and compute first systematic high-energy displacement cascade analysis in SiC, andradiation damage pulses by atomistic models in metals. Using Molecular Dynamics we explainthe experimental evidence of low-temperature amorphization by damage accumulation in SiC.
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275
(IFP/20) Recent Progress of an Integrated Implosion Codeand Modeling of Element Physics
H. Nagatomo1), H. Takabe1), K. Mima1), N. Ohnishi1), A. Sunahara1), T. Takeda1), K. Nishi-hara1), A. Nishiguchu2), K. Sawada3)
1) Institute of Laser Engineering, Osaka University, Osaka Japan2) Osaka Institute of Technology, Osaka, Japan3) Department of Aeronautics and Space Engineering, Tohoku University, Sendai, Japan
Abstract. Physics of the inertial fusion is based on a variety of elements such as compressiblehydrodynamics, radiation transport, non-ideal equation of state, non-LTE atomic process, andrelativistic laser plasma interaction. In addition, implosion process is not in stationary state andfluid dynamics, energy transport and instabilities should be solved simultaneously. In order tostudy such complex physics, an integrated implosion code including all physics important in theimplosion process should be developed. The details of physics elements should be studied andthe resultant numerical modeling should be installed in the integrated code so that the implosioncan be simulated with available computer within realistic CPU time. Therefore, this task can bebasically separated into two parts. One is to integrate all physics elements into a code, which isstrongly related to the development of hydrodynamic equation solver. We have developed 2-Dintegrated implosion code which solves mass, momentum, electron energy, ion energy, equationof states, laser ray-trace, laser absorption radiation, surface tracing and so on. The reasonableresults in simulating Rayleigh-Taylor instability and cylindrical implosion are obtained usingthis code. The other is code development on each element physics and verification of thesecodes. We had progress in developing a nonlocal electron transport code and 2 and 3 dimensionradiation hydrodynamic code.
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276
Session ICP — InnovativeConcepts
Contents
(ICP/01) Charged Particles Beams Measurements in Plasma Focus Discharges278
(ICP/16) Physics and Engineering Design of a Low Aspect Ratio Quasi-Axisymmetric Stellarator CHS-qa . . . . . . . . . . . . . . . . . . . . . . 292
(ICP/17) D-T Ignition in a Z Pinch Compressed by an Imploding Liner . . 293
277
(ICP/01) Charged Particles Beams Measurements in PlasmaFocus Discharges
L. Jakubowski1), M. Sadowski1), J. Zebrowski1)
1) The Andrzej Soltan Institute for Nuclear Studies, Swierk by Warsaw, Poland
Abstract. Experimental studies performed with many Plasma-Focus (PF) facilities have shownthat simultaneously with the emission of X-ray pulses and intense relativistic electron beams(REBs) there also appears the emission of pulsed ion streams of a relatively high energy (up toseveral MeV). Such ions are emitted mainly along the z-axis of the PF discharge, although theion angular distribution is relatively wide. From PF discharges with deuterium filling fast neu-trons produced by nuclear fusion reactions are also emitted. The paper concerns studies of theenergetic ion beams and their correlation with the pulsed REBs. Time-integrated measurementswere performed with an ion pinhole camera equipped with solid-state nuclear track detectors(SSNTDs), and time-resolved studies were carried out with a scintillation detector, enabling thedetermination of an ion energy spectrum on the basis of the time-of-flight (TOF) technique.
A. A. Skovoroda1), V. V. Arsenin1), E. D. Dlougach1), V. M. Kulygin1), A. Yu. Kuyanov1),A. V. Timofeev1), V. A. Zhil’tsov1), A. V. Zvonkov1)
1) OGRA, Nuclear Fusion Institute, RRC Kurchatov Institute, Moscow, Russia
Abstract. Within the framework of the conceptual project “Adaptive Plasma EXperiment” atrap with the closed magnetic field lines “Experimental Pseudo-Symmetric trap” is examined.The project APEX is directed at the theoretical and experimental development of physical foun-dations for stationary thermonuclear reactor on the basis of an alternative magnetic trap withtokamak-level confinement of high β plasma. The fundamental principle of magnetic field pseu-dosymmetry that should be satisfied for plasma to have tokamak-like confinement is discussed.The calculated in paraxial approximation examples of pseudosymmetric curvilinear elementswith poloidal direction of B isolines are adduced. The EPSILON trap consisting of two straightaxisymmetric mirrors linked by two curvilinear pseudosymmetric elements is considered. Theplasma currents are short-circuited within the curvilinear element what increases the equilib-rium β. The untraditional scheme of MHD stabilization of a trap with the closed field lines bythe use of divertor inserted into axisymmetric mirror is analyzed. The experimental installationEPSILON-OME that is under construction for experimental check of divertor stabilization isdiscussed. The possibility of ECR plasma production in EPSILON-OME under conditions ofhigh density and small magnetic field is examined.
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279
(ICP/04) Experimental Results on STPC-E Machine withan Alternative Non-Inductive Current Drive
A. Sinman1), S. Sinman1)
1) Middle East Technical University, Electrical and Electronics Engineering Department, InonuBulvari 06531 Ankara, Turkey
Abstract. At the Maastricht Conference, A Paramagnetic Spherical Tokamak with PlasmaCenterpost (STPC) machine consisting of the conceptual design and the computational experi-mental results, as well as the cross-sectional layout of the constructional properties of the STPCmachine were presented. In this study, the preliminary experimental results and their assessmentobtained from the STPC-E (experimental set-up version of STPC) machine whose building hasrecently been completed is given. In order to form and control the plasma core of a sphericaltokamak in the STPC-E machine, a novel non-inductive current drive method is applied. In thismethod, an energetic pulse forming line and its direct coupling open circuit transverse termina-tion have been connected to form the plasma core of a spherical tokamak. The measured basicplasma parameters of the STPC-E machine are: rms electron density, nrms
e = 1020 − 1022m−3
in confinement time, tconf = 45− 60ms, electron temperature, Te = 30− 45eV, average plasmacurrent, 〈Ipl〉 = 1.5 − 1.8kA in helical form, maximum toroidal and poloidal magnetic fieldsBmax
t = 1.2kG, Bmaxp = 0.8kG and sustainment time, tst
∼= 10ms. Increasing the number ofSTPC-E machine’s modules, it is possible to extend the sustainment time up to 100–120 msorders.
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280
(ICP/05) Measurements of Strongly Localized Potential WellProfiles in an Inertial Electrostatic Fusion Neutron Source
K. Yoshikawa1), K. Takiyama2), T. Koyama1), K. Taruya1), K. Masuda1), Y. Yamamoto1),T. Toku1), T. Kii1), H. Hashimoto1), N. Inoue1), M. Ohnishi3), H. Horiike4)
1) Institute of Advanced Energy, Kyoto University, Uji, Kyoto 611-0011, Japan2) Hiroshima University, Higashi Hiroshima 739-8527, Japan3) Kansai University, Suita, Osaka 564-8680, Japan4) Osaka University, Suita, Osaka 565-0871, Japan
Abstract. Direct measurements of localized electric fields are made by the laser-induced flu-orescence (LIF) method by use of the Stark effects in the central cathode core region of anInertial-Electrostatic Confinement Fusion (IECF) neutron (proton) source, which is expectedfor various applications, such as luggage security inspection, non-destructive testing, land minedetector, or positron emitter production for cancer detection, currently producing continuouslyabout 107 n/sec D-D neutrons. Since 1967 when the first fusion reaction was successfully provedexperimentally in a very compact IECF device, potential well formation due to space chargeassociated with spherically converging ion beams has been a central key issue to be clarifiedin the beam-beam colliding fusion, which is the major mechanism of the IECF neutron source.Many experiments, but indirect, were made so far to clarify the potential well, but none of themproduced definitive evidence, however. Results by the present LIF method show a double wellpotential profile with a slight concave for ion beams with relatively larger angular momenta,whereas for ions with smaller angular momenta, potential but much steeper peak to develop.
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281
(ICP/06) An Exploration of Pulsed Magnetic Field DrivenFusion in Z Pinch Configuration
T. C. Kaushik, N. K. Gupta, S. K. H. Auluck
High Pressure Physics Division, Bhabha Atomic Research Centre, Trombay, Mumbai–400 085(India)
Abstract. Scheme to exploit the magnetic field driven near-isentropic compression of matterand explosion of metallic conductors is explored to attain density and temperatures suitable togenerate fusion reactions. A cylindrical metallic shell, containing a capsule of deuterated matter,is radially compressed by (JxB) force and then made to explode “just at the end” of currentpeak by optimum thickness. An attempt is also made to enhance the final pressure at center byallowing the return conductor also to burst. Resultant shock may be directed inwards throughthe effect of impedance mismatch at different interfaces. A 1-D radiation hydrodynamic code,modified to include ohmic heating and self-generated magnetic field pressures by pulsed cur-rents, is validated against results for some of the reported experiments. The model is then usedto simulate a target structure for relatively slow sub-MJ capacitor banks such as the recentlycommissioned 280 kJ/ 40 kV capacitor bank at Trombay.
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282
(ICP/07) Ion Motion Modelling within Dynamic Filamen-tary PF Pinch Column
A. Galkowski1), A. Pasternak2), M. Sadowski2)
1) Institute of Plasma Physics and Laser Microfusion, Warsaw, Poland2) The Andrzej Soltan Institute for Nuclear Studies, Swierk by Warsaw, Poland
Abstract. The paper reports on results of the 3-D modelling of ion motion within the plasmafocus (PF) pinch column, which is treated as a lamentary, non-stationary system. Beside mag-netic fields also induced electrical fields as well as ion-ion and ion-electron collisions have beenincluded in the equations of ion (deuteron) motion. Obtained results show that lamentary struc-ture influences 3-D ion trajectories and 2-D angular distribution of ions. It has been confirmedthat 3-D flower-like filaments can explain some important characteristics of the ion emissionfrom PF discharges.
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283
(ICP/08) Optimal Regimes for Ignition and the Ignitor Ex-periment
B. Coppi1), A. Airoldi2), F. Bombarda1), G. Cenacchi1), P. Detragiache1), L. E. Sugiyama3)
1) Ignitor Project Group - ENEA, Italy2) CNR - Istituto di Fisica del Plasma, Italy3) Massachusetts Institute of Technology, Cambridge MA, USA
Abstract. The optimal conditions under which confined plasmas can reach ignition are iden-tified referring in particular to the parameters of the Ignitor machine. The key importance ofthe radial profiles of the particle density, the thermal energy diffusivity, the plasma pressure,the function q(ψ), etc., is demonstrated. Peaked density profiles, such as those obtained inthe Alcator A and C experiments (at about the same central density and magnetic field as inIgnitor), characterized by minimal thermal diffusivities and high plasma purity, are shown to bebest suited for ignition. The H-mode regime in Ignitor is accessible but not considered a prioritybecause of the typically flat density profiles. The role of collective modes and their interactionwith both high and low energy α-particle populations are assessed. For the modes generatingsawtooth oscillations and involving magnetic reconnection the stabilizing effect of “shoulder”q(ψ) profiles is pointed out together with the role of the trapped ion population.
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284
(ICP/09) Spheromak Formation Studies in SSPX
D. N. Hill1), R. H. Bulmer1), B. I. Cohen1), E. B. Hooper1), L. L. LoDestro1), N. Mattor1),H. S. McLean1), C. T. Holcomb2), T. R. Jarboe2), C. R. Sovinec3), Z. Wang3), G. Wurden3)
1) Lawrence Livermore National Laboratory, USA2) University of Washington, Seattle, USA3) Los Alamos National Laboratory, USA
Abstract. We present results from the Sustained Spheromak Physics Experiment (SSPX) atLLNL, which has been built to study energy confinement in spheromak plasmas sustained forup to 2 ms by coaxial DC helicity injection. Peak toroidal currents as high as 600kA havebeen obtained in the 1m dia. (0.23m minor radius) device using injection currents between200–400kA; these currents generate edge poloidal fields in the range of 0.2–0.4T. The internalfield and current profiles are inferred from edge field measurements using the CORSICA code.Density and impurity control is obtained using baking, glow discharge cleansing, and titaniumgettering, after which long plasma decay times (τ ≥ 1.5ms) are observed and impurity radiationlosses are reduced from ∼ 50% to < 20% of the input energy. Thomson scattering measurementsshow peaked electron temperature and pressure profiles with Te(0) ∼ 120eV and βe ∼ 7%. Edgefield measurements show the presence of n=1 modes during the formation phase, as has beenobserved in other spheromaks. This mode dies away during sustainment and decay so that edgefluctuation levels as low as 1% have been measured. These results are compared with numericalsimulations using the NIMROD code.
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285
(ICP/10) Measurement of Solid Liner Implosion for Mag-netized Target Fusion
R. E. Siemon1), T. Cavazos3), D. Clark1), S. K. Coffey4), J. H. Degnan2), R. J. Faehl1),M. H. Frese4), D. Fulton1), J. C. Gueits2), D. Gale3), T. W. Hussey2), T. P. Intrator1), R. Kirk-patrick1), G. F. Kiuttu2), F. M. Lehr2), J. D. Letterio2), I. Lindemuth1), W. McCullough2),R. Moses1), R. E. Peterkin2), R. E. Reinovsky1), N. F. Roderick5), E. L. Ruden2), J. S. Shlachter1),K. F. Schoenberg1), W. Sommars2), J. M. Taccetti1), P. J. Turchi1), M. G. Tuszweski1), G. Wur-den1), F. Wysocki1)
1) Los Alamos National Laboratory, Los Alamos, NM2) Air Force Research Laboratory, Kirtland AFB, NM, USA3) Maxwell Technologies, Inc., Albuquerque, NM4) NumerEx, Albuquerque, NM5) University of New Mexico, Albuquerque, NM
Abstract. Data are presented on the implosion symmetry of a 1-mm-thick 10-cm-diameter30-cm-long solid aluminum cylinder (called a liner). At the moment when radial compression ofmore than 10:1 is achieved, the inward velocity of the inner liner surface is 5 km/sec and linersymmetry is excellent (rms variation in radius of about 6%). This technology is important forMagnetized Target Fusion, the approach being developed where magnetically insulated plasmais compressed to fusion conditions by means of an imploded liner. The construction of a thetapinch to inject a high-density field-reversed configuration (n ∼ 1017cm−3 with T ∼ 300 eV) intothe liner is presently underway.
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286
(ICP/11) Physics Issues of Compact Drift Optimized Stel-larators
D. A. Spong1), S. Hirshman1), L. A. Berry1), J. F. Lyon1), R. H. Fowler1), D. Strickler1),M. J. Cole1), B. N. Nelson1), E. E. Williamson1), A. Ware2), D. Alban2), R. Sanchez3), G. Y. Fu4),D. Monticello4), W. Miner5), P. Valanju5)
1) Oak Ridge National Laboratory, Oak Ridge, TN, U.S.A.2) University of Montana, Missoula, MT, U.S.A.3) Universidad Carlos III de Madrid, Madrid, Spain4) Princeton Plasma Physics Laboratory, Princeton, NJ, U.S.A.5) University of Texas, Austin, TX, U.S.A.
Abstract. Physics issues are discussed for compact stellarator configurations which achievegood confinement by the fact that the magnetic field modulus, |B|, in magnetic coordinates isdominated by poloidally symmetric components. Two distinct configuration types are consid-ered: (1) those which achieve their drift optimization and rotational transform at low β and lowbootstrap current by appropriate plasma shaping; and (2) those which have a greater relianceon plasma β and bootstrap currents for supplying the transform and obtaining quasi poloidalsymmetry. Stability analysis of the latter group of devices against ballooning, kink and verticaldisplacement modes has indicated that stable 〈β〉’s on the order of 15% are possible. The firstclass of devices is being considered for a low β near-term experiment that could explore some ofthe confinement features of the high beta configurations.
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287
(ICP/12) Evidence of Flow Stabilization in the ZaP Z PinchExperiment
U. Shumlak1), E. Crawford1), R. P. Golingo1), B. A. Nelson1), A. Zyrmpas1), D. J. Den Hartog2),D. J. Holly2)
1) Aerospace & Energetics Research Program, Univ. Washington, Seattle, Washington, USA2) Sterling Scientific, Inc., Madison, Wisconsin, USA
Abstract. The stabilizing effect of an axial flow on the m = 1 kink instability in Z pinches hasbeen studied numerically with a linearized ideal MHD model to reveal that a sheared axial flowstabilizes the kink mode when the shear exceeds a threshold. The sheared flow stabilizing effectis investigated with the flow-through Z pinch experiment, ZaP. An azimuthal array of surfacemounted magnetic probes located at the midplane of the 50 cm long pinch plasma measures thefluctuation levels of the azimuthal modes m = 1, 2, and 3. After pinch formation a quiescentperiod is found where the mode activity is reduced to a few percent of the average field. Opticalimages from a fast framing camera and a HeNe interferometer also indicate a stable pinch plasmaduring this time. Doppler shift measurements of a C-III line correspond to an axial flow velocityof 9.6× 104m/s internal to the pinch. During the time when the axial plasma flow is high, theplasma experiences a quiescent period which lasts approximately 800 exponential growth timespredicted by linear theory for a static plasma.
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288
(ICP/13) Reactor Advantages of the Belt Pinch and LiquidMetal Walls
M. Kotschenreuther1), J. Manickam2), J. Menard2), H. Rappaport, L.-J. Zheng, B. Dorland3),R. Miller4), A. D. Turnbull4)
1) Institute for Fusion Studies, Austin, TX USA2) Princeton Plasma Physics Laboratory, Princeton, NJ USA3) Institute for Plasma Research, College Park, MD USA4) General Atomics, San Diego, CA USA
Abstract. MHD stability of highly elongated tokamaks (termed a belt pinch) are consideredfor high bootstrap fraction cases. By employing high triangularity or indentation, and invokingwall stabilization, and β can be increased by a factor of roughly 3 by increasing κ from 2 to 4.Axisymmetric stability up to κ = 4 tolerable by employing a shell which conforms more closelyto the boundary than in present experiments. Engineering difficulties with a close fitting shell ina reactor environment may be overcome by employing a liquid lithium alloy shell. Rapid metalflows can lead to potentially deleterious plasma shifts and damping of the flow.
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289
(ICP/14) Electrically Nonneutralized Plasmas for Hydro-dynamic Confinement
H. Himura1), Z. Yoshida1), Y. Ogawa2), J. Morikawa2), C. Nakashima3), K. Yagi1), H. Saitoh1),N. Shibayama3), S. Tahara3), T. Tatsuno1), S. Kondoh3), M. Iqbal3), F. Volponi3), A. Ito3),S. Osaki1), R. Numata3), and M. Savkovic3)
1) 1) University of Tokyo, Graduate School of Frontier Sciences, Tokyo 113-0033, Japan2) High Temperature Plasma Center, University of Tokyo, Tokyo 113-8656, Japan3) University of Tokyo, Graduate School of Engineering, Tokyo 113-8656, Japan
Abstract. A high-β equilibrium was theoretically predicted by the Beltrami/Bernoulli con-ditions. In order to obtain such attractive plasmas an innovative method using electricallynon-neutralized plasmas has been proposed, and the experiments to demonstrate it have beenconducted on Proto-RT and BX-U. Energetic electrons are successfully injected across vacuummagnetic fields and a strong self-electric field, which is sufficient to drive a super-Alfvenic flowin a target plasma, is formed by the electrons. However, in the target plasma frictional forceswould damp the flow, resulting in loss of the electrons. An initial experimental result on theloss of electrons is presented.
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290
(ICP/15) Contribution of Muon Catalyzed Fusion to FusionEnergy Development
K. Nagamine1)2), T. Matsuzaki1), K. Ishida1), S. N. Nakamura1)3), N. Kawamura1), Y. Mat-suda1)
1) Muon Science Laboratory, RIKEN, Wako, Saitama, Japan2) Meson Science Laboratory, High Energy Accelerator Research Organization (KEK), Tsukuba,Ibaraki, Japan3) Department of Physics, Graduate School of Science, Tohoku University, Sendai, Miyagi, Japan
Abstract. Recent experimental studies on muon catalyzed fusion (µCF) process of D-T mixturehave uncovered anomalously large muon (µ−) regeneration from the (µα)+ stuck atom formedafter nuclear fusion in dtµ-molecule. The result has opened a new direction towards a realizationof the break-even. In addition, high-intensity hadron accelerator projects for neutron source etc.will realize kW µCF reactor once advanced muon generator be installed. Considering these newtrends, we may be able to develop the fusion energy related R&D program based upon the µCFprocess such as materials irradiation facility, tritium breeding, fundamental plasma physics, etc.
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291
(ICP/16) Physics and Engineering Design of a Low AspectRatio Quasi-Axisymmetric Stellarator CHS-qa
S. Okamura1), K. Matsuoka1), S. Nishimura1), M. Isobe1), I. Nomura1), C. Suzuki1), A. Shimizu1),S. Murakami1), N. Nakajima1), M. Yokoyama1), A. Fujisawa1), K. Ida1), K. Itoh1), P. Merkel2),, M. Drevlak2), R. Zille2), S. Gori2), J. Nuhrenberg2)
1) National Institute for Fusion Science, Toki, 509-5292, Japan2) Max-Planck-Institut fur Plasmaphysik, Teilinstitut Greifswald, D-17489, Germany
Abstract. A low-aspect-ratio quasi-axisymmetric stellarator CHS-qa was designed. An opti-mization code was used to design a magnetic field configuration with evaluations of physicalquantities of quasi-axisymmetry, rotational transform, MHD stability and alpha particle colli-sionless confinement. It is shown that the electron neoclassical diffusion coefficient is similarto tokamaks for the low collisional regime. A self-consistent equilibrium with bootstrap cur-rent confirms the global mode stability up to 130 kA for R = 1.5 m and Bt = 1.5 T device.The evaluation of plasma rotation viscosity is greatly suppressed compared with conventionalstellarators. Engineering design was completed with 20 main modular coils and auxiliary coilswhich provide flexibility of configuration study for confinement improvement and MHD stability.
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292
(ICP/17) D-T Ignition in a Z Pinch Compressed by an Im-ploding Liner
L. Bilbao1), L. Bernal1), J. G. Linhart2), G. Verri2)
1) Instituto de Fısica del Plasma, Universidad de Buenos Aires, Buenos Aires (Argentina)2) INFM, Dipartimento di Fisica, Universita di Ferrara, Ferrara (Italy)
Abstract. It has been shown that an m = 0 instability of a Z-pinch carrying a current of the or-der of 10 MA, with a rise time inferior to 10 nsec can generate a spark capable of igniting a fusiondetonation in the adjacent D-T plasma channel. A possible method for generating such currents,necessary for the implosion of an initial large radius, low temperature Z-pinch, can be a radialimplosion of a cylindrical fast liner. The problem has been addressed in a previous publication,without considering the role played by an initially impressed m = 0 perturbation, a mechanismindispensable for the generation of a spark. The liner/Z-pinch dynamics can be solved at sev-eral levels of physical model completeness. The first correspond to a zero-dimensional model inwhich the liner has a given mass per cm length and a zero thickness, the plasma is compressedadiabatically and is isotropic, there are no energy losses and no Joule heating. The second levelis one-dimensional. The Z-pinch plasma is described by the full set of MHD, 2 fluid equations.The liner is treated first as thin and incompressible and subsequently it is assumed that it hasa finite thickness and is composed of a heavy ion plasma, having an artificial but realistic EOS.Both plasma and liner are considered uniform in the Z-direction and only D-T reactions areconsidered. We shall show that, given sufficient energy and speed of the liner, the Z-pinch canreach a volume ignition. The third level is two-dimensional. Plasma and liner are treated asin the 2nd level but either the Z-pinch or the liner is perturbed by an m = 0 non-uniformity.Provided the liner energy is high enough and the initial m = 0 perturbation is correctly chosen,the final neck plasma can act as a spark for D-T ignition.
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293
Session FTP1 — TechnologyDevelopments
Contents
(FTP1/01(R)) Engineering Design of KSTAR Tokamak Main Structure . . 296
(FTP1/02(R)) Development of Long Pulse Heating and Current Drive Sys-tems for the KSTAR Tokamak . . . . . . . . . . . . . . . . . . . . . . . . 297
(FTP1/12) Investigations in the Area of Thermonuclear Structural MaterialScience in the Republic of Kazakhstan . . . . . . . . . . . . . . . . . . . 307
(FTP1/27(R)) Simulation Study of Evacuation and Effects of Metal Vaporin Laser Fusion Liquid Wall Chamber . . . . . . . . . . . . . . . . . . . . 322
(FTP1/29) Oxidation of Carbon Based First Wall Materials of ITER . . . . 323
(FTP1/30) Iron Neutron Data Benchmarking for 14 MeV Source . . . . . . 324
(FTP1/31) Gyrotrons for Fusion. Status and Prospects. . . . . . . . . . . . . 325
295
(FTP1/01(R)) Engineering Design of KSTAR Tokamak MainStructure
K. H. Im1), S. Cho1), N. I. Her1), D. L. Kim1), G. S. Lee1), M. Kwon1), C. J. Do1), J. B. Kim2),Y. C. Kim2), J. S. Lee2), I. K. Yu1), S. R. In3), B. J. Yoon3), G. H. Hong1), B. C. Kim1),G. H. Kim1), W. C. Kim1), J. W. Sa1), and the KSTAR Team
1) Korea Basic Science Institute, Taejon, Republic of Korea2) Hyundai Heavy Industries, Ulsan, Republic of Korea3) Korea Atomic Energy Research Institute, Taejon, Republic of Korea
Abstract. The main components of the KSTAR (Korea Superconducting Tokamak AdvancedResearch) tokamak including vacuum vessel, plasma facing components, cryostat, thermal shieldand magnet supporting structure are in the final stage of engineering design. Hundai Heavy In-dustries (HHI) has been involved in the engineering design of these components. The currentconfiguration and the final engineering design results for the KSTAR main structure are pre-sented.
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296
(FTP1/02(R)) Development of Long Pulse Heating and Cur-rent Drive Systems for the KSTAR Tokamak
B. G. Hong1), B. H. Oh1), B. H. Choi1), W. Namkung2), Y. D. Bae1), Y. S. Bae2), M. H. Cho2),Y. S. Cho1), C. K. Hwang1), S. R. In1), S. H. Jeong1), H. S. Kang3), K. R. Kim1), J. G. Kwak1),K. W. Lee1), S. Wang4), J. S. Yoon1
1) Korea Atomic Energy Research Institute, Taejon, Korea2) Pohang University of Science and Technology, Pohang, Korea3) Pohang Accelerator Laboratory, Pohang, Korea4) Soongsil University, Seoul, Korea
Abstract. The heating and current drive systems of the Korea Superconducting Tokamak Ad-vanced Research (KSTAR) tokamak consist of neutral beam, ion cyclotron, lower hybrid andelectron cyclotron system and are being developed to provide heating as well as current drivecapability for long pulse length up to 300 sec. The systems also provide flexibility in the controlof current density abd pressure profiles for the study of advanced tokamak plasmas. The designhas been completed and the long pulse relevant heating and current drive technologies are beingdeveloped.
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297
(FTP1/03(R)) Development of KSTAR Diagnostics for theAdvanced Tokamak Operation
M. Kwon1), J. G. Bak1), Y. S. Chung1), S. M. Hwang1), S. H. Jeong2), B. C. Kim1), J. Y. Kim1),S. G. Lee1), H. G. Lee1), H. K. Na1), H. L. Yang1), S. J. Yoo1)
1) Korea Basic Science Institute, Korea2) Korea Atomic Energy Research Institute, Korea
Abstract. The KSTAR is aiming at the successful preformance of the advanced tokamak (AT)mode of operation as well as the steady state operation with fully superconducting magnets. Tosupport the current density and pressure profile control experiments for AT operations, devel-opment activities are being concentrated on the profile diagnostics such as the motional Starkeffect (MSE) polarimetry, the Faraday Rotation polarimetry and densiometry, interferometry,and reflectometry. Even for active feedback control of the MHD unstable states of plasmas,extensive sets of magnetic diagnostics are being prepared and tested. The edge and divertorareas are also going to be actively diagnosed to understand the multiply-coupled characteristicswith the AT operations. The present status and activities for the development of the KSTARdiagnostics will be described with test results obtained by utilizing HANBIT mirror plasmas.
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298
(FTP1/04) Investigation of Irradiated Ferroelectric ThinFilms
R. Bittner1), K. Humer1), H. W. Weber1), M. Tyunina2)3), L. Cakare4)2), A. Sternberg2)
1) Atomic Institute of the Austrian Universities, Vienna, Austria2) Institute of Solid State Physics, University of Latvia, Riga, Latvia3) Microelectronics Laboratory and EMPART Research Group of Infotech Oulu, University ofOulu, Finland4) Jozef Stefan Institute, Ljubljana, Slovenia
Abstract. Irradiation effects on highly oriented Pb1Zr0.53Ti0.47O3 (PZT), Pb0.94La0.06Zr0.65Ti0.35O3
(PLZT-6), and Pb1Zr1O3 (PZ) ferroelectric (FE) and antiferroelectric (AF) thin films are in-vestigated with respect to their possible application as a temperature sensitive element in anew bolometer system for ITER. The PZT and PZ films were deposited by a sol-gel techniqueon a Pt/TiO2/Si substrate, whereas the PLZT-6 film was deposited by pulsed laser deposition(PLD) on a LSCO/MgO (100) substrate. The dielectric properties, i.e. the hysteresis loop andthe dielectric constant of the films, were investigated in a frequency range from 20 Hz to 100kHz and at temperatures up to 300 o C, before and after neutron irradiation to a fast neutronfluence of 5×1021m−2 (E > 0.1MeV). The dielectric constant was measured during cooling with2 o C.min−1. The dielectric properties of the films were measured before and after annealing to300 o C.
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299
(FTP1/05) Membrane Technologies for Tritium Recoveringin the Fusion Fuel Cycle
S. Tosti1), L. Bettinali1), C. Rizzello2), V. Violante1)
1) Euratom-ENEA Fusion Association, C. R. ENEA Frascati, 00044 Frascati (RM), Italy2) Tesi Sas, Rome
Abstract. Palladium and palladium-silver permeators have been obtained by coating porousceramic tubes with a thin metal layer. Three coating techniques have been studied and char-acterized: sputtering, chemical electroless deposition and cold-rolling. The Pd-Ag membranesobtained by cold-rolling and annealing of thin metal foils have shown complete hydrogen selec-tivity and chemical and physical stability meeting the requirements of the fuel cycle applications.These rolled membranes have been tested at 300–400 o C with a hydrogen transmembrane pres-sure in the range of 100–280 kPa and hydrogen flow rates up to 2.5× 10−6m3/s. By filling thePd-Ag membranes with a catalyst selective for the water gas shift reaction, membrane reactorshave been obtained for recovering hydrogen isotopes in elemental form from tritiated water.Particularly, a closed-loop process based on a Pd-Ag membrane reactor has been studied for thetritium recovery system of an ITER scale fusion reactor.
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300
(FTP1/06(R)) Development of SiC/SiC Composite for Fu-sion Application
A. Kohyama1), Y. Katoh1), L. L. Snead2), R. H. Jones3)
1) Institute of Advanced Energy, Kyoto University, Kyoto, Japan2) Oak Ridge National Laboratory, Oak Ridge, TN., USA3) Pacific Northwest National Laboratory, Richland, WA., USA
Abstract. The recent efforts to develop SiC/SiC composite materials for fusion applicationunder the collaboration with Japan and the USA are provided, where material performancewith and without radiation damage has been greatly improved. One of the accomplishments isdevelopment of the high performance reaction sintering process. Mechanical and thermal con-ductivity are improved extensively by process modification and optimization with inexpensivefabrication process. The major efforts to make SiC matrix by CVI, PIP and RS methods areintroduced together with the representing baseline properties. The resent results on mechanicalproperties of SiC/SiC under neutron irradiation are quite positive. The composites with newSiC fibers, Hi-Nicalon Type-S, did not exhibit mechanical property degradation up to 10 dpa.Based on the materials data recently obtained, a very preliminary design window is providedand the future prospects of SiC/SiC technology integration is provided.
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301
(FTP1/07(R)) Reduced Activation Ferritic Steel R&D inUS/Japan Collaborative Research
A. K. Kimura1), A. Kohyama1), K. Shiba2), R. L. Klueh3), D. S. Gelles4), G. R. Odette5)
1) Institute of Advanced Energy, Kyoto University, Uji, Kyoto, Japan2) Japan Atomic Energy Research Institute, Tokai-mura, Ibaraki, Japan3) Oak Ridge National Laboratory, Oak Ridge, Tennessee, USA4) Pacific Northwest National Laboratory, Richland, Washington, USA5) University of California, Santa Barbara, California, USA
Abstract. Material performance of reduced activation ferritic steels (RAFS) and their re-sponse to neutron irradiation, which have been investigated by utilizing fission reactors underthe US/Japan collaborative research program (JUPITER), are summarized. Rather high resis-tance to neutron irradiation and helium was recognized for 9Cr-2W RAFS; irradiation hardeningand helium embrittlement of RAFS were evaluated to be much less than for other candidatematerials. Alloy design of high-temperature steels and the development of oxide dispersion-strengthened steels have been progressing.
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302
(FTP1/08) Low Activated Materials as Plasma Facing Com-ponent
T. Hino1), Y. Hirohata1), Y. Yamauchi1), S. Sengoku2)
1) Laboratory for Plasma and Vacuum Science, Hokkaido University, Sapporo, Japan2) JFT-2M Group, Japan Atomic Energy Research Institute, Japan
Abstract. Low activated materials such as ferritic steel, vanadium alloy and SiC/SiC com-posite have to be developed for realization of a fusion demonstration reactor. Major issuesconcerning these low activated materials have been evaluation of neutron irradiation effects andfeasibility as blanket materials. Since these are also in-vessel materials, issues of plasma materialinteractions have to be investigated. Ferritic steel, F82H, is well oxidized in the atmosphere.Thus, pre-baking is necessary before installation. The required baking temperature is higherthan 900 K. Vanadium alloy, V-4Cr-4Ti, absorbs hydrogen well and hydrogen embrittlementtakes place when the hydrogen concentration exceeds a critical level. In order to avoid hydrogenabsorption, the formation of an oxide layer on the alloy was found to be very useful. In JFT-2M,the vanadium alloy was exposed to a deuterium discharge environment for 9 months. On thealloy surface, an oxide deposition layer with a thickness of 200 nm was formed. The deuteriumconcentration observed was very low, only 1.3 wppm. SiC/SiC composite may be employedas divertor plates in addition to its use as blanket material. Fuel hydrogen retention was verysimilar to that of graphite but the chemical erosion was negligibly small.
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303
(FTP1/09) Development of Manufacturing Technology forHigh Purity Low Activation Vanadium Alloys
T. Muroga1), T. Nagasaka1)
1) National Institute for Fusion Science, Toki, Gifu, Japan
Abstract. Vanadium alloys are promising candidate low activation materials for structuralcomponents of fusion reactors. Establishment of industrial infrastructure is, however, remainingto be a critical issue because of lack of other large scale commercial applications. In the presentstudy, technologies for large scale manufacturing of high purity V-4Cr-4Ti alloy were developedby improving the present commercial production processes of vanadium metal, and optimiz-ing alloying, plating, sheeting and wiring techniques. Efforts were focused on reducing carbon,nitrogen and oxygen impurities, which are known to deteriorate workability, weldability andradiation resistance of vanadium alloys. Especially, improvements were made in atmosphericcontrol during calcination, aluminothermic reduction, vacuum arc remelting, and hot forgingand rolling. A medium size (30kg) high purity V-4Cr-4Ti ingot was produced and designated asNIFS-HEAT-1. The specimens produced out of the ingot are being submitted to Round-robintests by Japanese universities. Two larger ingots of 166kg in total weight were produced recently(NIFS-HEAT-2(A) and (B)). By these efforts, technology for fabricating large V-4Cr-4Ti alloyproducts with < 100ppm C, ∼ 100ppm N and 100 ∼ 200ppm O was demonstrated.
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304
(FTP1/10) Optical Fibers for Application in Diagnosticsfor Burning Plasmas
T. Shikama1), T. Kakuta2), T. Nishitani3), S. Kasai3), S. Yamamoto4), N. Shamoto5)
2) Tokai Research Establishment, JAERI, Tokai, Ibaraki, 319-1195 Japan3) Naka Fusion Research Establishment, Naka, Ibaraki, 311-01934) ITER-JCT, Garching, near Munich, D-85748, Germany5) Optics and Electronics Lab., Fujikura Ltd., 285-8550, Sakura, Japan
Abstract. Attempts have been made to develop fused-silica-core optical fibres which can beused in burning plasma fusion devices such as International Thermonuclear Experimental Re-actor (ITER). Nine kinds of silica core optical fibres were developed and irradiated in a highflux fission reactor. A transmission loss of one of fluorine doped optical fibre, F-4, was less than20 dB/m at 630nm even at the end of irradiation where a total ionizing irradiation dose wasabove 109Gy and a fast neutron fluence was above 1023n/m2. These irradiation parameters willbe corresponding to those for the ITER-Engineering-Phase radiation parameters near its coreregion. The results showed that optical fibres could be applied near to burning plasma for visibleas well as infrared diagnostics.
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305
(FTP1/11) Gamma and Proton Induced Degradation in Ce-ramics Materials – A Proposal
B. Constantinescu1)
1) Institute of Atomic Physics, POB MG-6, Bucharest, Romania
Abstract. Ceramic materials will play very important roles in developing fusion reactors, wherethey will be used under heavy irradiation environments (neutrons, gamma-rays, protons, heliumand other ions) for substantial periods for the first time. The programme at the Institute ofAtomic Physics in Bucharest forms a part of the on going ceramics programmes to assess thesuitability of SiO2 based materials for both diagnostic and remote handling application. Theauthors’ proposal focuses on comparison of the ionization and displacement induced damage(influence on the UV and visible optical transmission properties) and on radiation enhanced hy-drogen isotope diffusion in these materials; the work is performed in cooperation with CIEMATMadrid and SCK/CEN Mol. The irradiation facilities are: IRASM – 200 kCi Co-60 source,minimum 2kGy/h, ethanol chlorine benzene and ESR dosimetry; HVEC 8 MV TANDEM –protons up to 16 MeV and 200 nA; and 600 kV DISKTRON – H isotopes up to 600 keV, tensof microamperes.
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306
(FTP1/12) Investigations in the Area of ThermonuclearStructural Material Science in the Republic of Kazakhstan
I. Tazhibayeva1), V. Shestakov1), Yu. S. Cherepnin2)
1) Scientific Research Institute of Experimental and Theoretical Physics of Kazakh State Uni-versity, Almaty, Kazakhstan2) National Nuclear Center, Kurchatov, Kazakhstan
Abstract. The investigations in the area of structural materials for fusion program initiatedwithin the framework of ITER project in the Republic of Kazakhstan are devoted basically inthe following direction: to studying the behaviour of hydrogen isotopes in structural elementsof the first wall and the divertor in conditions simulating real conditions of material operation,accident situations arising during steam interaction with the beryllium armour of the first wallduring accidental coolant loss, to establish an experimental facility for study aspects of tritiumsafety of thermonuclear installations, for example, levels of tritium accumulation and release;efficiency of barrier layers and protective coating; influence of brazing and welding zones ontritium permeation. The work on determination of tritium release from lead/lithium eutecticalloy by mass-spectrometry method and the development of permeation barriers has begun. Atpresent, work has begun to create Kazakhstan’s own tokamak type reactor for investigation ofthe behaviour of various first wall materials and divertor plates during normal and accidentconditions. The concept of spherical tokamak will be used in the construction of KTM reactor.
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307
(FTP1/13) Structural Materials for Fusion Reactors
Abstract. In order to preserve the condition of an environmentally safe machine, present se-lection of materials for structural components of a fusion reactor is made not only on the basisof adequate mechanical properties, behavior under irradiation and compatibility with other ma-terials and cooling media, but also on their radiological properties, i.e. activity, decay heat,radiotoxicity. These conditions strongly limit the number of materials available to a few familiesof alloys, generically known as low activation materials. We discuss the criteria for deciding onsuch materials, the alloys resulting from the application of the concept and the main issues andproblems of their use in a fusion environment.
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308
(FTP1/14) Large Superconducting Conductors and Jointsfor Fusion Magnets: From Conceptual Design to Test atFull Size Scale
D. Ciazynski1), J. L. Duchateau1), P. Decool1), P. Libeyre1), B. Turck1)
1) Association EURATOM-CEA, DRFC, CEA/CADARACHE, France
Abstract. A new kind of superconducting conductor, using the so-called cable-in-conduit con-cept, is emerging mainly involving fusion activity. It is to be noted that at present time no largeNb3Sn magnet in the world is operating using this concept. The difficulty of this technologywhich has now been studied for 20 years, is that it has to integrate major progresses in multipleinterconnected new fields such as: large number (1000) of superconducting strands, high currentconductors (50 kA), forced flow cryogenics, Nb3Sn technology, low loss conductors in pulsedoperation, high current connections, high voltage insulation (10 kV), economical and industrialfeasibility. CEA was very involved during these last 10 years in this development which tookplace in the frame of the NET and ITER technological programs. One major milestone wasreached in 1998–1999 with the successful tests by our Association of three full size conductorand connection samples in the Sultan facility (Villigen, Switzerland).
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309
(FTP1/15) Achieved Capability of the Superconducting Mag-net System for the Large Helical Device
T. Satow1), S. Imagawa1), N. Yanagi1), K. Takahata1), T. Mito1), S. Yamada1), H. Chikaraishi,A. Nishimura1), I. Ohtake1), Y. Nakamura1), S. Satoh1), O. Motojima1)
1) National Institute for Fusion Science, Toki, Japan
Abstract. The Large Helical Device (LHD) is a plasma physics experimental device with amagnetic stored energy of 960 MJ, consisting of two sc (superconducting) helical coils and sixsc poloidal coils. The trial operation and the first plasma discharge of the eight-year Phase Iproject for LHD were finished on 31 March 1998 as initially planned. The second experimentalcampaign was conducted by additional heating using two NBI devices. The third campaignstarted in June 1999 and was finished in January 2000. Many plasma heating tests up to aplasma field of 2.90 T were carried out. Major test results on the sc magnet system for LHD areas follows: (1) The LHD cryogenic system succeeded in 13,400-hour operation and proved itshigh reliability. (2) A central field of 2.91 T at a radius of 3.60 m was achieved at an H-I currentof 11.08 kA, H-M current of 11.83 kA and an H-O current of 12.02 kA. (3) All six poloidal coilswere excited stably. (4) Nine flexible sc bus-lines with a total length of 497 m were operatedstably and safe.
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310
(FTP1/16) Vibration Test Using Sub-scaled Tokamak Modelto Validate Numerical Analysis on Seismic Response of Fu-sion Reactor
N. Takeda1), M. Nakahira1), H. Takahashi1), K. Koizumi1), K. Shimizu2), T. Nakamura2)
1) Japan Atomic Energy Research Institute, Tokai-mura, Japan2) Mitsubishi Heavy Industry, Japan
Abstract. This paper describes the latest status of the fabrication and testing of a sub-scaledtokamak model for the vibration test to validate numerical analysis on seismic response of fusionreactor. The sub-scale model referred to the 1998 ITER design and the scale was decided as tobe 1/8 considering the capacity of test facility and the scaling ratio was chosen so that the stressbecomes equivalent to that of the real machine. The partial test had been performed with thegravity supports and TF coil case, and their vibration characteristics were verified.
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311
(FTP1/17(R)) Progress in Development of Negative IonSources on JT-60U
L. R. Grisham1), M. Kuriyama2), M. Kawai2), T. Itoh2), N. Umeda2), and JT-60U Team2)
1) Princeton University Plasma Physics Laboratory, P. O. Box 451, Princeton, N. J. 08543USA2) Japan Atomic Energy Research Institute, Naka Fusion Research Establishment, Naka-machi,Naka-gun, Ibaraki-ken 311-0193, Japan
Abstract. The negative ion neutral beam system now operating on JT-60U was the first ap-plication of negative ion technology to the production of beams of high current and power forconversion to neutral beams, and has successfully demonstrated the feasibility of negative ionbeam heating systems for ITER and future tokamak reactors. It also demonstrated significantelectron heating and high current drive efficiency in JT-60U. Because this was such a largeadvance in the state of the art with respect to all system parameters, many new physical pro-cesses appeared during the earlier phases of the beam injection experiments. We have exploredthe physical mechanisms responsible for these processes, and implemented solutions for someof them, in particular excessive beam stripping, the secular dependence of the arc and beamparameters, and nonuniformity of the plasma illuminating the beam extraction grid. This hasreduced the percentage of beam heat loading on the downstream grids by roughly a third, andpermitted longer beam pulses at higher powers. Progress is being made in improving the neg-ative ion current density, and in coping with the sensitivity of the cesium in the ion sources tooxidation by tiny air or water leaks, and the cathode operation is being altered.
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312
(FTP1/18(R)) Advanced Negative Ion Beam Technology toImprove the System Efficiency of Neutral Beam Injectors
Y. Okumura1), T. Amemiya1), Y. Fujiwara1), M. Hanada1), M. Kashiwagi1), T. Morishita2),T. Takayanagi3), M. Taniguchi1), K. Watanabe1)
1) Japan Atomic Energy Research Institute, Naka-machi, Ibaraki-ken, 319-0193, Japan 2) KeioUniv., Yokohama, Japan3) Ibaraki Univ., Ibaraki-ken, Japan
Abstract. Technologies producing high current density (> 20mA/cm2) negative ions at a lowoperating gas pressure (< 0.3Pa) has been established. Reduction of the gas pressure is efficientto improve the acceleration efficiency of neutral beam injectors. Beam optics in a multi-aperture,multi-stage electrostatic accelerator has been studied using a 400keV negative ion source. It wasfound that the ion beam tends to expand because of beamlet-beamlet interaction. The effectwas confirmed by 3D beam trajectory simulation code. Beam steering technology by aperturedisplacement has been established to compensate the effect and to focus the beam preciselytoward the injection port, which improves the geometrical efficiency of neutral beam injectors.A plasma neutralizer producing a high-density plasma (1.7 × 1012cm−3) with a high degree ofionization (0.15) has been developed.
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313
(FTP1/19(R)) The Next Step in a Development of NegativeIon Beam Plasma Neutraliser for ITER NBI
V. M. Kulygin1), E. D. Dlougach1), E. P. Gorbunov1), E. Yu. Klimenko1), A. A. Mehed’kin2),A. A. Panasenkov1), Yu. M. Pustovoit1), A. A. Skovoroda1), V. A. Smirnov1), V. A. Zhil’tsov1),V. F. Zubarev1)
1) Institute of Nuclear Fusion, RRC Kurchatov Institute, Moscow, Russia2) Moscow Radio Technical Institute, Moscow, Russia
Abstract. Injectors of deuterium atom beams developing for ITER plasma heating and currentdrive are based on the negative ion acceleration and further neutralization with a gas target.The maximal efficiency of a gas stripping process is 60%. The replacement of the gas neutralizerby plasma one must increase the neutral yield to 80%. The experimental study overview of themicrowave discharge in a multi-cusp magnetic system chosen as a base device for Plasma Neu-tralizer realization and the design development for ITER Neutral Beam Injectors are presented.The experimental results achieved at a plasma neutralizer model PNX-U is discussed. Plasmaconfinement, gas flows, ionization degree were investigated. The plasma in the volume 0.5m3
with density ne ∼ 1018m−3 has been achieved at power density 80kW/m3 in operation withArgon.
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314
(FTP1/20) Development of 100 GHz Band High Power Gy-rotron for Fusion Experimental Reactor
K. Sakamoto1), A. Kasugai1), H. Shoyama1), K. Hayashi1), K. Takahashi1), M. Tsuneoka1),Y. Ikeda1), K. Kajiwara1), S. Moriyama1), M. Seki1), T. Fujii1), T. Kariya2), Y. Mitsunaka2),T. Imai1)
1) JAERI, Naka, Ibaraki, Japan2) Toshiba Co., Otawara, Tochigi, Japan
Abstract. In JAERI, 1MW gyrotrons of 170GHz and 110GHz are under development for ITER(International Thermonuclear Experimental Reactor) and JT-60U, respectively. Both gyrotronshave a depressed collector for an efficiency improvement and a low loss synthetic diamond win-dow that enables Gaussian beam output over 1MW. Three 110GHz gyrotrons are used on anelectron cyclotron heating and current drive(ECH/ECCD) system on JT-60U, in which the out-put power of ∼0.8MW/3sec was generated from each gyrotron. As for 170GHz, output power of1.2MW with electron beam of 85kV/49A was obtained on a short pulse gyrotron. The efficiencyof ∼57% was attained at 1.1MW with the depressed collector. Based on these results, the 1MW170GHz gyrotron for long pulse operation was fabricated.
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315
(FTP1/21) Liquid Lithium Experiments in CDX-U
R. Majeski1), R. Doerner2), R. Kaita1), G. Antar2), J. Timberlake1), J. Spaleta1), D. Hoff-man1), B. Jones1), T. Munsat1), H. Kugel1), G. Taylor1), D. Stutman3), V. Soukhanovskii3),R. Maingi4), S. Molesa5), M. Ulrickson6), P. Efthimion1), J. Menard1), M. Finkenthal3), S. Luck-hardt2), D. G. Whyte2), R. Causey6), D. Buchaneauer6)
1) Princeton Plasma Physics Laboratory, Princeton, NJ USA2) University of California at San Diego, La Jolla, CA USA3) Johns Hopkins University, Baltimore, MD USA4) Oak Ridge National Laboratory, Oak Ridge, TN USA5) Hope College, Holland, MI USA6) Sandia National Laboratory, Alberquerque, NM USA
Abstract. Abstract. The initial results of experiments involving the use of liquid lithium asa plasma facing component in the Current Drive Experiment - Upgrade (CDX-U) are reported.Studies of the interaction of a steady-state plasma with liquid lithium in the Plasma Interactionwith Surface and Components Experimental Simulator (PISCES-B) are also summarized. InCDX-U a solid or liquid lithium covered rail limiter was introduced as the primary limiting sur-face for spherical torus discharges. Deuterium recycling was observed to be reduced, but so farnot eliminated, for glow discharge-cleaned lithium surfaces. Some lithium influx was observedduring tokamak operation. The PISCES-B results indicate that the rates of plasma erosion oflithium can exceed predictions by an order of magnitude at elevated temperatures. Plans to ex-tend the CDX-U experiments to large area liquid lithium toroidal belt limiters are also described.
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316
(FTP1/22) Operating Windows of Pebble Divertor
K. Matsuhiro1), M. Isobe1), Y. Ohtsuka1), Y. Ueda1), M. Nishikawa1)
1) Graduate School of Engineering, Osaka University, Osaka, Japan
Abstract. A marked feature of the pebble divertor is an effect by use of functional multi-layercoated pebble, which consists of a surface plasma facing layer, an intermediate tritium perme-ation barrier layer, and a kernel for heat removal. The dimensions, structure and the irradiationconditions of pebbles are the important issues for the development of the pebble divertor. Fromthe view point of resistance of the induced thermal stress, the pebble is taken as small as pos-sible in size. On the other hand, from the view point of the pumping performance, the suitableirradiation temperature range of the surface layer of pebble was estimated from the experimentsand the numerical analysis. The pumping process enhanced by dynamic retention is available toextend the higher allowable irradiation temperature range from 900K to 1100K. As taking thetemperature rise limitation due to pumping effect and the fractural strength due to the inducedthermal stress limitation, it was found that the diameter of the pebble is possible to be 1–2 mm inabout 20 MW/m2 for the SiC kernel and 2–3 mm in less than 30 MW/m2 for the graphite kernel.
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317
(FTP1/23) Development of a Large-scale Monoblock Diver-tor Mock-up for Fusion Experimental Reactors in JAERI
M. Akiba, S. Suzuki, K. Sato, K. Ezato, M. Taniguchi
Japan Atomic Energy Research Institute, Ibaraki-ken, Japan
Abstract. In a design of fusion experimental reactors, the divertor is required to withstand asteady-state heat load range of 5–10 MW/m2, and a transient heat load of up to 20 MW/m2.JAERI has developed a high thermal conductivity carbon fiber composites (CFCs), reducedactivation brazing techniques, and three layer high strength copper tubes. Based on these tech-nologies, a 1 m long divertor mock-up, which is relevant to a reactor-scale divertor plate, hasbeen fabricated. The mock-up has 30 CFC armors, whose dimension is 30 mm long, 30 mmwide, and 60 mm high. The center of the armor is drilled and directly brazed to the coolingtube, so-called the monoblock type. After heating tests with an intense ion beam, it has beensuccessfully demonstrated that the mock-up withstands a heat load of 5 MW/m2, 30s for 3000cycles, and a heat load of 20 MW/m2, 10s for 1000 cycles without failure.
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318
(FTP1/24) Experimental and Design Activity on LiquidLithium Divertor
V. A. Evtikhin1), I. E. Lyublinski1), A. V. Vertkov1), A. N. Chumanov1), N. M. Afanasiev1),V. N. Shpoliansky1), B. I. Khripunov2), V. B. Petrov2), S. V. Mirnov3), V. M. Korzhavin4),L. G. Golubchikov4)
Abstract. Results of lithium divertor investigation designed for 400 MW power removal understeady-state operation condition (DEMO-S project) are presented and a possibility to solve theproblems of efficient condensation of evaporated lithium, of heat removal, of lithium flow in astrong magnetic field, of heat output, of tritium recovery from lithium and the others is shown.Feasibility of the lithium divertor concept has been confirmed by the results of successful tests oflithium limiter in various modifications in the T-11M tokamak experiments with hydrogen andhelium plasmas and by experimental studies of lithium capillary-pore systems interacting withpulsed plasma and with electron beam in the steady state. Further directions of experimental,calculated and design studies needed for the development and substantiation of the lithium di-vertor concept with capillary-pore systems are analyzed.
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319
(FTP1/25(R)) Overview of the IFE Chamber and TargetTechnologies R&D in the U.S.
W. R. Meier1), M. A. Abdou2), G. L. Kulcinski3), R. W. Moir1), A. Nobile4), P. F. Peterson5),D. A. Petti6), K. R. Schultz7), M. Tillack8), M. Yoda9)
1) Lawrence Livermore National Laboratory, Livermore, CA USA2) University of California, Los Angeles, CA USA3) University of Wisconsin, Madison, WI USA4) Los Alamos National Laboratory, Los Alamos, NM USA5) University of California, Berkeley, CA USA6) Idaho National Engineering and Environmental Laboratory, Idaho Falls, ID USA7) General Atomics, San Diego, CA USA8) University of California, San Diego, CA USA9) Georgia Institute of Technology, Atlanta, GA USA
Abstract. The U.S. Department of Energy, Office of Fusion Energy Science (OFES) formed theVirtual Laboratory for Technology (VLT) to develop the technologies needed to support nearterm fusion experiments and to provide the basis for future magnetic and inertial fusion energypower plants. The scope of the inertial fusion energy (IFE) element of the VLT includes thefusion chamber, driver/chamber interface, target fabrication and injection, and safety and en-vironmental assessment for IFE. Lawrence Livermore National Laboratory, in conjunction withother laboratories, universities and industry, has written an R&D plan to address the criticalissues in these areas over the next 5 years in a coordinated manner. This paper provides anoverview of the U.S. research activities addressing these critical issues.
S. Nakai1), Y. Kozaki1), Y. Izawa1), M. Yamanaka1), T. Kanabe1), Y. Kato1), K. Mima1),H. Nagatomo1), H. Furukawa2), K. Yamamoto1), M. Kabetani1), T. Norimatsu1), K. Nagai1),M. Nakatsuka1), T. Jitsuno1), T. Yamanaka1), C. Yamanaka2)
1) Institute of Laser Engineering, Osaka University, Osaka, Japan2) Institute for Laser Technology, Osaka, Japan
Abstract. Inertial fusion energy (IFE) is becoming feasible due to the increasing understand-ing of implosion physics. Reactor technology issues have begun to be developed. Based on theconceptual design of Laser Driven IFE Power Plant, the technical and physical issues have beenexamined. R&D on key issues that affect the feasibility of power plant have been proceededtaking into account the collaboration in the field of laser driver, fuel pellet, reaction chamberand system design. It is concluded that the technical feasibility of IFE power plant seems to bereasonably high. Coordination and collaboration scheme of reactor technology experts in Japanon Laser Driven IFE Power Plant is being proceeded.
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321
(FTP1/27(R)) Simulation Study of Evacuation and Effectsof Metal Vapor in Laser Fusion Liquid Wall Chamber
Y. Kozaki1), M. Kabetani1), H. Nagatomo1), T. Norimatsu1), H. Furukawa2), K. Yamamoto1),K. Mima1), A. Sunahara1), S. Nakai1), T. Yamanaka1)
1) Institute of Laser Engineering, Osaka University, Osaka, JAPAN2) Institute for Laser Technology, Osaka, JAPAN
Abstract. Simulation studies on evacuation process in liquid wall chamber have been carriedout. The evacuation speed in liquid wall concepts is a critical issue for pulse repetition rate.Then we have analyzed the evacuation process in high vacuum region using DSMC code, whichsolve the Boltzmann equation. The simulation results of the DSMC code show that the effect ofviscosity can’t be neglected in the region where the gas pressure goes down to 10−3 ∼ 10−4 Torr,near the saturated vapor pressure. When the surface temperature of liquid wall is adequatelycooled, the evacuation speed from 10−2 to 10−4 Torr is rather fast because of the gas flow tothe wall. A preliminary estimation on the influence of chamber gas to the performance of a fuelpellet and trajectory of pellet injection is carried out to identify the required level of residualgas pressure.
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322
(FTP1/29) Oxidation of Carbon Based First Wall Materialsof ITER
Abstract. The safety relevance of oxidation reactions on carbon materials in fusion reactorsis discussed. Becau-se tritium codeposited in ITER will probably exceed tolerable limits, coun-termeasures have to be developed: In this paper ozone is tested as oxidising agent for removalof codeposited layers on thick a-C:D-flakes from TEXTOR. In preceeding experiments the ad-vantageous features of using ozonised air instead of ozonised oxygen, reported in literature forreactions with graphite, is not found for nuclear grade graphite. At 185o C = 458 K ozone(0.8–3.4 vol-% in oxygen) is able to gasify the carbon content of these flakes with initial rates,comparable to initial rates in oxygen (21 kPa) for the same material at > 200K higher tem-peratures. The layer reduction rate in ozone drops with increasing burn-off rapidly from about0.9–2.0 µm/h to 0.20–0.25 µm/h, but in oxygen it drops to zero for all temperatures ≤ 450o C = 723 K, before carbon is completely gasified. Alltogether, ozone seems to be a promisingoxidising agent for removal of codeposited layers, but further studies are necessary with respectto rate dependence on temperature and ozone concentration even on other kinds of codepositedlayers. Furtheron, the optimum reaction temperature considering the limited thermal stabilityof ozone has to be found out and studies on the general reaction mechanism have to be done.Besides these examinations on codeposited layers, a short overview on the status of our oxidationstudies on different types of fusion relevant C-based materials is given; open problems in thisfield are outlined.
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323
(FTP1/30) Iron Neutron Data Benchmarking for 14 MeVSource
S. I. Belousov1), Kr. D. Ilieva1)
1) Institute for Nuclear Research and Nuclear Energy of Bulgarian Academy of Science
Abstract. Iron neutron data benchmarking is presented. The iron is widely used as construc-tional and shields material for nuclear fusion and fission systems. The Iron cross section datafrom FENDL/MG-1.0 and VITAMIN-B6 were tested by benchmark experiment of integral neu-tronics’ experiment collection implemented by the IAEA Nuclear Data Section. A discouragingbad consistency of calculated and measured results has been obtained for the neutron leakagefrom iron media. While the discrepancy for the FENDL/MG-1.0 is especially significant in theenergy range 0.5–10 MeV for neutron transmission through media with thickness greater than20 cm, for the VITAMIN-B6 the most inconsistency was obtained in the vicinity of the fusionpeak for neutron transmission through lower thickness media. The presented results demon-strate that the tested Iron muligroup cross section data have to be significantly improved sothat the neutron transport calculation results to be reliable in order to warrant nuclear safetyrequirements.
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324
(FTP1/31) Gyrotrons for Fusion. Status and Prospects.
A. G. Litvak1), V. V. Alikaev1), G. G. Denisov1), V. I. Kurbatov1), V. E. Myasnikov1), E. M. Tai1),V. E. Zapevalov1)
1) Institute of Applied Physics Russian Academy of Sciences, GYCOM Ltd, Nizhny Novgorog,Russia
Abstract. No abstract available
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325
Session FTP2 — EngineeringDesign
Contents
(FTP2/01) Status and Perspectives of Thermal-Hydraulic Analysis of Su-perconducting Magnets for Nuclear Fusion Applications . . . . . . . . . 327
(FTP2/02) The KTM Tokamak and Studies of Construction Materials forThermonuclear Reactors . . . . . . . . . . . . . . . . . . . . . . . . . . . . 328
(FTP2/20) The Impact of Physics Assumptions on Fusion Economics . . . . 344
326
(FTP2/01) Status and Perspectives of Thermal-HydraulicAnalysis of Superconducting Magnets for Nuclear FusionApplications
R. Zanino, L. Savoldi
Dipartimento di Energetica, Politecnico, Torino, Italy
Abstract. An overview is presented of the work performed over the last years at Politec-nico di Torino in the field of thermal-hydraulic modeling of superconducting magnets for fusion,with particular reference to present International Thermonuclear Experimental Reactor (ITER)-related activities, and to future possibilities and needs. The two-fluid Mithrandir code and itsmulti-conductor version, the M&M code, are used together with the cryogenic network solverFlower. We discuss the capabilities of these validated tools to simulate the evolution of thermal-hydraulic transients characterized by widely disparate time scales, from heat slug propagationto quench, both in single dual-channel conductors like that of the QUELL experiment, and inthe ITER model coils (central solenoid CSMC, toroidal field TFMC).
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327
(FTP2/02) The KTM Tokamak and Studies of ConstructionMaterials for Thermonuclear Reactors
E. A. Azizov1), O. I. Buzhinskij1), E. P. Velikhov2), G. G. Gladush1), V. A. Glukhikh3),N. Ya. Dvorkin4), V. N. Dokouka1), I. A. Kovan1), V. Krylov3), I. N. Leykin4), A. Mineev3),N. A. Obysov5), Yu. M. Semenets1), V. S. Shkolnik6), V. Shestakov7), I. Tazhibayeva7), L. N. Tikhomirov8),Yu. S. Cherepnin8), R. Khayrutdinov1), O. G. Filatov3), V. A. Yagnov1)
1) Troitsk Institute for Innovation’s and Thermonuclear Researches, Troitsk, Russia2) Kurchatov Institute, Moscow, Russia3) Efremov Institute, St-Petersburg, Russia4) State Enterprise “Leningradsky Severny Zavod”, St-Petersburg, Russia5) Minatom of RF6) Ministry of fuel, energy, trade, science and technologies of Kazakhstan Republic7) NII ETPh of Kazakh State University, Alma-Ata, Kazakhstan8) National Nuclear Centre of Kazakhstan Republic
Abstract. The Kazakh Tokamak for Material studies (KTM) is designed for modeling plasma-material interactions in divertor region under conditions expected for ITER. KTM is a tokamakwith low aspect ratio A=2. The device enables divertor plates to be changed without distur-bance of vacuum. Optimization of power supply sources of equilibrium coils and inductor hasbeen done. The influence of various plasma parameters (density, temperature, effective charge)and final shape of plasma configuration on volt-second of inductor has been studied. Plasmaequilibrium has been analyzed with respect to vertical stability. The characteristic times of pas-sive stabilization were obtained and “active” coils were selected for active vertical stabilization.Recommendations were given on passive stabilization coils. A system of position and shapecontrol has been designed.
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328
(FTP2/03) Staged Deployment of the International FusionMaterials Irradiation Facility
H. Takeuchi1), M. Sugimoto1), H. Nakamura1), T. Yutani1), M. Ida1), S. Jitsukawa1), T. Kondo1),S. Matsuda1), H. Matsui2), T. E. Shannon3), R. A. Jameson4), F. W. Wiffen5), J. Rathke6),C. Piaszczyk6), S. Zinkle7), A. Moslang8), K. Ehrlich8), E. Daum8), F. Cozzani9), H. Klein10),J.-M. Lagniel11), R. Ferdinand11), M. Martone12), S. Ciattaglia12), V. Chernov13)
1) Japan Atomic Energy Research Institute (JAERI)2) Tohoku University3) The University of Tennessee4) Los Alamos National Laboratory (LANL)5) US Department of Energy6) Advanced Energy Systems Inc.7) Oak Ridge National Laboratory (ORNL)8) Forschungszentrum Karlsruhe (FZK)9) Commission of the EC10) Johann-Wolfgang-Goethe Universitat11) Commissariat a l’Energie Atomique (CEA)/Saclay12) Ente per le Nuove Tecnologie, l’Energia e l’Ambiente (ENEA)13) SSC RF-A. A. Bochvar Institute of Inorganic Materials
Abstract. The International Fusion Materials Irradiation Facility (IFMIF) employs an accel-erator based D-Li intense neutron source as defined in the 1995–96 Conceptual Design Activity(CDA) study. In 1999, IEA mandated a review of the CDA IFMIF design for cost reductionwithout change to its original mission. This objective was accomplished by eliminating thepreviously assumed possibility of potential upgrade of IFMIF beyond the user requirements.The total estimated cost was reduced from $797.2 M to $487.8 M. An option of deploymentin 3 stages was also examined to reduce the initial investment and annual expenditures duringconstruction. In this scenario, full performance is achieved gradually with each interim stageas follows. 1st Stage: 20% operation for material selection for ITER breeding blanket, 2ndStage: 50% operation to demonstrate materials performance of a reference alloy for DEMO, 3rdStage: full performance operation (2MW/m2 @ 500cm3) to obtain engineering data for poten-tial DEMO materials under irradiation up to 100–200 dpa. In summary, the new, reduced costIFMIF design and staged deployment still satisfies the original mission. The estimated cost ofthe 1st Stage facility is only $303.6 M making it financially much more attractive. Currently,IFMIF Key Element Technology Phase (KEP) is underway to reduce the key technology riskfactors.
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329
(FTP2/05) Optimization of Tokamak Poloidal Field Con-figuration by Genetic Algorithms
S. Khorasani1)2), R. Amrollahi1), H. Minoo1)2), F. Dini1)2)
1) Department of Physics, KNT University of Technology, 322 Mirdamad, Tehran, Iran2) Department of Plasma Physics, Nuclear Fusion Research Center, Atomic Energy Organizationof Iran, Amirabad, Tehran, Iran
Abstract. The design of the poloidal field system in tokamaks is a lengthy procedure dueto functional dependence of design parameters on pre-assumed plasma parameters. Thereforechanging an individual design parameter could in principle influence the overall design. Thiscomplicated task could easily lead to non-optimal final results. In the paper it is shown thatgenetic algorithms, as powerful optimization methods, are useful in approaching an optimal de-sign. The work may be extended to the whole plasma scenario and magnetic configuration inother small and large scale systems.
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330
(FTP2/06) Design of Next Step Tokamak: Consistent Anal-ysis of Plasma Perfomance, Flux Consumption and PoloidalField System
J. M. Ane1), V. Grandgirard2), F. Albajar1), J. Johner1)
Abstract. A consistent and simple approach to derive plasma scenarios for next step tokamakdesign is presented. It is based on successive plasma equilibria snapshots from plasma breakdownto end of ramp-down. Temperature and density profiles for each equilibrium are derived from a2D plasma model. The time interval between two successive equilibria is then computed fromthe toroidal field magnetic energy balance, the resistive term of which depends on n, T profiles.This approach provides a consistent analysis of plasma performance, flux consumption and PFsystem, including average voltages waveforms across the PF coils. The plasma model and thePoynting theorem for the toroidal magnetic energy are presented. Application to ITER-FEATand to M2, a Q=5 machine designed at CEA, are shown.
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331
(FTP2/07) Edge-Plasma Analysis for Liquid-Wall MFE Con-cepts
R. W. Moir1), M. E. Rensink1), T. D. Rognlien1)
1) University of California Lawrence Livermore National Laboratory
Abstract. A thick flowing layer of liquid (e.g., flibe – a molten salt, or Sn80Li20 – a liquid metal)protects the structural walls of the magnetic fusion configuration so that they can last the life ofthe plant even with intense 14 MeV neutron bombardment from the D-T fusion reaction. Thesurface temperature of the liquid rises as it passes from the inlet nozzles to the exit nozzles dueto absorption of line and bremsstrahlung radiation, and neutrons. The surface temperature canbe reduced by enhanced turbulent convection of hot surface liquid into the cooler interior. Thissurface temperature is affected by the temperature of liquid from a heat transport and energyrecovery system. The evaporative flux from the wall driven by the surface temperature mustalso result in an acceptable impurity level in the core plasma. The shielding of the core by theedge plasma is modeled with a 2D-transport code for the DT and impurity ions; these impurityions are either swept out to the divertor, or diffuse to the hot plasma core. An auxiliary plasmabetween the edge plasma and the liquid wall may further attenuate evaporating flux of atomsand molecules by ionization near the wall.
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332
(FTP2/08) Compact Stellarator Coils
N. Pomphrey1), L. A. Berry2), A. H. Boozer3), A. Brooks1), R. Hatcher1), S. Hirshman2),L. P. Ku1), W. Miner4), H. Mynick1), W. Reiersen1), D. Strickler2), P. Valanju4)
1) Princeton Plasma Physics Laboratory, Princeton, NJ 08543, USA2) Oak Ridge National Laboratory, Oak Ridge, TN 37831-8070, USA3) Columbia University, New York, NY 10027, USA4) University of Texas at Austin, Austin, TX 78712-1081, USA
Abstract. Experimental devices to study the physics of high-beta (β & 4%), low aspect ratio(A . 4.5) stellarator plasmas require coils that will produce plasmas satisfying a set of physicsgoals, provide experimental flexibility, and be practical to construct. In the course of design-ing a flexible coil set for the National Compact Stellarator Experiment, we have made severalinnovations that may be useful in future stellarator design efforts. These include: the use ofSingular Value Decomposition methods for obtaining families of smooth current potentials ondistant coil winding surfaces from which low current density solutions may be identified; theuse of a Control Matrix Method for identifying which few of the many detailed elements of thestellarator boundary must be targeted if a coil set is to provide fields to control the essentialphysics of the plasma; the use of Genetic Algorithms for choosing an optimal set of discrete coilsfrom a continuum of potential contours; the evaluation of alternate coil topologies for balancingthe tradeoff between physics objective and engineering constraints; the development of a newcoil optimization code for designing modular coils, and the identification of a “natural” basis fordescribing current sheet distributions
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333
(FTP2/09) Advanced Study of Tokamak Transmutation Sys-tem
L. J. Qiu1), Y. C. Wu1), B. Wu1), X. P. Liu1), Y. P. Chen1), W. N. Xu1), Q. Y. Huang1)
1) Institute of Plasma Physics, Chinese Academy Science
Abstract. An advanced tokamak transmutation system is proposed as an alternative applica-tion of fusion energy based on a review of the previous studies. This system includes: (1) a lowaspect ratio tokamak as fusion neutron driver, (2) a Radioactivity Clean Nuclear Power System(RCNPS) as blanket, (3) a novel concept of liquid metal center conductor post (CCP) as partof toroidal field coils. A preliminary feasibility study has been carried out for the system, whichincluded the aspects of core plasma physics, blanket neutronics and design. A driver of 100MWfusion power under 1MW/m2 neutron wall loading can transmute the amount of High LevelWaste (including minor actinides and fission products) produced by 10 standard pressurizedwater reactors of 1 GW electrical power output. Meanwhile, the system can produce tritiumand output about 2 GW of electrical energy. After 30 years of operation , the biological hazardpotential (BHP) level of the whole system will decrease by 2 orders of magnitude.
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334
(FTP2/10) High-Frequency Operation of the Dynamic Er-godic Divertor (DED)
K. H. Finken1), S. S. Abdullaev1), Th. Eich1), D. W. Faulconer2), M. Kobayashi1)3), R. Koch2),G. Mank1), A. L. Rogister1)
1) Institut fur Plasmaphysik, Forschungszentrum Julich GmbH, EURATOM Association, D-52425 Julich, Germany, Partner in the Trilateral Euregio Cluster2) Ecole Royal Militaire-Koninklijke Militaire School, Association EURATOM-Belgian State,Brussels, Belgium, Partner in the Trilateral Euregio Cluster3) Department of Energy Engineering and Science, Nagoya University, Nagoya, Japan
Abstract. After the introduction of the experimental options with the DED and after a discus-sion of the static aspects of the ergodic and laminar zones, the dynamic features of the rotatingDED field are emphasised. The rotating perturbation field induces a shielding current which ismodelled under different assumptions. A result of the interaction of the shielding current withthe external one is the transfer of a torque from the DED-coils to the plasma. The location ofthe maximum of this transfer function with respect to the frequency depends critically on thewidth of the shielding current. From the transferred torque, at first the toroidal velocity of theplasma is computed and from the poloidal force component, the radial electric field is estimatedassuming a revisited neoclassical theory.
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335
(FTP2/11) The FTU-D Project
E. Barbato1), A. Bruschi1), G. Candela1), R. Cesario1), P. Chuillon1), S. Cirant, R. Clae-sen1), V. Cocilovo1), A. Coletti, C. Crescenzi1), F. Crisanti1), A. Cucchiaro1), R. De Angelis1),G. Fermani1), M. Gasparotto1), C. Gormezano2), G. Granucci2), E. Lazzaro1), G. Maddaluno1),M. Marinucci1), G. Mazzitelli2), S. Novak1), S. Papastergiou1), V. Pericoli1), L. Pieroni2),G. Ramponi1), M. Roccella1), F. Romanelli1), M. Santinelli1), L. Semeraro2), C. Sozzi1), F. Starace1),A. A. Tuccillo1), O. Tudisco1), V. Vitale1), L. Zannelli3), R. Albanese3), G. Ambrosino3), M. Ar-iola3), A. Pironti3), F. Villone1), G. Vlad
1) EURATOM-ENEA Fusion Association, Frascati, Italy2) Istituto di Fisica del Plasma CNR Milano, Italy3) CREATE
Abstract. A modification of the FTU tokamak (toroidal field BT = 8T, plasma current IP = 1.6MA,minor radius a = 0.3m, major radius R = 0.93m) is proposed in order to extend the FTU op-eration to strongly shaped plasmas (FTU-D: R = 1m, a = 0.18− 0.2m, elongation κ = 1.6,triangularity δ up to δ = 0.8). FTU has a circular vacuum vessel and was built to producecircular plasmas, however unbalancing the currents in the windings of the air core transformera plasma shaping can be produced. Single Null (SN) and Double Null (DN) equilibria havebeen studied with a maximum current in the range 0.350–0.450 MA. The scientific aim of theproject is the investigation of the advanced tokamak operation, characterised by the simulta-neous achievement of high βN (normalised beta) and high bootstrap current fraction (fB) inregimes with high-energy confinement obtained by current and pressure profile control. Themain features of FTU-D, with respect to other existing tokamaks, are the high magnetic field(BT = 5− 2.5T), the high density and aspect ratio value (A = R/a = 5–6) and the possibilityof investigating regimes with dominant electron heating.
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336
(FTP2/12) Helical Reactor Design Studies Based on NewConfinement Scalings
K. Yamazaki1), K. Y. Watanabe1), A. Sagara1), H. Yamada1), S. Sakakibara1), K. Narihara1),K. Tanaka1), M. Osakabe1), K. Nishimura1), O. Motojima1), M. Fujiwara1), the LHD Group
1) National Institute for Fusion Science, Oroshi-cho, Toki, Japan
Abstract. The design requirements for helical reactors are investigated on plasma confinement,density regime and beta limit, comparing with recent LHD (Large Helical Device) experimen-tal data. Several new confinement scaling laws are derived using LHD database in additionto the medium-sized helical system database. In the previous reactor designs two times bet-ter plasma confinement than the conventional LHD scaling law was assumed, which has beenachieved experimentally as “New LHD” scaling laws. One and half times higher density thanthe conventional helical density limit scaling laws has been achieved, which condition is requiredat the start-up phase of reactors. Half of beta value required in reactors is achieved in theinward-shifted configuration in LHD experiment, which value is beyond the theoretical stabilitylimit. This magnetic configuration satisfies the high beta and low effective helical ripple opera-tion required for reactors. Almost these normalized requisites have been achieved in the LHDexperiment Based on new LHD scaling laws the reactor system design has been carried out. TheCOE (cost of electricity) value of large reactor system is not so high in comparison with thatof compact design, however the compact reactor has advantage of rather lower direct cost. Thepresent LHD experiment can justify the future prospect of the LHD-type helical devices towardsa steady-state efficient and reliable reactor.
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337
(FTP2/13) Study on the Steady-state Tokamak Reactorwith Combined Heating and Current Drive
I. Senda1), H. Takase2), T. Shoji1), M. Araki1), T. Tsunematsu1)
1) Naka Fusion Research Establishment, Japan Atomic Energy Research Institute, Ibaraki, Japan2) Power and Industrial Systems Research and Development Center, Toshiba Corp., Kanagawa,Japan
Abstract. The control of burning plasma in the compact ITER is investigated by analyses ofthe heating and current drives combined with the 1.5D transport simulations. The transportcoefficients are determined by analyses in ELMy H mode consistently with IPB98(y,2) scalingand the shear-dependent reduction of diffusivity is introduced in the reversed shear (RS) mode.The heating and current drive by the neutral beam injection and the electron cyclotron wave areconsidered. The long-pulse operations in ELMy H mode with Q = 4.2 and higher are obtained.The loop voltage in the long-pulse operation is less than 0.02V, which corresponds to the burntime 2500 seconds. The steady state operations in RS mode are investigated and operationswith Q > 6.5 are obtained. The thermal energy confinement time in RS mode increases morethan 30% compared with IPB98(y,2) scaling due to the formation of the internal transport bar-rier. The position of the transport barrier moves toward the plasma edge as the fusion poweris increased. It is found that the burning plasma in RS mode has characteristics to organize itsprofiles by the balance among itself. For both ELMy H and RS modes, the local heating at theregion of low diffusivity improves the performance of the plasma. The possibility to obtain highfusion power (> 1GW) in RS mode of the compact ITER is discussed.
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338
(FTP2/14) Conceptual Design of Advanced Steady-stateTokamak Reactor (A-SSTR2) – Compact and Safety Ori-ented Commercial Power Plant
S. Nishio, K. Ushigusa, S. Ueda, et al.
Japan Atomic Energy Research Institute
Abstract. Based on the last decade JAERI reactor design studies, the advanced commercialreactor concept (A-SSTR2) which meets both economical and environmental requirements hasbeen proposed. The A-SSTR2 is a compact power reactor (Rp = 6.2m, ap = 1.5m, Ip = 12MA)with a high fusion power (Pf = 4GW) and a net thermal efficiency of 51%. The machine config-uration is simplified by eliminating a center solenoid (CS) coil system. SiC/SiC composite forblanket structure material, helium gas cooling with pressure of 10MPa and outlet temperatureof 900o C, and TiH2 for bulk shield material are introduced. For the toroidal field (TF) coil, ahigh temperature (TC) superconducting wire made of bismuth with the maximum field of 23Tand the critical current density of 1000A/mm2 at a temperature of 20K is applied. In spite ofthe CS-less configuration, a computer simulation gives a satisfactory plasma equilibria, plasmainitiation process and current ramp up scenario.
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339
(FTP2/15) ARIES-AT: An Advanced Tokamak, AdvancedTechnology Fusion Power Plant
F. Najmabadi1), S. C. Jardin2), M. Tillack1), L. M. Waganer3)
1) University of California, San Diego, La Jolla, CA, USA2) Princeton Plasma Physics Laboraotry, Princeton, NJ, USA3) Boeing High Energy Systems, St. Louis, MI, USA
Abstract. The ARIES-AT study was initiated to assess the potential of high-performancetokamak plasmas together with advanced technology in a fusion power plant. Several avenueswere pursued in order to arrive at plasmas with a higher β and better bootstrap alignmentcompared to ARIES-RS that led to plasmas with higher βN and β. Advanced technologiesthat are examined in detail include: (1) Possible improvements to the overall system by usinghigh-temperature superconductors, (2) Innovative SiC blankets that lead to a high thermal cycleefficiency of ∼60%; and (3) Advanced manufacturing techniques which aim at producing near-finished products directly from raw material, resulting in low-cost, and reliable components. The1000-MWe ARIES-AT design has a major radius of 5.4 m, minor radius of 1.3 M, a toroidal βof 9.2% (βN = 6.0) and an on-axis field of 5.6 T. The plasma current is 13 MA and the currentdrive power is 24 MW. The ARIES-AT study shows that the combination of advanced tokamakmodes and advanced technology leads to attractive fusion power plant with excellent safety andenvironmental characteristics and with a cost of electricity (5c/kWh), which is competitive withthose projected for other sources of energy.
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340
(FTP2/16) Mission and Design of the Fusion Ignition Re-search Experiment (FIRE)
D. M. Meade1), S. C. Jardin1), J. Schmidt1), R. Thome2), N. R. Sauthoff1), P. Heitzenroeder1),B. E. Nelson3), M. Ulrickson4), C. Kessel1), J. Mandrekas5), C. Neumeyer1), J. H. Schultz2),P. Rutherford1), J. C. Wesley6), K. M. Young1), W. M. Nevins7), W. A. Houlberg3), N. A. Uckan3),R. W. Woolley1) and C. C. Baker8)
1) Princeton Plasma Physics Laboratory, Princeton, NJ 08543, USA2) Massachusetts Institute of Technology, Cambridge, MA 02139, USA3) Oak Ridge National Laboratory, Oak Ridge, TN 37831, USA4) Sandia National Laboratory, Albuquerque, NM 87185, USA5) Georgia Institute of Technology, Atlanta, GA 30332, USA6) General Atomics, San Diego, CA 92186, USA7) Lawrence Livermore National Laboratory, Livermore, CA 94551, USA8) University of California at San Diego, San Diego, CA 92093, USA
Abstract. Experiments are needed to test and extend present understanding of confinement,macroscopic stability, alpha-driven instabilities, and particle/power exhaust in plasmas dom-inated by alpha heating. A key issue is to what extent pressure profile evolution driven bystrong alpha heating will act to self-organize advanced configurations with large bootstrap cur-rent fractions and internal transport barriers. A design study of a Fusion Ignition ResearchExperiment (FIRE) is underway to assess near term opportunities for advancing the scientificunderstanding of self-heated fusion plasmas. The emphasis is on understanding the behavior offusion plasmas dominated by alpha heating (Q ≥ 5) that are sustained for durations comparableto the characteristic plasma time scales (≥ 20τE and ∼ τskin, where τskin is the time for theplasma current profile to redistribute at fixed current). The programmatic mission of FIREis to attain, explore, understand and optimize alpha-dominated plasmas to provide knowledgefor the design of attractive magnetic fusion energy systems. The programmatic strategy is toaccess the alpha-heating-dominated regime with confidence using the present advanced tokamakdata base (e.g., Elmy-H-mode, ≤ 0.75 Greenwald density) while maintaining the flexibility foraccessing and exploring other advanced tokamak modes (e. g., reversed shear, pellet enhancedperformance) at lower magnetic fields and fusion power for longer durations in later stages ofthe experimental program. A major goal is to develop a design concept that could meet thesephysics objectives with a construction cost in the range of $1B.
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341
(FTP2/17) Toroidal Reactor Designs as a Function of As-pect Ratio
C. P. C. Wong1), J. C. Wesley1), R. D. Stambaugh1), and E. T. Cheng2)
1) General Atomics, San Diego, California USA2) TSI Research Inc., Solana Beach, California USA
Abstract. This paper presents a “common basis” systems study of superconducting (SC)and normal-conducting (NC) DT-burning fusion power and materials testing reactor designs.Figures-of-merit for power and materials-testing reactors are respectively; projected cost-of-electricity (COE) and direct cost (DC). A common 0-D plasma modeling basis is used and theplasma geometry and engineering aspects of the SC and NC designs are treated in an equivalentmanner that is consistent with the limitations of their respective magnet technologies and otherdesign constraints. Aspect ratios A in the range 1.2 ≤ A ≤ 6 and plasma elongations in therange 1 ≤ κ ≤ 3 are explored and a MHD stability (beta limit) physics basis that accuratelydescribes the increase of normalized beta βN and toroidal beta βT with a decreasing A and/orincreasing κ is incorporated. With this MHD basis taken into account and with the usual reac-tor geometry, physics and engineering constraints and costing bases applied, the results of thestudy show that for power reactors the minimum COE is pointing towards lower A ∼ 2 thangenerally found in previous studies. The minimum is broader with higher κ. For test reactorswith similar fusion power output, the direct cost for NC options is significantly lower than forSC coil options. With the NC category, testing designs that combine intermediate A and higherelongation show promise as a D-T burn next step device that could provide scientific and testingdata to support future SC and NC reactors. For example, a NC coil design with A ∼ 2, κ = 3could produce 200 MW fusion power at 1.23MW/m2 average neutron wall loading at a totaldirect cost of about $643 M. This NC design with a fissile blanket could also convert ∼ 1270kgof fission reactor waste per full power year.
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342
(FTP2/18) Compact Stellarators as Reactors
J. F. Lyon1), P. Valanju2), M. C. Zarnstorff3), S. Hirshman1), D. A. Spong1), D. Strickler1),A. Ware4), D. E. Williamson1)
1) Oak Ridge National Laboratory, Oak Ridge, TN, U.S.A.2) University of Texas at Austin, Austin, TX, U.S.A.3) Princeton Plasma Physics Laboratory, Princeton, NJ, U.S.A.4) University of Montana, Missoula, MT, U.S.A.
Abstract. Two types of compact stellarators are examined as reactors: two- and three-field-period (M = 2 and 3) quasi-axisymmetric devices with volume-average 〈β〉 = 4− 5% and M = 2and 3 quasi-poloidal devices with 〈β〉 = 10 − 15%. These low-aspect-ratio stellarator-tokamakhybrids differ from conventional stellarators in their use of the plasma-generated bootstrap cur-rent to supplement the poloidal field from external coils. Using the ARIES-AT model withBmax = 12T on the coils gives Compact Stellarator reactors with R = 7.3−8.2m, a factor of 2–3smaller R than other stellarator reactors for the same assumptions, and neutron wall loadingsup to 3.7MWm−2.
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343
(FTP2/20) The Impact of Physics Assumptions on FusionEconomics
D. Ward1), I. Cook1), P. J. Knight1)
1) EURATOM/UKAEA Fusion Association, Culham Science Centre, Abingdon, OX14 3DB,UK.
Abstract. The development of fusion promises a long term supply of energy with widespreadresources and good safety and environmental properties. However the introduction of fusioninto the future energy market will rely on the development of an economically viable fusionpower plant. Although predictions of the likely cost of electricity produced by a future fusionpower plant are uncertain, it is important that an assessment is made to ensure that the likelyeconomics are not unreasonable. In this paper the impact of different physics (and other) con-straints on the economics of fusion is considered. Comparison with the expected future cost ofelectricity from other sources must take account of the trends in the energy market, particularlyat present towards sources with low external costs related to impact on human health and thenatural environment. Although these costs depend on the country concerned, a range of ex-pected future costs can be derived. Comparison with the expected range of fusion costs showsthat fusion can contribute to the future energy market.
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344
Session SEP — Safety &Environment
Contents
(SEP/01) Study of Decay Heat Removal and Structural Assurance by LBBConcept of Tokamak Components . . . . . . . . . . . . . . . . . . . . . . 346
(SEP/02) Experimental Verification of Integrated Pressure Suppression Sys-tems in Fusion Reactors at In-Vessel Loss-of-Coolant Events . . . . . . 347
(SEP/03) Studies on Nuclear Fusion Energy Potential Based on a Long-termWorld Energy and Environment Model . . . . . . . . . . . . . . . . . . . 348
(SEP/01) Study of Decay Heat Removal and Structural As-surance by LBB Concept of Tokamak Components
Y. Neyatani1), D. Tsuru1), M. Nakahira2), T. Araki1), F. Araya2) and K. Nomoto1)
1) Naka Fusion Research Establishment, Japan Atomic Energy Research Institute 801-1Mukoyama,Naka-machi, Naka-gun, Ibaraki-ken, JAPAN2) Tokai Research Establishment, Japan Atomic Energy Research Institute 2-4Shirakata, Tokai-mura, Naka-gun, Ibaraki-ken, JAPAN
Abstract. Since decay heat density in ITER is quite low, thermal analyses have shown thatonly natural dissipation due to thermal radiation can be sufficient for removal of decay heateven in loss of all coolant. Owing to this attractiveness, no cooling system would be requiredfor decay heat removal. In addition, because a magnetically confined plasma terminates by asmall amount of impurity ingress, there is no possibility of uncontrolled production of energy,which will damage the integrity of the vacuum vessel containing tritium and other radioactivematerials. This statement can be assured with a high level of confidence resulted from the LBB(Leak Before Break) concept.
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346
(SEP/02) Experimental Verification of Integrated PressureSuppression Systems in Fusion Reactors at In-Vessel Loss-of-Coolant Events
K. Takase1), H. Akimoto1)
1) Japan Atomic Energy Research Institute
Abstract. An integrated ICE (Ingress-of-Coolant Event) test facility was constructed to demon-strate that the ITER safety design approach and design parameters for the ICE events are ad-equate. Major objectives of the integrated ICE test facility are: to estimate the performanceof an integrated pressure suppression system; to obtain the validation data for safety analysiscodes; and to clarify the effects of two-phase pressure drop at a divertor and the direct-contactcondensation in a suppression tank. A scaling factor between the test facility and ITER-FEATis around 1/1600. The integrated ICE test facility simulates the ITER pressure suppressionsystem and mainly consists of a plasma chamber, vacuum vessel, simulated divertor, relief pipeand suppression tank. From the experimental results it was found quantitatively that the ITERpressure suppression system is very effective to reduce the pressurization due to the ICE event.Furthermore, it was confirmed that the analytical results of the TRAC-PF1 code can simulatethe experimental results with high accuracy.
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347
(SEP/03) Studies on Nuclear Fusion Energy Potential Basedon a Long-term World Energy and Environment Model
K. Tokimatsu1), J. Fujino2), Y. Asaoka3), Y. Ogawa4), K. Okano3), T. Yoshida3), R. Hiwatari3),S. Konishi5), S. Nishio5), K. Yamaji4), and Y. Kaya1)
1) Research Institute of Innovative Technology for the Earth (RITE), Tokyo, Japan2) National Institute for Environmental Studies (NIES), Tsukuba, Japan3) Central Research Institute of Electric Power Industry (CRIEPI), Tokyo, Japan4) The University of Tokyo (UT), Tokyo, Japan5) Japan Atomic Energy Research Institute (JAERI), Naka, Japan
Abstract. This study investigates introduction conditions and potential of nuclear fusion en-ergy as energy supply and CO2 mitigation technologies in the 21st century. Time horizon ofthe 21st century, 10 regionally allocated world energy/environment model (Linearized DynamicNew Earth 21) is used for this study. Following nuclear fusion technological data are takeninto consideration: cost of electricity (COE) in nuclear fusion introduction year, annual COEreduction rates, regional introduction year, and maximum regional plant capacity constraints bymaximum plant construction speed. We made simulation under a constraint of atmospheric CO2
concentration of 550 parts per million by volume (ppmv) targeted at year 2100, assuming thatsequestration technologies and unknown innovative technologies for CO2 reduction are available.The results indicate that under the 550ppm scenario with nuclear fusion within maximum con-struction speed, 66mill/kWh is required for introducing nuclear fusion in 2050, 92 mill/kWh in2060, and 106 mill/kWh in 2070. Therefore, tokamak type nuclear fusion reactors of presentseveral reactor cost estimates are expected to be introduced between 2060 and 2070, and elec-tricity generation fraction by nuclear fusion will go around 20% in 2100 if nuclear fusion energygrowth is limited only by the maximum construction speed. CO2 reduction by nuclear fusionintroduced in 2050 from business-as-usual (BAU) scenario without nuclear fusion is about 20%of total reduction amount in 2100. In conclusion, nuclear fusion energy is revealed to be one ofthe candidates of energy supply technologies and CO2 mitigation technologies. Cost competi-tiveness and removal of capacity constraint factors are desired for use of nuclear fusion energyin a large scale.
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348
(SEP/04) Economic and Environmental Performance of Fu-ture Fusion Plants in Comparison
T. Hamacher1), R. M. Saez1), P. Lako5), H. Cabal1), B. Hallberg2), R. Korhonen4), Y. Lechon1),S. Lepicard8), L. Schleisner6), T. Schneider8), D. Ward7), J. R. Ybema5), G. Zankl3)
1) CIEMAT,Spain2) Studsvik Eco & Safety AB, SE-611 82 Nykoping, Sweden3) IPP, Germany4) VTT Energy, Finland,5) ECN Netherlands Energy Research Foundation, Netherlands6) RISO National Laboratory, Denmark7) UKAEA, Abingdon, UK8) CEPN-Association CEA-EURATOM, Fontenay-aux-Roses, France
Abstract. If the good performance of fusion as technology with no CO2 emission during normaloperation and rather low external costs, reflecting the advantageous environmental and safetycharacteristics, are considered in future energy regulations, fusion can win considerable marketshares in future electricity markets. The economic performance was elaborated for WesternEurope for the time period till 2100. The software tool MARKAL widely used in energy re-search was used to simulate and optimise the development of the Western European energysystem. Two different scenarios were considered, the main difference was the interest rate forinvestments. Stringent CO2-emission strategies lead to considerable market shares for fusion.As a comprehensive indicator of the environmental and safety performance of fusion plants theexternal costs following the ExternE method was used. External costs of fusion are rather low,much below the cost of electricity, and are in the same range as photovoltaics and wind energy.
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349
Session TH2 — Turbulence,Flows, Streamers
Contents
(TH2/1) Secondary Instability in Drift Wave Turbulence as a Mechanismfor Avalanche and Zonal Flow Formation . . . . . . . . . . . . . . . . . . 351
(TH2/2) Bursty Transport in Tokamaks with Internal Transport Barriers . 352
(TH2/3) Role of Zonal Flow in Turbulent Transport Scalings . . . . . . . . . 353
(TH2/4) Nonlinear Features of the Electron Temperature Gradient Modeand Electron Thermal Transport in Tokamaks . . . . . . . . . . . . . . . 354
(TH2/1) Secondary Instability in Drift Wave Turbulence asa Mechanism for Avalanche and Zonal Flow Formation
P. H. Diamond1), S. Champeaux1), M. Malkov1), A. Das1)2), I. Gruzinov1), M. N. Rosenbluth1),C. Holland1), B. Wecht1), A. Smolyakov3), F. L. Hinton4), Z. Lin5), T. S. Hahm5)
1) University of California, San Diego, La Jolla, CA 92093-0319 USA2) Permanent Address: Institute For Plasma Research, Bhat, Gandhinagar 382428 India3) University of Saskatchewan, Saskatoon, SK S7N 5E2, Canada4) General Atomics, San Diego, CA 92138-5608 USA5) Princeton Plasma Physics Laboratory, Princeton, NJ 08543-0451
Abstract. We report on recent developments in the theory of secondary instability in drift-ITGturbulence. Specifically, we explore secondary instability as a mechanism for avalanche forma-tion. A theory of radially extended streamer cell formation and self-regulation is presented.Aspects of streamer structure and dynamics are used to estimate the variance of the drift-waveinduced flux. The relation between streamer cell structures and the avalanche concept is dis-cussed, as are the implications of our results for transport modeling.
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351
(TH2/2) Bursty Transport in Tokamaks with Internal Trans-port Barriers
S. Benkadda1), O. Agullo1), P. Beyer1), N. Bian1), P. H. Diamond3), C. Figarella1), X. Garbet2),P. Ghendrih2), V. Grandgirard1), Y. Sarazin2)
1) CNRS – Universite de Provence, Marseille, France2) Association EURATOM – CEA, St. Paul-lez-Durance, France3) University of California, La Jolla, CA
Abstract. Large scale transport events are studied using two different three dimensional sim-ulation codes related to resistive ballooning and ion temperature gradient turbulence. Theturbulence is driven by a constant incoming flux. In the case of resistive ballooning simulations,the underlying structures are found to be radially elongated at the low field side, and distortedby magnetic shear in parallel direction (streamers). The non linear character of these structuresis emphasized. Bursty transport is investigated in presence of zonal flows and internal trans-port barriers generated either by a strong shear flow or with a magnetic shear reversal. Finally,a low dimensional model that captures the main features of bursty transport dynamics is derived.
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352
(TH2/3) Role of Zonal Flow in Turbulent Transport Scal-ings
Z. Lin1), T. S. Hahm, J. A. Krommes, W. Lee, J. Lewandowski, H. Mynick, H. Qin, G. Rewoldt,W. M. Tang, R. White
1) Princeton Princeton Plasma Physics Laboratory, Princeton, NJ 08543, USA
Abstract. Transport scalings with respect to collisionality (ν∗) and device size (ρ∗) are obtainedfrom massively parallel gyrokinetic particle simulations of electrostatic toroidal ion-temperature-gradient (ITG) turbulence in the presence of zonal flows. Simulation results show that ionthermal transport from electrostatic ITG turbulence depends on ion–ion collisions due to theneoclassical damping of self-generated E×B zonal flows that regulate the turbulence. Fluc-tuations and heat transport level exhibit bursting behavior with a period corresponding to thecollisional damping time of poloidal flows. Results from large-scale full torus simulations withdevice-size scans of realistic parameters show that Bohm-like transport can be driven by micro-scopic scale fluctuations in the ITG turbulence with isotropic spectra. These simulation resultsresolve some apparent physics contradictions between experimental observations and turbulenttransport theories.
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353
(TH2/4) Nonlinear Features of the Electron TemperatureGradient Mode and Electron Thermal Transport in Toka-maks
P. K. Kaw1), R. Singh1),J. G. Weiland2)
1) Institute for Plasma Research, Bhat, Gandhinagar 382428, India2) Department of Electromagnetics, Chalmers University ofTechnology, S-41296, Gothenburg, Sweden
Abstract. Analytical investigations of several linear and nonlinear features of ETG turbulenceare reported. The linear theory includes effects such as finite beta induced electromagneticshielding, coupling to electron magnetohydrodynamic modes like whistlers etc. It is argued thatnonlinearly, turbulence and transport are dominated by radially extended modes called ‘stream-ers’. A nonlinear mechanism generating streamers based on a modulational instability theoryof the ETG turbulence is also presented. The saturation levels of the streamers using a KelvinHelmholtz secondary instability mechanism are calculated and levels of the electron thermaltransport due to streamers are estimated.
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354
(TH2/5) Gyrokinetic Simulations of Tokamak Microturbu-lence
W. Dorland1), B. N. Rogers1), F. Jenko2), M. Kotschenreuther3), G. W. Hammett4), D. R. Mikkelsen4),D. W. Ross5), M. A. Beer4), P. B. Snyder6), R. Bravenec5), M. J. Greenwald7), D. R. Ernst4),R. Budny4)
1) University of Maryland, College Park, MD, 20742, USA2) Max-Planck-Institut fur Plasmaphysik, EURATOM, 85748 Garching, Germany3) Institute for Fusion Studies, University of Texas, Austin, TX, 78712, USA4) Princeton Plasma Physics Laboratory, Princeton, NJ, 08543, USA5) Fusion Research Center, University of Texas, Austin, TX, 78712, USA6) General Atomics, La Jolla, CA, USA7) Plasma Science and Fusion Center, MIT, Cambridge, MA, 02139 USA
Abstract. Gyrokinetic simulations of ITG (k⊥ρi . 1) and ETG (k⊥ρi 1) turbulence arepresented. Comparisons of toroidal turbulence in these two limits provide insights into the dy-namics of streamers and zonal flows. We address the generation of zonal flows by secondaryinstabilities and the regulation of zonal flows by collisionless tertiary instabilities. We presentthe first toroidal electromagnetic gyrokinetic simulations of small scale turbulence, and gyrofluidmodels which explain two important gyrokinetic results: (1) Near marginal stability of the linearITG mode, the turbulence can generate zonal flows that are sufficiently weak to remain stablebut sufficiently strong to suppress the linear ITG mode. This stable region corresponds to theparameter regime of the nonlinear Dimits shift. (2) “Long” wavelength (k⊥ρi 1 > k⊥ρe) ETGturbulence drives experimentally relevant thermal transport, because the secondary modes thatproduce saturation become weak. Finally, preliminary comparisons of simulations with experi-mental data are described.
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355
(TH2/6) Gyrokinetic Theory of Drift Waves in NegativeShear Tokamaks
Y. Idomura1), S. Tokuda1), Y. Kishimoto1), M. Wakatani2)
1) Naka Fusion Research Establishment, Japan Atomic Energy Research Institute, Naka, Ibaraki,311-0193, Japan2) Graduate School of Energy Science, Kyoto University, Uji, Kyoto, 611-0011, Japan
Abstract. Linear and nonlinear properties of slab drift waves in the negative sheared slab con-figuration modeling the qmin-surface region of negative shear tokamaks are studied, where qmin
is the minimum value of a safety factor q. Linear calculations show that both the slab ion tem-perature gradient driven (ITG) mode and the slab electron temperature gradient driven (ETG)mode become strongly unstable around the qmin-surface. Nonlinear simulations are performedfor the ETG turbulence which evolves in a much faster time scale than the ITG turbulence. Itis found that quasi-steady Er ×B zonal flows are generated by an inverse wave energy cascadeprocess. Linear stability analyses of the electrostatic Kelvin-Helmholtz (K-H) mode show thatthe quasi-steady Er×B zonal flow profile is closely related to the q-profile or the magnetic shear,which has a stabilizing effect on the K-H mode. It is shown that the microscopic quasi-steadyEr ×B zonal flows arising from the ETG turbulence have a strong stabilizing effect on the slabITG mode.
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356
(TH2/D) Discussion of Section TH2
The file contains the discussion contributions relating to TH2/1, TH2/2, TH2/3, TH2/4, TH2/6.
F. Porcelli1), A. Airoldi3), C. Angioni4), A. Bruschi3), P. Buratti5), F. Califano6), S. Cirant3),I. Furno4), D. Grasso1), E. Lazzaro3), A. Martynov4), M. Ottaviani7), F. Pegoraro6), G. Ram-poni3), E. Rossi8), O. Sauter4), C. Tebaldi9), O. Tudisco5)
1) INFM and Department of Energetics, Politecnico di Torino, Italy2) Plasma Science and Fusion center, M.I.T., Cambridge, MA, USA3) IFP, Ass. EURATOM-ENEA-CNR, Via Cozzi 53, Milan, Italy4) CRPP, Ass. EURATOM-Conf. Suisse, EPFL, Lausanne, Switzerland5) Ass. EURATOM-ENEA, Frascati, Italy6) INFM and Department of Physics, University of Pisa, Italy7) DRFC, CEA Cadarache, 13108 St Paul lez Durance, France8) IFS, University of Texas at Austin, USA9) Department of mathematics, University of Lecce, Italy
Abstract. We present experimental and theoretical investigations on the dynamics of coherentmagnetic islands in high temperature, magnetically confined plasmas of thermonuclear interest,and of their effects on plasma transport.
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359
(TH3/2) Recent Progress in MHD Stability Calculations ofCompact Stellarators
G. Y. Fu1), L. P. Ku1), M. H. Redi1), C. Kessel1), D. Monticello1), A. Reiman1), W. A. Cooper2),C. Nuhrenberg3), R. Sanchez4), A. Ware5), S. Hirshman6), D. A. Spong6)
1) Princeton Plasma Physics Laboratory, Princeton, NJ 08543, U.S.A.2) Center of Research for Plasma Physics, EPFL, Lausanne, Switzerland3) Max-Planck-Institute for Plasma Physics, Greifswald, Germany4) Universidad Carlos III de Madrid, Madrid, Spain5) University of Montana, Missoula, MT 59812, U.S.A.6) Oak Ridge National Laboratory, Oak Ridge, TN 37831, U.S.A.
Abstract. A key issue for compact stellarators is the stability of beta-limiting MHD modes, suchas external kink modes driven by bootstrap current and pressure gradient. We report here recentprogress in MHD stability studies for low-aspect-ratio Quasi-Axisymmetric Stellarators (QAS)and Quasi-Omnigeneous Stellarators (QOS). We find that the N = 0 periodicity-preservingvertical mode is significantly more stable in stellarators than in tokamaks because of the ex-ternally generated rotational transform. It is shown that both low-n external kink modes andhigh-n ballooning modes can be stabilized at high beta by appropriate 3D shaping without aconducting wall. The stabilization mechanism for external kink modes in QAS appears to be anenhancement of local magnetic shear due to 3D shaping. The stabilization of ballooning modein QOS is related to a shortening of the normal curvature connection length.
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360
(TH3/3) Gyrokinetic Simulation of Internal Collapse in Re-versed Magnetic Shear Tokamak
T. Matsumoto1), H. Naitou2), S. Tokuda1), Y. Kishimoto1)
1) Naka Fusion Research Establishment, Japan Atomic Energy Research Institute, Japan2) Department of Electrical and Electronic Engineering, Yamaguchi Univ., Japan
Abstract. The gyrokinetic particle simulation and the gyro-reduced MHD simulation are ex-ecuted to clarify the kinetic modifications of the MHD phenomena in the reversed shear con-figuration (RSC) of a tokamak plasma. The kinetic (collisionless) double tearing modes in theRSC, which is induced not by the resistivity but by the electron inertia, are found to grow up atthe Alfven time scale by the coupling of two perturbations originated in each resonant surface.It is also found that the internal collapse occurs at the Alfven time scale. After the internalcollapse, the secondary reconnection is induced by the m/n = 2/1 flow. As a result of currentreconcentration, a reversed shear configuration with q < 2 is constructed again.
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361
(TH3/4) The Reversed Field Pinch as a Magnetically Quietand Non Chaotic Configuration
D. F. Escande1)2), S. Cappello2), F. D’Angelo2), C. Marchetto2), R. Paccagnella2), D. Benisti2)
1) UMR 6633 CNRS-Universite de Provence, Marseille, France2) Consorzio RFX, Padua, Italy
Abstract. Recent progress in experiments open a path beyond the standard paradigm that abath of magnetic turbulence is intrinsic to the reversed field pinch (RFP): quasi single helicity(QSH) states have been found in several RFP’s. This motivates a thorough theoretical study ofthe single helicity (SH) states of the RFP which correspond to a laminar dynamo produced bya single mode, and an integrable magnetic field with good flux surfaces, a feature favourable togood confinement. Numerical simulations of visco-resistive MHD reveal a bifurcation from SHto multiple helicity related with temporal intermittency, which is ruled by the product of resis-tivity by viscosity. Furthermore a mechanism of magnetic chaos healing is shown to exist whenthe magnetic separatrix of the dominant mode of QSH states disappears due to a saddle-nodebifurcation.
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362
(TH3/5) Theoretical Understanding of Tokamak PressureLimits
H. R. Wilson1), J. W. Connor1), C. G. Gimblett1), R. J. Hastie2), F. Waelbroeck3)
1) EURATOM/UKAEA Fusion Association, Culham Science Centre, Abingdon, Oxon, OX143DB UK2) MIT Plasma Science and Fusion Center, 167 Albany Street, NW16-234 Cambridge, MA 02139USA3) Institute for Fusion Studies, University of Texas, Austin, Texas 78712 USA
Abstract. A self-consistent theory for the role of the polarisation current in magnetic islandevolution is presented, which suggests that it can, under certain circumstances, provide a thresh-old island width for growth of neoclassical tearing modes. However, at high β coupling to anunstable resistive wall mode (RWM) removes this threshold, and would limit the achievableplasma pressure. Techniques for stabilising the RWM using a rotating shell are described, thusproviding the possibility of high β operation.
A. D. Turnbull1), D. Brennan2)∗, M. S. Chu1), L. L. Lao1), J. R. Ferron1), A. M. Garofalo3)∗,P. B. Snyder1), J. Bialek3), I. N. Bogatu4), J. D. Callen5), M. S. Chance6)∗, K. Comer5),D. H. Edgell4), S. A. Galkin7), D. A. Humphreys1), J. S. Kim4), R. J. La Haye1), T. C. Luce1),M. Okabayashi6), B. W. Rice8), E. J. Strait1), T. S. Taylor1), and H. R. Wilson9)
1) General Atomics, P.O. Box 85608, San Diego, California USA2) Oak Ridge Institute for Science Education, Oak Ridge, Tennessee3) Columbia University, New York, New York4) FARTECH, P.O. Box 221053, San Diego, California USA5) University of Wisconsin-Madison, Madison, Wisconsin USA6) Princeton Plasma Physics Laboratory, Princeton, New Jersey USA7) University of California-San Diego, La Jolla, California USA8) Lawrence Livermore National Laboratory (present address: Xenogen, 860 Atlantic, Alameda,California USA9) Culham Laboratory, UKAEA, Abingdon, Oxfordshire OX14 3DB, UK∗ Present address: General Atomics, P.O. Box 85608, San Diego, California USA
Abstract. Results from an array of theoretical and computational tools developed to treatthe instabilities of most interest for high performance tokamak discharges are described. Thetheory and experimental diagnostic capabilities have now been developed to the point wheredetailed predictions can be productively tested so that competing effects can be isolated andeither eliminated or confirmed. The predictions using high quality discharge equilibrium recon-structions are tested against the observations for the principal limiting phenomena in DIII-D:L-mode negative central shear (NCS) disruptions, H-mode NCS edge instabilities, and tearingand resistive wall modes (RWMs) in long pulse discharges. In the case of predominantly idealMHD instabilities, agreement between the code predictions and experimentally observed sta-bility limits and thresholds can now be obtained to within several percent, and the predictedfluctuations and growth rates to within the estimated experimental errors. Edge instabilities canbe explained by a new model for edge localized modes as predominantly ideal low to interme-diate n modes. Accurate ideal calculations are critical to demonstrating RWM stabilization byplasma rotation and the ideal eigenfunctions provide a good representation of the RWM struc-ture when the rotation slows. Ideal eigenfunctions can then be used to predict stabilization usingactive feedback. For non-ideal modes, the agreement is approaching levels similar to that for theideal comparisons; ∆′ calculations, for example, indicate that some discharges are linearly unsta-ble to classical tearing modes, consistent with the observed growth of islands in those discharges.
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364
(TH3/D) Discussion of Section TH3
The file contains the discussion contributions relating to TH3/1, TH3/3, TH3/4, TH3/5, TH3/6.
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365
Session TH4 — Transport,Barrier, Edge Physics
Contents
(TH4/1) Driftwave-Based Modeling of Poloidal Spin-up Precursor and Step-Wise Expansion of Internal Transport Barriers in Tokamaks . . . . . . 367
(TH4/2) Theoretical Issues in Tokamak Confinement: (i) Internal/EdgeTransport Barriers and (ii) Runaway Avalanche Confinement . . . . . 368
(TH4/3) Plasma Confinement with a Transport Barrier . . . . . . . . . . . . 369
(TH4/4) Turbulence Simulations of X Point Physics on the L-H Transitions 370
(TH4/1) Driftwave-Based Modeling of Poloidal Spin-up Pre-cursor and Step-Wise Expansion of Internal Transport Bar-riers in Tokamaks
G. M. Staebler1), J. E. Kinsey2)∗, P. Zhu3), G. Bateman2), W. Horton3), A. H. Kritz2), T. On-jun2), A. Pankin2), and R. E. Waltz1)
1) General Atomics, San Diego, California USA2) Lehigh University, Bethelehem, Pennsylvania, USA3) University of Texas-Austin, Austin, Texas USA∗ Present address: General Atomics
Abstract. The rich phenomenology of internal transport barriers observed in tokamaks in-cludes a poloidal spin-up precursor for balanced injection neutral beam heating and step-wiseexpansion of the barrier for unbalanced injection. Examples of numerical simulations of thesephenomena are presented. Two driftwave-based predictive transport models (GLF23 and Multi-Mode) are used. Both models include the suppression of ion temperature gradient modes as theEB shear approaches the computed maximum linear growth rate. Modeling of discharges withinternal transport barriers from the DIII-D, JET and TFTR tokamaks are compared.
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367
(TH4/2) Theoretical Issues in Tokamak Confinement: (i)Internal/Edge Transport Barriers and (ii) Runaway AvalancheConfinement
J. W. Connor1), P. Helander1), A. Thyagaraja1), F. Andersson2), T. Fulop2), L.-G. Eriksson3),M. Romanelli4)
1) EURATOM/UKAEA Fusion Association, Culham Science Centre, Abingdon, UK2) Association EURATOM/NFR, Dept of Electromagnetics, Chalmers University of Technology,Goteborg, Sweden3) Association EURATOM-CEA sur la Fusion, CEA Cadarache, France4) ENEA, Centro Ricerche Energia FRASCATI,Via Enrico Fermi, Frascati (RM), Italy
Abstract. This paper summarises a number of distinct, but related, pieces of work on keyconfinement issues for tokamaks, in particular the formation of internal and edge transport bar-riers, both within turbulent and neoclassical models, and radial diffusion of avalanching runawayelectrons. First-principle simulations of tokamak turbulence and transport using the two-fluid,electromagnetic, global code CUTIE are described. The code has demonstrated the spontaneousformation of internal transport barriers near mode rational surfaces, in qualitative agreementwith observations on JET and RTP. The theory of neoclassical transport in an impure, toroidalplasma has been extended to allow for steeper pressure and temperature gradients than are usu-ally considered, and is then found to become nonlinear under conditions typical of the tokamakedge. For instance, the particle flux is found to be a nonmonotonic function of the gradients,thus allowing for a bifurcation in the ion particle flux. Finally, it is shown that radial diffusioncaused by magnetic fluctuations can effectively suppress avalanches of runaway electrons if thefluctuation amplitude exceeds δB/B ∼ 10−3.
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368
(TH4/3) Plasma Confinement with a Transport Barrier
T. Tamano1), I. Katanuma1)
1) Plasma Research Center, University of Tsukuba
Abstract. Plasma confinement with a transport barrier such as an H-mode or an internaltransport barrier mode (ITB-mode) is examined under the constraint due to the conservation oftotal angular momentum. The results are tested against actual experimental data and generalcharacteristics of the plasma confinement with a transport barrier are well understood fromthis constraint. This implies that the confinement of tokamak plasmas can be determined bythe decay rate of the total angular momentum. It also suggests that the confinement with atransport barrier is good since electrostatic fluctuations cannot affect this constraint, but thatelectromagnetic fluctuations such as ELM’s can can cause the confinement to deteriorate.
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369
(TH4/4) Turbulence Simulations of X Point Physics on theL-H Transitions
X. Q. Xu1), R. H. Cohen1), W. M. Nevins1), G. D. Porter1), M. E. Rensink1), T. D. Rognlien1),J. R. Myra2), D. A. D’Ippolito2), R. A. Moyer3), P. B. Snyder4), T. N. Carlstrom4)
1) Lawrence Livermore National Lab., Livermore, CA 94551, USA2) Lodestar Research Corporation, Boulder, CO, 80301, USA3) University of California, San Diego, La Jolla, CA 92093, USA4) General Atomics, San Diego, CA 92186 USA
Abstract. We show that the resistive X-point mode is the dominant mode in boundary plasmasin X-point divertor geometry. The poloidal fluctuation phase velocity from the simulation resultsof the resistive X-point turbulence shows experimentally measured structure across separatrix inmany fusion devices. The fluctuation phase velocity is larger than E×B velocity both in L andH mode phases. We also demonstrate that there is a strong poloidal asymmetry of particle fluxin the proximity of the separatrix. Turbulence suppression in the L-H transition results whensources of energy and particles drive sufficient gradients as in the experiments.
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370
(TH4/5) Nonlinear Zonal Dynamics of Drift and Drift AlfvenTurbulences in Tokamak Plasmas
Abstract. The present work addresses the issue of identifying the major nonlinear physicsprocesses which may regulate drift and drift-Alfven turbulence using a weak turbulence ap-proach. Within this framework, based upon the nonlinear gyrokinetic equation for both elec-trons and ions, we present an analytic theory for nonlinear zonal dynamics described in termsof two axisymmetric potentials, δφz and δA‖z, which spatially depend only on a (magnetic)flux coordinate. Spontaneous excitation of zonal flows by electrostatic drift microinstabilities isdemonstrated both analytically and by direct 3D gyrokinetic simulations. Direct comparisonsindicate good agreement between analytic expressions of the zonal flow growth rate and numeri-cal simulation results for Ion Temperature Gradient (ITG) driven modes. Analogously, we showthat zonal flows may be spontaneously excited by drift-Alfven turbulence, in the form of mod-ulational instability of the radial envelope of the mode as well, whereas, in general, excitationsof zonal currents are possible but they have little feedback on the turbulence itself.
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371
(TH4/6) Edge Transport Modulation by Coherent ShearFlows
Abstract. Different from the core, in the transitional core/edge regime a rich variety of distinctradially and poloidally coherent flow structures have been found by numerical and analytic stud-ies, ranging from solitary propagating zonal flows to sinusoidal structures with definite radialwavenumber. Toward the edge, the fluctuating flows get weaker until the coherent component ofthe flows vanishes completely, which is analytically analogous to the Bose-Einstein phase tran-sition. The coherent flow structures strongly modulate the transport via wave-kinetic effects.The clear signature and the transport relevance of the coherent flows in the transitional regimemake them an ideal target for experimental investigation.
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372
(TH4/D) Discussion of Section TH4
The file contains the discussion contributions relating to TH4/1, TH4/2, TH4/3, TH4/4, TH4/5,TH4/6.
Abstract. The theory of low frequency instabilities in axisymmetric toroidal plasmas is pre-sented from the point of view of the two-fluids equations, assuming the standard drift waveordering. Attention is focused on the limit in which neighboring rational surfaces are sufficientlyfar apart that mode overlapping is non-existent. Owing to field line bending, poloidal side-bandsm ± 1,. . . coexist with the primary mode m, enhancing noticeably the role of the parallel iondynamics. The electron and ion branches are investigated accurately under those conditions.It is found that the radial widths of the eigenmodes increase with respect to the slab values;the shear damping rate of the electron branch, respectively the growth rate of the ion branchincreases correspondingly. Other interesting results are obtained concerning the frequency, thegrowth rate and the poloidal variation of the amplitude of the ion mode fluctuations. Those ex-plain the origin of internal transport barriers; they also suggest ways of interpreting fluctuationsasymmetries observed in tokamaks and (when collisions are included) the Radiative Improvedconfinement mode.
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376
(THP1/02) Self-Similarity and Structures of Plasma Tur-bulence
D. E. Newman1), B. A. Carreras2), V. E. Lynch2), D. Lopez-Bruna3)
1) Department of Physics, University of Alaska, Fairbanks, Alaska, U.S.A.2) Oak Ridge National Laboratory, Oak Ridge, Tennessee 37831-8070, U.S.A3) Asociacion Euratom-CIEMAT, 28040 Madrid, Spain
Abstract. Plasma edge fluctuations and induced fluxes measured in several types of confine-ment devices have been found to be self-similar over time scales between 10 times the turbulencedecorrelation time and the plasma confinement time. These self-similarity parameters vary lit-tle from one device to another. In exploring the self-similarity properties, it has become clearthat time and space measurements lead to different information on the structure of turbulence.Therefore, it is often not possible to clearly separate the poloidal and temporal structures ofthe turbulence with a single-point measure. This in turn implies that using the standard Taylorfrozen flow hypothesis can be very misleading when applied to plasma turbulence. We have usedsimple 2 and 3-D turbulence models to investigate how 1) the multiple nonlinearities intrinsic toplasmas affect the self-similarity parameter for both temporal and poloidal structures and 2) howpoloidal flows influence the single-point measurements. Understanding the temporal and spatialdynamics individually, as well as the relationships between the temporal and spatial dynamics forturbulent plasma systems is crucial to improving the comparison between model and experiment.
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377
(THP1/03) Progress in Gyrokinetic Simulations of ToroidalITG Turbulence
W. M. Nevins1), A. M. Dimits1), B. I. Cohen1), D. E. Shumaker1)
1) Lawrence Livermore National Laboratory, California, USA
Abstract. The 3-D nonlinear toroidal gyrokinetic simulation code PG3EQ is used to studytoroidal ion temperature gradient (ITG) driven turbulence – a key cause of the anomaloustransport that limits tokamak plasma performance. Systematic studies of the dependence ofion thermal transport on various parameters and effects are presented, including dependence on~E × ~B and toroidal velocity shear, sensitivity to the force balance in simulations with radialtemperature gradient variation, and the dependences on magnetic shear and ion temperaturegradient.
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378
(THP1/04) The Role of Dynamo Fluctuations in Anoma-lous Ion Heating, Mode Locking, and Flow Generation
P. W. Terry1), R. Gatto1), R. Fitzpatrick2), C. C. Hegna3), G. Fiksel1)
1) Department of Physics, University of Wisconsin-Madison, Madison, WI 537062) Institute for Fusion Studies, University of Texas at Austin, Austin, TX 787123) Department of Engineering Physics, University of Wisconsin-Madison, Madison, WI 53706
Abstract. Anomalous ion heating intrinsic to magnetic fluctuation-induced electron heat trans-port, the locking of global modes through wall conditions, and flow generation via the magneticReynolds stress all derive from the global, m = 1 tearing modes familiar in the RFP as the dy-namo modes. These important processes are investigated analytically and numerically, yieldingnew insights and predictions for comparison with experiment.
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379
(THP1/05) Effect of Zonal Flows on Drift Wave Turbulence
J. G. Weiland1), S. Mahajan1)
1) Chalmers University of Technology, Gothenburg, Sweden
Abstract. Nonlinearly generated zonal flows are obtained from ion temperature gradient driven(ITG) modes and electromagnetic drift interchange modes. It is found that in general the flow isnot important at the correlation lengths but gives an absorbing boundary for long wavelengths.For ITG modes, however, resonant excitation is found close to marginal stability. This mayexplain the strong effects of zonal flows seen in some nonlinear gyrokinetic simulations.
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380
(THP1/06) Advances in Global Linear Gyrokinetic Simu-lations
L. Villard1), K. Appert1), W. A. Cooper1), G. L. Falchetto1), G. Jost1), M. Maccio1), T. M. Tran2),J. Vaclavik3)
1) CRPP, Association Euratom - Confederation Suisse, Lausanne, Switzerland
2) Service Informatique Central, EPFL, Lausanne, Switzerland
3) La Conversion, Switzerland
Abstract. Substantial advances have been made in the global calculation of microinstabilities.First, we present results of the world’s first global linear gyrokinetic code in fully 3D configura-tions. We show that the unstable Ion Temperature Gradient (ITG) modes in quasi-axisymmetricand quasi-helical configurations exhibit 3D features, while the global stability properties seemrather insensitive to the three-dimensionality of the configuration. Second, the inclusion of equi-librium radial electric fields in both finite element particle-in-cell (PIC) and spectral global codesyield the remarkable result that the value of the E × B flow can be as effective as its shear instabilizing toroidal ITG modes. Third, electron dynamics and finite beta effects are addressedby including electromagnetic perturbations with a two-potential formulation.
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381
(THP1/07) Electrostatic Turbulence with Finite ParallelCorrelation Length and Radial Electric Field Generation
M. Vlad1), F. Spineanu1), J. H. Misguich2), R. Balescu3)
1) Association Euratom-NASTI Romania, National Institute of Laser, Plasmas and RadiationPhysics, P.O.Box MG-36, Magurele, Bucharest, Romania2) Association Euratom-C.E.A. sur la Fusion, CEA-Cadarache, France3) Association Euratom-Etat Belge sur la Fusion, Universite Libre de Bruxelles, Bruxelles, Bel-gium
Abstract. Particle diffusion in a given electrostatic turbulence with a finite correlation lengthalong the confining magnetic field is studied in the test particle approach. An anomalous diffu-sion regime of amplified diffusion coefficients is found in the conditions when particle trappingin the structure of the stochastic potential is effective. The auto-generated radial electric fieldis calculated.
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382
(THP1/08) Mixed SOC Classical Diffusive Dynamics as aParadigm for Transport in Fusion Devices
R. Sanchez1), D. E. Newman2), B. A. Carreras3)
1) Universidad Carlos III de Madrid, 28911 Leganes, Madrid, SPAIN2) University of Alaska-Fairbanks, Fairbanks, AK, U.S.A.3) Oak Ridge National Laboratory, Oak Ridge, TN, U.S.A.
Abstract. A recently proposed paradigm for plasma turbulent transport dynamics, based onthe concept of self-organized criticality (SOC), is extended to include other transport mechanismsexistent in real plasmas. This extension might clarify the experimentally observed violation atfluctuation scales of the scale-invariance essential to the SOC model. It might also provide withnew experimental tests that could help to validate its relevance. Finally, it might give somehints to understand the central role played by the plasma edge in transport dynamics.
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383
(THP1/09) Numerical Assessment of the Ion TurbulentThermal Transport Scaling Laws
M. Ottaviani1), G. Manfredi2)
1) Departement de Recherches sur la Fusion Controlee, Commissariat a l’Energie Atomique,Centre de Cadarache, 13108 St. Paul lez Durance, France2) Laboratoire de Physique des Milieux Ionises, Universite Henri Poincare, Nancy-1, BP 239,F-54506 Vandoeuvre-les-Nancy cedex, France
Abstract. Numerical simulations of ion temperature gradient (ITG) driven turbulence werecarried out to investigate the parametric dependence of the ion thermal transport on the reducedgyroradius and on the local safety factor. Whereas the simulations show a clear proportionalityof the conductivity to the gyroradius, the dependence on the safety factor cannot be representedas a simple power law like the one exhibited by the empirical scaling laws.
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384
(THP1/10) Flow Shear Stabilization of Ion TemperatureGradient Modes in an Internal Transport Barrier
J. Q. Dong1)
1) Southwestern Institute of Physics, China
Abstract. Ion temperature gradient (ITG) driven instability is studied with gyrokinetic theoryin an internal transport barrier (ITB) of tokamak plasmas. The stabilization effects of a parallelvelocity shear on the modes are investigated. It is found that the mode structures and stabilityproperties, as well as the effects of a velocity shear, in an ITB are significantly different fromthat in off-ITB regions.
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385
(THP1/12) The Influence of Zonal ExB Flows on Edge Tur-bulence in Tokamaks and Stellarators
B. Scott1), F. Jenko1),A. Kendl1)
1) Max-Planck-IPP, Garching, Germany
Abstract. We report on fluid, gyrofluid and gyrokinetic numerical studies of edge turbulence inboth tokamak and stellarator geometry, regarding both its physical character and its interactionwith the flux surface averaged (“zonal”) ExB flows and magnetic fields. We run several elec-tromagnetic models under drift ordering in globally consistent flux tube geometry: three fluidmodels of increasing complexity, a gyrofluid model, and a collisionless gyrokinetic phase spacecontinuum model. All treat both electron and ion dynamics. The fluid models employ a Landauclosure for the parallel heat fluxes, which are treated dynamically. The fluid and gyrofluid modelsall run at arbitrary collisionality. We find turbulence of drift wave and ion temperature gradientmode (ITG) character at all parameters of interest to experiment. Although an ExB shear layerimposed by the equilibrium is effective in suppressing turbulence, a strong enough shear layer toaccount for the L-to-H confinement transition is never self-generated by the turbulence in threedimensions. The turbulence correlation length is more sensitive to the equilibrium ExB shearthan to any other parameter. Zonal fields in contrast to zonal flows have little effect below theideal MHD boundary. Similar results are found for both stellarator and tokamak geometry, inagreement with experimentally observed universality of turbulence.
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386
(THP1/13) Radial Current in Radial Electric Field
X. D. Zhang1)
1) Institute of plasma physics, academia sinica
Abstract. The collision effect of cross-field transport in a radial electric field for ions andelectrons is studied and the transport equations are obtained, leading to a better understandingof the relation of collision effect with a radial electric field. The collisional transport equationsindicate that there is a radial current generated by the damping of ExB drift. This currentis in the direction of the radial electric field, so it will reduce the radial electric field and isunfavorable for forming the H mode.
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387
(THP1/14) New Coulomb Logarithm and Its Effects toFokker-Planck Equation, Relaxation Times and Cross-FieldTransports in Fusion Plasmas
Ding Li1)
1) Department of Modern Physics, University of Science and Technology of China, Hefei 230027,China
Abstract. It is shown that a new cutoff at small scattering angle should be introduced for con-stant relative velocity meanwhile a new cutoff on the velocity increment should be introduced forvaried relative velocity based on the effective collision conditions. It is found that the Coulomblogarithm should be reduced to half of the well-known result. Consequently, the Fokker-Planckcoefficients are modified and applicable to both weakly and moderately coupled plasmas. Therelaxation times increase and the cross-field electrical resistivity, electron diffusion and ion ther-mal conductivity are reduced to half level for Maxwellian scatters. The non-Maxwellian effectcan furthermore modify the Fokker-Planck coefficients, increase the relaxation times and reducethe cross-field transport coefficients.
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388
(THP1/16) X Transport and its Effect on H-mode and EdgePedestal in Tokamaks
C. S. Chang1), D. Darrow3), S. H. Ku5), H. Weitzner2), R. White3), Z. Lin3), W. Lee3),T. N. Carlstrom4), J. S. deGrassie4)
1) New York University, New York, U.S.A. and KAIST, Korea2) New York University, New York, U.S.A.3) PPPL, Princeton, U.S.A.4) General Atomics, San Diego, U.S.A.5) Department of Physics, KAIST, Korea
Abstract. A new classical non-ambipolar transport mechanism has been identified which canbe a dominant source of strong Er and edge pedestal layer formation immediately inside theseparatrix in a diverted tokamak. Due to vanishingly small poloidal B-field and grad-B drifttoward x-point, plasma ions with small v‖ in the X-region do not have confined single particleorbits. This leads to a non-ambipolar convective transport in the X-region (X-transport), eithercollisional or collisionless, inducing a strong negative Er-shear layer. The X-transport can pro-vide basic understanding of many of the experimental observations.
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389
(THP1/17) Accretion Theory for “Spontaneous” Rotationof Toroidal Plasmas
B. Coppi1)
1) Massachusetts Institute of Technology, Cambridge MA, USA
Abstract. The accretion theory of spontaneous toroidal rotation connects this phenomenon tothe energy and particle transport properties of the plasma column and to the relevant collectivemodes. The consequent prediction that an inversion of the velocity direction in the transitionfrom the H to the L regime should occur has been verified by the experiments. The theory isconsistent also with the observation that the velocity is depressed when a peaked density profileappears as a result of a transport barrier. The fact that the rotation velocity increases with thetotal energy content is explained by the fact that the inflow of angular momentum, whose sourceis at the edge of the plasma column, results from the excitation of modes driven by the plasmapressure gradient. A quasi-linear derivation of the angular momentum transport produced bythese modes, whose novel feature is the inflow, is given and a model of the relevant (transport)equation is solved. Fluctuations at the edge of the plasma column are considered responsible forthe scattering, out of confinement, of particles that transfer, to the surrounding material wall,angular momentum in the same direction as that of the phase velocity of the prevalent modes.Thus the fact that the plasma rotates in the direction of the ion diamagnetic velocity in theH-regime, when the prevalent modes are expected to have phase velocity in the direction of theelectron diamagnetic velocity, can be explained and the rate of rotation decrease, as the plasmacurrent is increased, can be justified.
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390
(THP1/18) Ignitor Physics Assessment and ConfinementProjections
W. Horton1), F. Porcelli2), P. Zhu1), A. Aydemir1), Y. Kishimoto3), T. Tajima1)
1) Institute for Fusion Studies, The University of Texas, Austin, TX 78712, U.S.A.2) INFM and Dipartimento di Energetica, Politecnico di Torino, 10129 Torino, Italy3) Plasma Science and Fusion Center, M.I.T., Cambridge, MA, U.S.A.
Abstract. An independent assessment is presented of the physics of Ignitor, a physics demon-stration experiment for achieving thermonuclear ignition (where fusion alpha heating compen-sates for all forms of energy losses). Simulations show that a pulse of particle power up to 10-20MW is produced for a few seconds. Crucial issues are the production of peaked density profilesover several energy confinement times, the control of current penetration for the optimizationof ohmic heating, and sawtooth avoidance. The presence of a 10-20 MW ion cyclotron radiofrequency system and the operation of a high-speed pellet injector are considered essential toprovide added flexibility in order to counter unexpected, adverse plasma behavior.
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391
(THP1/19) Theory of Dynamics in Long Pulse Helical Plas-mas
K. Itoh1), H. Sanuki1), S. Toda1), S.-I. Itoh2), M. Yagi2), A. Fukuyama3),M. Yokoyama1)
1) National Institute for Fusion Science, Toki 509-5292, Japan2) Research Institute for Applied Mechanics, Kyushu University, Kasuga 816-8580, Japan3) Department of Nuclear Engineering, Kyoto University, Sakyo-ku, Kyoto, 606-8501, Japan
Abstract. Self-organized dynamics of toroidal helical plasma, which is induced by the nonlin-ear transport property, is discussed. Neoclassical ripple diffusion is a dominant mechanism thatdrives the radial electric field. The bifurcation nature of the electric field generation gives rise tothe electric field domain interface, across which the electric field changes strongly. This domaininterface is an origin of internal transport barrier in helical systems. This nonlinearity gives riseto the self-organized oscillations; the electric field pulsation is one of the examples. Based onthe model of density limit, in which the competition between the transport loss and radiationloss is analyzed, dynamics near the density limit of helical systems is also discussed.
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392
(THP1/20) Triggering Mechanisms for Transport Barriers
J. A. Heikkinen1), O. Dumbrajs2), S. Karttunen1), T. Kiviniemi2), T. Kurki-Suonio2), M. Mantsi-nen2), K. Rantamaki1), S. Saarelma2), R. Salomaa2), S. Sipila2), T. Tala1)
1) VTT Chemical Technology, Espoo, Finland2) Helsinki University of Technology, Espoo, Finland
Abstract. The radial shear ωE×B of the E × B flow is evaluated with the Monte Carlo orbitfollowing code ASCOT at the onset of the L-H transition and internal transport barriers (ITB)in JET, TFTR, ASDEX Upgrade, TEXTOR, and FT-2 tokamaks. Systematically, a large shear(sufficient for turbulence suppression) is found for local parameters close to the experimentalthreshold conditions at the barrier location. For L-H transition in JET and ASDEX Upgrade,the large shear is obtained by increasing the edge ion temperature. For TEXTOR, the radialelectric field and the electrode current bifurcate at a threshold electrode voltage. In a JETdatabase study, toroidal rotation is found to be dominant in triggering the JET ITB, and anempirical s-ωE×B fit is found for the transition threshold. For TFTR and FT-2, in which toroidalrotation does not play a role, ASCOT predicts a significant ωE×B shear for the ITB conditions.The ripple-induced transport is not found to be important here.
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393
(THP1/21) Stability of Neoclassical Rotation in Edge Plas-mas
U. Daybelge1), C. Yarim1), H. A. Claassen2), A. L. Rogister2)
1) Istanbul Technical University, Faculty of Aeronautics and Astronautics, 80626 Maslak, Is-tanbul/TURKEY2) IPP-T, Forschungszentrum Julich, GmbH, Association EUROATOM-FZJ, Trilateral EuregioCluster, D-52425, Julich/GERMANY
Abstract. Using a revised neoclassical theory, the stability of rotation velocities of a collisionaltokamak plasma near the separatrix is investigated. First, assuming equilibrium, full coupledequations of poloidal and toroidal velocities are solved by numerical means and depending onthe underlying temperature and density profiles behavior of solutions, such as shocks etc. isstudied. Secondly, using a generic quasilinear differential equation for the toroidal velocity, weinvestigate its evolution from an initial profile depending on the coefficients of the equation andgive a criterion for the development of a breaking, or a shock front represented by it.
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394
(THP1/22) Simulation Study of Detached Plasmas by Us-ing Advanced Particle Model and Fluid Model
T. Takizuka1), K. Shimizu1), N. Hayashi1), M. Hosokawa2)
1) Japan Atomic Energy Research Institute, Naka Fusion Research Establishment, Naka, Japan2) Research Organization for Information Science & Technology, Tokai, Japan
Abstract. Fluid simulations and particle simulations are performed to understand the physicsof detached plasmas in the tokamak divertor. Two dimensional fluid simulations show thatdetached divertor plasmas are formed for the high density operation in the W-shaped divertorconfiguration of JT-60U tokamak. Charge-exchange and recombination processes play importantroles to cause the detachment. The asymmetry of inner-and-outer divertor plasmas is studiedbased on a fluid model, and the bifurcated nature of the asymmetry caused by the SOL currentis found. Advanced particle simulations demonstrate that the E× B drift by the radial electricfield in a SOL plasma causes the asymmetry of flow pattern and density profile. A detachedplasma is formed in the divertor region from which the drift flows out, when the ratio of theE× B drift speed to the sound speed exceeds a threshold. Effects of the radial gradient includingdiamagnetic drift flow on SOL and divertor plasmas are also studied with the two-dimensionalparticle simulation.
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395
(THP1/23) Transport Simulations for Tokamak Edge Plas-mas
T. D. Rognlien1), R. H. Cohen1), L. L. LoDestro1), G. D. Porter1), M. E. Rensink1), D. D. Ryu-tov1), X. Q. Xu1)
1) University of California Lawrence Livermore National Laboratory
Abstract. The edge plasma plays key roles in tokamak devices: it generates the edge transport-barrier yielding the L-H core confinement transition, distributes the core charged-particle energyto surrounding material surfaces, shields the core from impurities, and removes helium ash infusion plasmas. The transport of density, momentum, and energy in the near-separatrix edgeregion, and the corresponding self-consistent electrostatic potential, require a two-dimensionaldescription, here incorporated into the UEDGE code. In the direction across the B-field, bothturbulent transport and classical cross-field flows are important. The role of classical flows isanalyzed in detail in the presence of an assumed diffusive turbulent transport. Results andexplanations are given for the generation of radial electric field near the separatrix, edge plasmaasymmetries and differences between double-null DIII-D and NSTX devices, comparisons withDIII-D diagnostics, and core/edge transport coupling.
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396
(THP1/24) Edge Modelling for W7-X
R. Schneider1), M. Borchardt1), J. Riemann1), X. Bonnin1), J. Nuhrenberg1), A. Mutzke1)
Abstract. The edge modelling activities for W7-X are summarized. The status of the new3D SOL transport code BoRiS is presented, including an algorithm for calculation of magneticcoordinates and metric coefficients. In addition, the analysis of a toroidally averaged islandtopology with respect to the effect of drift and currents is discussed using the 2D B2-solps5.0code.
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397
(THP1/25) On Radiative Density Limits in Stellarators
H. Wobig1)
1) Max-Planck-Institut fur Plasmaphysik, Garching bei Munchen, Germany
Abstract. Density limits in stellarators are caused mainly by enhanced impurity radiationleading to a collapse of the temperature. A simple model can be established, which computesthe temperature in the plasma with a fixed heating profile and a temperature-dependent ra-diation profile. If the temperature-dependent radiation function has one or several extrema,multiple solutions of the transport equation exist and radiative collapse occurs when the hightemperature branch merges with the unstable temperature branch. At this bifurcation point thetemperature decreases to a stable low temperature solution. The bifurcation point is a functionof the heating power and the plasma density. Thus a density limit can be defined as the pointwhere bifurcation occurs. It is shown that bifurcation and sudden temperature collapse does notoccur below a power threshold. Anomalous thermal conductivity and the details of the impurityradiation, which in the present model is assumed to be in corona equilibrium, determine thescaling of the density limit. A model of the anomalous transport is developed, which leads toGyro-Bohm scaling of the confinement time. The density limit based on this transport model isclose to experimental findings in Wendelstein 7-AS.
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398
(THP1/26) Effects of Tokamak Flux Surface Non-circularityon Charged Particle Transport Processes in the Weak Col-lisionality Regime
K. Schoepf1), B. H. Cho1), V. Ya. Goloborod’ko2), S. N. Reznik2), V. A. Yavorskij2)
1) Institut fur Theoretische Physik, Universitat Innsbruck, Assoz. EURATOM-OAW, Austria2) Institute for Nuclear Research, Ukrainian Academy of Sciences, Kyiv, Ukraine
Abstract. Introducing a novel analytical model for tokamak fields with prescribed flux sur-face shapes into a 3D Fokker-Planck simulation, the influence of parameters determining thenon-circularity of flux surfaces on the longitudinal transport of charged particles as well as ontheir radial fluxes is investigated for weak collisionality. It is demonstrated that a high trian-gularity can decrease the trapped particle fraction by up to 30%. An increase of elongationand triangularity at fixed plasma current and fixed B was found to reduce the effective radialexcursion of charged particles thus resulting in the improvement of plasma confinement. Anextreme sensitivity of the NBI induced current to the flux surface non-circularity was seen forZbeam ≈ Zeff . Another interesting result is that the triangularity and up-down asymmetry offlux surface strongly affect the slowing-down induced radial convection of high-energy toroidallytrapped charged particles being in resonance with the TF ripple perturbation. This transportmechanism is deemed to be consequential for the confinement of charged fusion products intokamaks. Further the ripple-induced stochastic transport domains are shown to vary with fluxsurface shapes.
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399
(THP1/27) Nonlinear Simulations of Turbulence Suppres-sion with External Flows and Impurity Injection in ToroidalPlasmas
R. D. Sydora1), J.-N. Leboeuf2), J. M. Dawson2), V. K. Decyk2), M. W. Kissick2), C. L. Ret-tig2), T. L. Rhodes2), G. Tynan3), J. A. Boedo3), J. Ongena4), A. Messiaen4), P. E. Vandenplas4)
1) Department of Physics, University of Alberta, Edmonton, Canada2) Department of Physics and Astronomy, UCLA, Los Angeles, USA3) Department of Mechanical and Aerospace Engineering, Univ. of California, San Diego, USA4) Laboratoire de Physique des Plasmas - Laboratorium voor Plasmafysica, ERM/KMS, Brus-sels, Belgium and IPP-Forschungszentrum, Julich, Germany
Abstract. Ion temperature gradient-driven (ITG) turbulence plays an important role in ex-plaining measured ion thermal transport in tokamaks, particularly under L-mode conditions.Nonlinear global toroidal gyrokinetic simulation results are presented with radiative impurityseeding in TEXTOR-like L-mode plasma. Reduced levels of ITG turbulence and ion heat trans-port are observed, possibly explaining the origin of the improved confinement regime radiativeimproved(RI)-mode. In a separate investigation, ITG turbulence in several DIII-D-like dis-charges has been analyzed using our nonlinear model under different experimental conditions,with high and low density and central ion temperature, and comparisons with experiment havebeen favorable. Turbulence radial correlation lengths have been compared and are found to besimilar when the effect of zonal flows is included.
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400
(THP1/28) Poloidal Plasma Rotation in the Presence of RFWaves in Tokamaks
B. Weyssow1), Caigen Liu1)
1) Physique Theorique and Mathematique, Unite de Physique des Plasmas, Association Euratom-Etat Belge, Universite Libre de Bruxelles, Campus Plaine, CP 231, Bvd du Triomphe, 1050Bruxelles, Belgium
Abstract. It is well known that one of the consequences of strong RF heating is the deformationof the equilibrium distribution function that induces a change in plasma transport and plasmarotation. The poloidal plasma rotation during RF wave heating in tokamaks is investigated usinga moment approach. A set of closed, self-consistent transport and rotation equations is derivedand reduced to a single equation for the poloidal particle flux. The formulas are sufficientlygeneral to apply to heating schemes that can be represented by a quasilinear operator.
(THP2/30) 3D Fokker Planck Model for High and Moderate Energy Ionsin Weakly Ripple Tokamaks . . . . . . . . . . . . . . . . . . . . . . . . . . 429
403
(THP2/01) Theoretical Modeling of the Feedback Stabiliza-tion of External MHD Modes of Toroidal Geometry
M. S. Chance1), M. S. Chu2), M. Okabayashi1)
1) Princeton Plasma Physics Laboratory, Princeton, NJ, USA2) General Atomics, San Diego, CA, USA
Abstract. A theoretical framework for understanding the feedback mechanism against exter-nal MHD modes has been formulated. Efficient computational tools – the gato stability codecoupled with a substantially modified vacuum code – have been developed to effectively designviable feedback systems against these modes. The analysis assumed a thin resistive shell and afeedback coil structure accurately modeled in θ, with only a single harmonic variation in φ. Anoptimized configuration and placement of the feedback and sensor coils as well as the time con-stants and induced currents in the enclosing resistive shell have been computed for the DIII-Ddevice. Up to 90% of the effectiveness of an ideal wall can be achieved.
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404
(THP2/02) Global Stability of the Field Reversed Config-uration
E. V. Belova, S. C. Jardin, H. Ji, R. M. Kulsrud, W. Park, M. Yamada
Princeton Plasma Physics Laboratory, Princeton, USA
Abstract. New computational results are presented which provide a theoretical basis for thestability of the Field Reversed Configuration (FRC). The FRC is a compact toroid with neg-ligible toroidal field in which the plasma is confined by a poloidal magnetic field associatedwith toroidal diamagnetic current. Although many MHD modes are predicted to be unstable,FRCs have been produced successfully by several formation techniques and show surprisingmacroscopic resilience. In order to understand this discrepancy, we have developed a new 3Dnon-linear hybrid code (kinetic ions and fluid electrons), M3D-B, which is used to study the roleof kinetic effects on the n = 1 tilt and higher n modes in the FRC. Our simulations show thatthere is a reduction in the tilt mode growth rate in the kinetic regime, but no absolute stabiliza-tion has been found for s . 1, where s is the approximate number of ion gyroradii between thefield null and the separatrix. However, at low values of s, the instabilities saturate nonlinearlythrough a combination of a lengthening of the initial equilibrium and a modification of the iondistribution function. These saturated states persist for many Alfven times, maintaining fieldreversal.
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405
(THP2/03) Is the Mercier Criterion Relevant to StellaratorStability?
B. A. Carreras1), V. E. Lynch1), K. Ichiguchi2), M. Wakatani3), T. Tatsuno4)
1) Oak Ridge National Laboratory, Oak Ridge, Tennessee, U.S.A.2) National Institute for Fusion Science, Toki, Japan3) Graduate School of Energy Science, Kyoto University, Kyoto, Japan4) Graduate School of Frontier Science, Tokyo University, Tokyo, Japan
Abstract. Local flattening of the pressure profile at the resonant surfaces may significantlychange the stellarator stability properties. This flattening may be an intrinsic consequence ofthe three-dimensional nature of the equilibrium and may invalidate the local stability criteriaoften used in stellarator design.
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406
(THP2/04) Studies of Spherical Tori, Stellarators and AnisotropicPressure with M3D
L. E. Sugiyama1), W. Park2), H. R. Strauss3), S. Hudson2), D. Stutman4), X.-Z. Tang2)
1) Massachusetts Institute of Technology, Cambridge MA, U.S.A.2) Princeton Plasma Physics Laboratory, Princeton NJ, U.S.A.3) New York University, New York NY, U.S.A.4) Johns Hopkins University, Baltimore MD, U.S.A.
Abstract. The M3D (Multi-level 3D) project simulates plasmas using multiple levels of physics,geometry, and grid models in one code package. The M3D code has been extended to fundamen-tally nonaxisymmetric and small aspect ratio, R/a & 1, configurations. Applications include thenonlinear stability of the NSTX spherical torus and the spherical pinch, and the relaxation ofstellarator equilibria. The fluid-level physics model has been extended to evolve the anisotropicpressures pj‖ and pj⊥ for the ion and electron species. Results show that when the densityevolves, other terms in addition to the neoclassical collisional parallel viscous force, such asB · ∇pe in the Ohm’s law, can be strongly destabilizing for nonlinear magnetic islands.
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407
(THP2/05) Dipole Equilibrium and Stability
M. E. Mauel1), D. T. Garnier1), T. Sunn Pedersen1), P. J. Catto2)3), R. J. Hastie2), J. Kesner2),J. J. Ramos2), A. N. Simakov2), S. I. Krasheninnikov4)
1) Columbia University, New York, NY, U.S.A.2) MIT, Cambridge, MA, U.S.A.3) Lodestar Research Corporation, Cambridge, MA, U.S.A.4) University of California, San Diego, CA, U.S.A.
Abstract. A plasma confined in a dipole field exhibits unique equilibrium and stability prop-erties. In particular, equilibria exist at all β values and these equilibria are found to be stableto ballooning modes when they are interchange stable. When a kinetic treatment is performedat low beta we also find a drift temperature gradient mode which couples to the MHD mode inthe vicinity of marginal interchange stability.
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408
(THP2/06) Mercier Instabilities in the Alcator C-Mod Toka-mak
R. J. Hastie1), Y. In1), J. J. Ramos1), P. J. Catto1)2), A. Hubbard1), I. H. Hutchinson1),E. S. Marmar1), M. Miklos1), J. A. Snipes1), S. Wolfe1), G. Taylor3), A. Bondeson4)
1) Plasma Science and Fusion Center, MIT, Cambridge, MA 02139, U.S.A.2) Lodestar Research Corporation, Cambridge, MA 02139, U.S.A.3) Princeton Plasma Physics Laboratory, Princeton, New Jersey 08543, U.S.A.4) Chalmers University of Technology, Goteborg, Sweden
Abstract. During current ramp discharges, highly localized MHD fluctuations were observedon ECE diagnostics of Alcator C-Mod tokamak. The electron temperature profile was hollow,while the density profile was almost flat. Assuming that the equilibration time was short enoughto quickly thermalize ions with Ti/Te ≈ 0.9, the pressure profile was also found to be hollow.Using the pressure profile as an additional constraint to the EFIT program, an equilibrium withreversed shear was constructed, whose q(0) 1. The localized MHD activity was observednear the inner q=5 rational surface in this reconstructed equilibrium. According to ideal MHDstability theory, it was found to be ideally unstable because of the reversed pressure gradient(dp/dr > 0), q 1 and moderate shear. When kinetic effects are added, the ideal Mercier modewas finite ion Larmor radius (FLR) stabilized. However, considering that the ions are collision-less (νii ω), and the thermal ion transit frequency is comparable to the ion diamagnetic driftfrequency (ωti ∼ ω∗pi), ion Landau damping was found to be strong enough to drive kineticMercier instability. As a result, the localized fluctuations in C-Mod can be attributed to a FLRmodified kinetic Mercier instability.
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409
(THP2/07) Nonlinear Evolution of Neoclassical Tearing Modesin the Presence of Sheared Flows
A. Sen1), P. K. Kaw1), D. Chandra1), M. P. Bora1)
1) Institute for Plasma Research, Bhat, Gandhinagar 382428, India
Abstract. The center manifold reduction technique is employed to study the nonlinear evo-lution of an (m = 2, n = 1) neoclassical tearing mode and its first harmonic, in the presence ofequilibrium sheared flows. A detailed bifurcation diagram of the reduced amplitude equations ispresented delineating the parametric regimes for the occurrence of single saturated island statesand oscillating island solutions.
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410
(THP2/08) Improved Stability Due to Local Pressure Flat-tening in Stellarators
M. Wakatani1), K. Ichiguchi2), T. Unemura1), T. Tatsuno3), B. A. Carreras4)
1) Graduate School of Energy Science, Kyoto University, Gokasho, Uji, Japan 611-00112) National Institute for Fusion Science, Oroshi-cho 322-6, Toki, Japan 509-52923) Graduate School of Frontier Science, University of Tokyo, Tokyo, Japan 113-86564) Oak Ridge National Laboratory, Oak Ridge, Tennessee 37831, USA
Abstract. It is demonstrated that the stability of low n pressure gradient driven modes isimproved by introducing local pressure flattening at low order rational surfaces in LHD (LargeHelical Device) with the inward magnetic axis shift of 25cm, where n is the toroidal mode number.
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411
(THP2/09) Nonlinear Tearing Mode in Turbulent Plasmas
M. Yagi1), S.-I. Itoh1), K. Itoh2), A. Fukuyama3)
1) Research Institute for Applied Mechanics, Kyushu University, Kasuga, Japan2) National Institute for Fusion Science, Toki, Japan3) Department of Nuclear Engineering, Kyoto University, Kyoto, Japan
Abstract. A kind of large eddy simulation model is proposed to incorporate the effect of mi-croturbulence into MHD. This model is applied to the analysis of magnetic island dynamicsin the presence of strong pressure gradient. It is found that the growth of magnetic island isaccelerated by the anomalous resistivity and the propagation frequency of magnetic island isdictated by turbulent viscosity.
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412
(THP2/10) An Innovative Method for Ideal and ResistiveMHD Stability Analysis of Tokamaks
S. Tokuda1)
1) Japan Atomic Energy Reserach Institute, Ibaraki, Japan
Abstract. An advanced asymptotic matching method of ideal and resistive MHD stabilityanalysis in tokamak is reported. The report explains a solution method of two-dimensionalNewcomb equation, dispersion relation for an unstable ideal MHD mode in tokamak, and a newscheme for solving resistive MHD inner layer equations as an initial-value problem.
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413
(THP2/11) Improved Theory of Forced Magnetic Recon-nection Due to Error Field and its Application to Seed Is-land Formation for NTM
A. Ishizawa1), S. Tokuda2), M. Wakatani1)
1) Kyoto University, Kyoto, Japan2) Japan Atomic Energy Research Institute, Ibaraki, Japan
Abstract. A seed island is required for destabilizing the neo-classical tearing mode (NTM),which degrades confinement in long sustained, high-confinement, high beta plasmas. The seedisland formation due to an MHD event, such as a sawtooth crash, is investigated by applyingthe improved boundary layer theory of forced magnetic reconnection. This improved theoryintroduces the non-constant-ψ matching and reveals the complicated feature of the reconnectiondescribed by two reconnected fluxes. In the initial evolution, these reconnected fluxes grow onthe time scale including the ideal time scale, typical time scale of the MHD event and the timescale of resistive kink mode. The surface current is negative, ∆′(t) < 0, to be consistent with theNTM theory. The theory also yields an integral equation which includes the typical time scaleof the resistive kink mode, and allows us to investigate the time evolution of the seed island att ≈ τAS1/3.
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414
(THP2/12) Theory and MHD Simulation of Fuelling Pro-cess by Compact Toroid (CT) Injection
Y. Suzuki1), T. Hayashi2), Y. Kishimoto1)
1) Japan Atomic Energy Research Institute, Ibaraki, Japan2) National Institute for Fusion Science, Gifu, Japan
Abstract. The fuelling process by a spheromak-like compact toroid (SCT) injection is investi-gated by using MHD numerical simulations, where the SCT is injected into a magnetized targetplasma region corresponding to a fusion device. In our previous study, the theoretical modelto determine the penetration depth of the SCT into the target region has been proposed basedon the simulation results, in which the SCT is decelerated not only by the magnetic pressureforce but also by the magnetic tension force. However, since both ends of the target magneticfield are fixed on the boundary wall in the simulation, the deceleration caused by the magnetictension force would be overestimated. In this study, the dependence of the boundary conditionof the target magnetic field on the SCT penetration process is examined. From these results,the theoretical model we have proposed is improved to include the effect that the wave lengthof the target magnetic field bent by the SCT penetration expands with the Alfven velocity. Inaddition, by carrying out the simulation with the torus domain, it is confirmed that the theo-retical model is applicable to estimate the penetration depth of the SCT under such conditions.Furthermore, the dependence of the injection position (the side injection and the top/bottominjection) on the penetration process is examined.
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415
(THP2/13) Dynamics of Relaxation Phenomena in Spheri-cal Tokamak
T. Hayashi, N. Mizuguchi, H. Miura, T. Sato
1) National Institute for Fusion Science, Toki, Japan
Abstract. Three-dimensional magnetohydrodynamic (MHD) simulations are executed to clar-ify the physical mechanisms of MHD relaxation activities, specifically the Internal ReconnectionEvent (IRE), which are observed in the spherical tokamak experiments. For a case of an initialequilibrium with q(0) less than one, tunnel-like plasma jet flow is formed by the occurrence ofthe external magnetic reconnection accompanying the growth of localized deformation of torus.When initial q(0) is slightly greater than one, upon increase of beta, multiple mode numberof finer-scale ballooning type instabilities is triggered, where the typical toroidal mode numbern is 11. In the nonlinear stage, the surface of the torus is deformed due to growth of thosemedium-n modes, and a part of heat energy is lost into the external region through occurrenceof external reconnection. Interestingly, low n modes, particularly n=2, are enhanced presumablyby nonlinear coupling among the unstable medium-n modes, and the n=2 type deformation ofthe torus becomes dominant in the later stage.
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416
(THP2/14) Three Dimensional Ideal MHD Stability Anal-ysis in L=2 Heliotron Systems
N. Nakajima1)
1) National Institute for Fusion Science, Oroshi-cho 322-6, Toki, Japan
Abstract. Global mode analyses of ideal MHD instability are performed under both fixed andfree boundary conditions for L/M = 2/10 heliotron configurations, where L and M are thepolarity and toroidal field period of helical coils, respectively. Under the fixed boundary condi-tion, dangerous low-n ballooning modes with n < M (n is a typical toroidal mode number) arenot destabilized, when the pressure gradient is quite weak near the boundary. However, suchballooning modes and/or peeling modes are destabilized under the free boundary condition evenfor the currentless condition, when the pressure gradient becomes strong there. Effects of thenet toroidal current on the external modes are also discussed.
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417
(THP2/16) Approximate Variational Solutions of the Grad-Shafranov Equation
G. O. Ludwig1)
1) Associated Plasma Laboratory, National Space Research Institute, Sao Jose dos Campos,SP, Brazil
Abstract. Approximate solutions of the Grad-Schluter-Shafranov equation based on variationalmethods are developed. The power series solutions of the Euler-Lagrange equations for equilib-rium are compared with direct variational results for a low aspect ratio tokamak equilibrium.
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418
(THP2/17) Stochastic Loss of Alpha Particles in a HeliasReactor
Ya. I. Kolesnichenko1), C. D. Beidler2), V. S. Marchenko1), I. Sidorenko2), H. Wobig2)
1) Scientific Centre “Institute for Nuclear Research”, Kyiv, 03680, Ukraine2) Max-Planck-Institut fur Plasmaphysik, EUROATOM Association D-85740 Garching bei Munchen,Germany
Abstract. It is shown that collisionless orbit transformation of the locally trapped particlesto the locally passing ones and vice versa in the Wendelstein-line optimized stellarators resultsin stochastic diffusion of energetic ions. This diffusion can lead to the loss of an essential frac-tion of energetic ion population from the region where the characteristic diffusion time is smallcompared to the slowing down time. The loss region and the magnitude of the loss can beminimized by shaping the plasma temperature and density profiles so that they satisfy certainrequirements. The predictions of the developed theory are in agreement with the results ofnumerical modelling of confinement of α-particles in a Helias reactor, which was carried out inthis work with the use of the orbit following code. The considered diffusion seems to representthe dominant mechanism of classical losses of α-particles in a Helias reactor.
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419
(THP2/18) Transport of Energetic Ions in MHD-active High-Beta Plasmas of Spherical Tokamaks
V. V. Lutsenko1), Ya. I. Kolesnichenko1), R. White2), Yu. V. Yakovenko1)
1) Scientific Center “Institute for Nuclear Research”, Kyiv, 03680, Ukraine2) Princeton Plasma Physics Laboratory, P.O. Box 451, Princeton, NJ, 08543, USA
Abstract. It is shown that high β (β is the ratio of plasma pressure to the magnetic fieldpressure) may deteriorate the confinement of trapped energetic ions in spherical tokamaks (ST)during MHD events, such as sawtooth oscillations and internal reconnection events (IRE). Thisresult indicates that moderate rather than very high β may be preferable in STs.
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420
(THP2/19) Drift-Kinetic Alfven Modes in High PerformanceTokamaks
A. Jaun1), A. F. Fasoli2), D. Testa2), J. Vaclavik3), L. Villard3)
1) Alfven Laboratory, Euratom-NFR Association, KTH, 100 44 Stockholm, Sweden2) Plasma Science Fusion Centre, MIT, Cambridge MA 02139, USA3) CRPP-EPFL, 1015 Lausanne, Switzerland
Abstract. The stability of fast-particle driven Alfven eigenmodes is modeled in high perfor-mance tokamaks, successively with a conventional shear, an optimized shear and a tight aspectratio plasma. A large bulk pressure yields global kinetic Alfven eigenmodes that are stabilizedby mode conversion in the presence of a divertor. This suggests how conventional reactor sce-narii could withstand significant pressure gradients from the fusion products. A large safetyfactor in the core q0 > 2.5 in deeply shear reversed configurations and a relatively large bulk ionLarmor radius in a low magnetic field can trigger global drift-kinetic Alfven eigenmodes thatare unstable in high performance JET, NSTX and ITER plasmas.
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421
(THP2/20) Energetic Particle Mode Dynamics in Toka-maks
F. Zonca1), S. Briguglio1), L. Chen2), G. Fogaccia1), G. Vlad1), L.-J. Zheng2)
1) ENEA C. R. Frascati, C.P. 65, 00044 Frascati, Rome, Italy2) Department of Physics and Astronomy, University of California, Irvine, CA 92717-4575, USA
Abstract. Energetic Particle Modes (EPM) are strongly driven oscillations excited via wave-particle resonant interactions at the characteristic frequencies of the energetic ions, ωtE , ωBEand/or ωdE , i.e., respectively the transit frequency for circulating particles and the bounce andprecessional drift frequencies for trapped ions. A sharp transition in the plasma stability atthe critical EPM excitation threshold has been observed by nonperturbative gyrokinetic codes interms of changes in normalized growth rate (γ/ωA, with ωA = vA/qR0), real frequency (ωr/ωA)and parallel wave vector (k‖qR0) both as α = −R0q
2β′ of the thermal plasma and that, αE offast ions are varied. The present work further explores theoretical aspects of EPM excitationsby spatially localized particle sources, possibly associated with frequency chirping, which canradially trap the EPM in the region where the free energy source is strongest. Results of a non-perturbative 3D Hybrid MHD Gyrokinetic code are also presented to emphasize that nonlinearbehaviors of EPM’s are different from those of Toroidal Alfven Eigenmodes (TAE) and KineticTAE (KTAE) and that particle losses and mode saturation are consistent with the mode-particlepumping model (particle radial convection). Results of theoretical analyses of nonlinear EPMdynamics are also presented and the possible overlap with more general nonlinear dynamicsproblems is discussed.
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422
(THP2/21) Analysis of Plasma Current Profile Driven byCoaxial Helicity Injection in Tokamak
Cheng Zhang1), Deng Zhou1),Sizheng Zhu1)
1) Institute of Plasma Physics, Chinese Academy of Sciences, China
Abstract. The plasma current profile driven by coaxial direct current helicity injection in a low-aspect-ratio toroidal configuration is investigated by applying the principle of minimum energydissipation rate. It is shown that the current profile modes are mainly decided by Langrangemultiplier β. Some critical values of βc are found. Different current profiles are obtained indifferent β ranges. Three typical current profiles are presented. The key features agree well withexperiment as the first case of β < 7. Large driven plasma current and a typical current profilein a normal tokamak can be obtained in the region of β = 7 to 9.65. There exist reversionsof both jϕ and Bϕ in the central part when β becomes higher than βc ≈ 9.65. For a selectedgeometry the values of βc weakly depend on other parameters. There exist critical values ofplasma temperature and bias voltage to induce the transformation of current profile mode. Theprofile of λ = µ0jϕ/Bϕ is non-uniform in the plasma and a larger deviation from force free statesappears when β becomes higher.
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423
(THP2/22) Non Inductive Current Drive and Beam PlasmaInteraction in Compact Tori
R. Farengo1), A. Lifschitz2), K. Caputti1), N. Arista1), R. A. Clemente3)
1) Centro Atomico Bariloche e Instituto Balseiro, Bariloche, RN, Argentina2) Consejo Nacional de Investigaciones Cientıficas y Tecnicas, Argentina3) Universidade Estadual de Campinas, Campinas, SP, Brasil
Abstract. The effect of a steady toroidal field (TF) on rotating magnetic field (RMF) currentdrive, the interaction of neutral beams with Field Reversed Configurations (FRC) and the re-laxed (minimum dissipation) states of tokamaks sustained by helicity injection are studied.
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424
(THP2/24) Analysis of IBW-Driven Plasma Flows in Toka-maks
L. A. Berry1), E. F. Jaeger1), E. F. D’Azevedo1), D. B. Batchelor1), J. A. Carlsson1), M. D. Carter1),R. Cesario2), AsOneFTU Team2)
1) Oak Ridge National Laboratory, Oak Ridge, Tennessee, United States2) Associazione EURATOM-ENEA sulla Fusione, Centro Ticerche Frascati, Frascati, Rome,Italy
Abstract. Both theory and experiment have suggested that damping of Ion Bernstein Waves(IBWs) at ion cyclotron frequency harmonics could drive poloidal flows and lead to enhancedconfinement for tokamaks. However, the early analyses were based on Reynolds stress closures ofmoment equations. More rigorous, finite Larmor radius (FLR) expansions of the radio frequency(RF) kinetic pressure for low harmonic interactions indicated that the Reynolds stress approxi-mation was not generally valid, and resulted in significant changes in the plasma flow response.These changes were largest for wave interactions driven by finite Larmour radius effects. Toprovide a better assessment of higher harmonic interactions and IBW flow drive prospects, theelectromagnetic (E&M) and RF kinetic force models are extended with no assumptions regard-ing the smallness of the ion Larmor radius. For both models, a spectral-width approximationwas used to make the numerical analysis tractable. In addition, it was necessary to includethe effects of plasma equilibrium gradients on the plasma conductivity and the RF-inducedmomentum in order to conserve energy and momentum. The analysis of high-harmonic IBWinteractions for TFTR and FTU parameters indicates significant poloidal flow shears (relative toturbulence correlation times) for power levels available in present experiments. Recent advancesin all-orders calculations of E&M fields in 2-D are also discussed.
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425
(THP2/25) Heating and Current Drive by Electron Bern-stein Waves in NSTX and MAST Type Plasmas
A. K. Ram1), A. Bers1), S. D. Schultz1), C. N. Lashmore-Davies2), R. A. Cairns3), P. Efthimion4),G. Taylor4)
1) Plasma Science & Fusion Center, M.I.T, Cambridge, MA 02139, U.S.A.2) Euratom - UKAEA Fusion Association, Culham Science Centre, Abingdon, Oxon, OX14 3DB,United Kingdom3) University of St. Andrews, St. Andrews, Fife, KY16 9SS, United Kingdom4) Princeton Plasma Physics Laboratory, Princeton, NJ 08543, U.S.A.
Abstract. The high β operating regime of spherical tokamaks (ST), such as in NSTX andMAST, make them attractive fusion devices. To attain the high β’s, there is a need to heatand to drive currents in ST plasmas. While ST plasmas are overdense to conventional electroncyclotron (EC) waves, electron Bernstein waves (EBW) offer an attractive possibility both forheating and for driving plasma currents. We consider techniques for the excitation of EBWs onNSTX and MAST-type plasmas. Emission of EBWs from inside the plasma and its conversionto the conventional EC modes at the plasma edge are also considered.
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426
(THP2/26) Global Analysis of ICRF Waves and AlfvenEigenmodes in Toroidal Helical Plasmas
A. Fukuyama1), E. Yokota1), T. Akutsu1)
1) Kyoto University, Kyoto, Japan
Abstract. The full wave code TASK/WM was extended to analyze the waves in a plasma withthree-dimensional inhomogeneity and was applied to the analysis of propagation and absorptionof the ICRF waves in the large helical device (LHD). We have studied various parameter de-pendence of radial power deposition profile and antenna loading impedance; frequency, minorityion density ratio, toroidal mode number. The dependence on the minority ion ratio agrees wellwith the experimental observation on LHD. The code also has a capability to describe the threekinds of weakly damped Alfven eigenmodes, GAE, TAE and HAE, in toroidal helical plasmas.We found several kinds of eigenmodes and studied their behavior.
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427
(THP2/27) Self-Consistent Modelling of ICRH
T. Hellsten1), J. Hedin1), T. Johnson1), M. Laxaback1), E. Tennfors1)
1) Division of Fusion Plasma Physics, Alfven Laboratory, Royal Institute of Technology, SE-100 44 STOCKHOLM, Sweden, Association EURATOM/NFR
Abstract. The performance of icrh is often sensitive to the shape of the high energy part ofthe distribution functions of the resonating species. This requires self-consistent calculations ofthe distribution functions and the wave-field. In addition to the wave-particle inte ractions andCoulomb collisions the effects of the finite orbit width and the rf-induced spatial transport arefound to be important. The inward drift dominates in general even for a symmetric toroidalwave spectrum in the centre of the plasma. An inward drift does not necessarily produce amore peaked heating profile. On the contrary, for low concentrations of hydrogen minority indeuterium plasmas it can even give rise to broader profiles.
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428
(THP2/30) 3D Fokker Planck Model for High and Moder-ate Energy Ions in Weakly Ripple Tokamaks
V. A. Yavorskij1), K. Schoepf2), V. Ya. Goloborod’ko1), J. W. Edenstrasser2)
1) Institute for Nuclear Research, Ukrainian Academy of Sciences, Kyiv, Ukraine2) Institut fur Theoretische Physik, Universitat Innsbruck, Association EURATOM-OEAW,Austria
Abstract. Fokker-Planck coefficients specifying diffusion and convection transport processesfor charged fusion products and NBI ions in tokamaks with weak TF ripples are derived in theCOM space describing the collisional ripple transport processes of fast ions, both in the moder-ate and in the high energy range. The collisional diffusion coefficient of toroidally trapped ionswhich are in resonance with the TF ripple perturbations (so-called superbananas) is shown toexhibit only a weak dependence on the ripple magnitude and, further, to reach a maximum inthe medium collisionality regime where the bounce frequency of superbananas is close to theeffective collision frequency. This maximum appears in the energy range from a few tens of kEVto a few hundreds of keV and may exceed the well-known Boozer- ripple plateau diffusion in thecase of weak ripples. The radial convection of superbananas induced by slowing down is shownalso to exceed the corresponding convection of bananas. The collisional superbanana transportis supposed to be responsible for the enhanced NBI ion loss observed in the intermediate energyrange and, hence, should be essentially embodied into any modeling of charged fusion productsat energies E < 1MeV.
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429
Session PD — Post DeadlinePapers
Contents
(PD/1) Overview of Recent JET Results and Future Perspectives . . . . . . 431
(PD/2) Full Absorption of 3rd Harmonic ECH in TCV Target PlasmasProduced by 2nd Harmonic ECH and ECCD . . . . . . . . . . . . . . . 432
(PD/3) Interactions between Neoclassical Tearing Modes and their Stabi-lization by an Externally Applied Helical Field . . . . . . . . . . . . . . 433
(PD/1) Overview of Recent JET Results and Future Per-spectives
J. Pamela1) and JET-EFDA Team1)
1) EFDA Close Support Unit, Culham Science Centre, Abingdon, Oxon, OX14 3EA, UK
Abstract. Recent JET results show progress in two directions: consolidation of the ITER refer-ence scenario, the ELMy H-mode, and development of regimes with internal transport barriers(ITBs). The beneficial effect of high triangularity in ELMy H-mode – one of the features ofthe new ITER design, has been confirmed in particular to extend the operation limits. Indeedthree different ELMy H-mode scenarios are shown to allow operation at or above the Greenwalddensity limit with limited or no confinement degradation: (1) operation at high triangularity (upto δ = 0.47) with controlled gas puffing; (2) operation well above the L-H transition threshold(Pin > 2.5PL−H) with controlled gas puffing; (3) impurity seeding. The level of confidence inthe ITER ELMy H-mode reference scenario has therefore increased further and is high; researchshould now focus towards milder ELMs. A very significant progress in LHCD coupling has beenachieved, with routine power in the 2-4 MW range. LHCD preheat allows to control efficientlythe q profile, giving access to strongly reversed shear plasmas, with very broad ITBs on Ti, Te
and ne, achieved at reduced power thresholds. ITBs are being developed, either towards highperformance (HβN ∼ 7.5 achieved) or towards steady state (full non-inductive current driveachieved during 1 second).
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431
(PD/2) Full Absorption of 3rd Harmonic ECH in TCV Tar-get Plasmas Produced by 2nd Harmonic ECH and ECCD
S. Alberti1), T. Goodman1), M. A. Henderson1), A. Manini1), J.-M. Moret1), P. Gomez1),P. Blanchard1), S. Coda1), O. Sauter1), C. Angioni1), K. Appert1), R. Behn1), P. Bosshard1),R. Chavan1), I. Condrea1), A. Degeling1), B. P. Duval1), D. Fasel1), J.-Y. Favez1), I. Furno1),F. Hofmann1), P. Lavanchy1), J. B. Lister1), X. Llobet1), Z. A. Pietrzyk1), A. Gorgerat1),P. Gorgerat1), J.-P. Hogge1), P.-F. Isoz1), B. Joye1), J.-C. Magnin1), B. Marletaz1), P. Marmil-lod1), Y. R. Martin1), A. Martynov1), J.-M. Mayor1), J. Mlynar1), P. Nikkola1), P. J. Paris1),A. Perez1), Y. Peysson2), R. A. Pitts1), A. Pochelon1), H. Reimerdes1), J. H. Rommers1), E. Scav-ino1), G. Tonetti1), M. Q. Tran1) and H. Weisen1)
1) Centre de Recherches en Physique des Plasmas, Association EURATOM-Confederation Su-isse, Ecole Polytechnique Federale de Lausanne, CH-1015 Lausanne, Switzerland2) Association Euratom-CEA sur la Fusion, DRFC, CEA-Cadarache, France
Abstract. An experimental study of the extraordinary mode (X-mode) absorption at the thirdcyclotron harmonic frequency (118GHz) has been performed on the TCV Tokamak in plasmaspreheated by X-mode at the second harmonic (82.7GHz). Various preheating configurationshave been experimentally investigated, ranging from counter-ECCD, ECH to CO-ECCD at var-ious power levels. Full absorption of the 470kW of injected X3 power was measured with as littleas 350kW of X2-CO-ECCD preheating. The measured absorption exceeds that predicted by thelinear ray tracing code TORAY by more than a factor of 2 for the CO-ECCD case. Experimentalevidence indicates that a large fraction of the X3 power is absorbed by electrons in an energetictail created by the X2-ECCD preheating.
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432
(PD/3) Interactions between Neoclassical Tearing Modesand their Stabilization by an Externally Applied HelicalField
Q. Yu1), S. Gunter1), K. Lackner1)
1) Max-Planck Institute of Plasma Physics, EURATOM Association, D-85748, Garching, Ger-many
Abstract. The interaction between the neoclassical tearing modes (NTMs) of different helici-ties is investigated theoretically. It is found that once two magnetic islands are close, the moreunstable island survives and suppresses the less unstable one, in agreement with the experi-mental results. The mechanism responsible for this suppression is the decreased fundamentalharmonic pressure perturbation of the NTM in the presence of magnetic perturbations of dif-ferent helicity. Based on the mechanism found, the effect of static helical magnetic fields on thenonlinear growth of neoclassical tearing modes (NTM) is investigated, and the NTM is found tobe stabilized by an externally applied helical field of a different helicity if the field magnitude issufficiently large, suggesting a very simple method for stabilizing the NTM.
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433
(PD/D) Discussion of Section PD
The file contains the discussion contributions relating to PD/1, PD/2, PD/3.
(PDP/1) Suppression of Pulsation by Laser Beam Smooth-ing and ICF with Volume Ignition
H. Hora1), R. J. Stening1), M. Aydin1), T. P. Rowlands1), F. Osman2), P. Evans2), R. Castillo3),M. Collins3), T. S. Gardener3), W. K. Chan3)
1) Department of Theoretical Physics, University of New South Wales, Sydney 2052, Australia.2) School of Science, University of Western Sydney Nepean, P.O. Box 10 Kingswood 2747, Aus-tralia.3) Department of Physics, University of Western Sydney Macarthur, P.O. Box 555 Campbelltown2560, Australia.
Abstract. A dominating mechanism responsible for the anomalies of the laser-plasma interac-tion at direct drive laser fusion is the 10 picosecond stochastic pulsation as recognised numericallyand experimentally since 1974 and measured in many details by Maddever and Luther-Davies(Australian National University, Canberra) few years ago. These fundamental new results arenow in the focus of interest in view of the present difficulties with the big laser-fusion facilities. Adrastic reconsideration and economic solution may be possible based on our recent detailed nu-merical studies which indicate that the stochastic pulsation can be suppressed by an appropriatesmoothing of the laser beam, permitting the operation with red light by saving expensive higherharmonic production avoiding damage by UV light, and providing much higher laser energy forfusion. By this way direct drive laser fusion will be favourable using the most robust volumeignition scheme with very high gain.
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436
(PDP/2) Verification of DBEs Selected for ITER EDA 1998Design by GEMSAFE Methodology and Consideration forSiting Events
T. Sawada1), M. Saito1)
1) RLNR, Tokyo Institute of Technology, Tokyo, Japan
Abstract. General Methodology of Safety Analysis and Evaluation for Fusion Energy Sys-tems (GEMSAFE) has been applied to the International Thermonuclear Experimental Reactor(ITER) design in the final stage of Engineering Design Activities (EDA) to select design ba-sis events (DBEs) and to identify related safety features and requirements to ensure its safety.We have classified DBEs into three categories considering their occurrence probabilities andexpected scales of their consequences. By the GEMSAFE methodology applied to the ITERfinal design, we have selected 21 DBEs: 8 in the category 1, 8 in the category 2 and 5 in thecategory 3. The selected DBEs were compared with the Reference Events which were addressedin the ITER non-site specific safety report (NSSR-2). As a result, it has been made clear thatthere is no significant differences between the GEMSAFE DBEs and NSSR-2 Reference Events.Furthermore, in the framework of the GEMSAFE methodology, we have proposed a concept ofsiting events selection with the suggestion for siting events as those that were beyond designbasis events developed by using function-based safety analysis (FBSA) method.
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437
(PDP/3) Recent Advances in Indirect Drive ICF TargetPhysics at CEA
J. Tassart1)
1) Commissariat a l’Energie Atomique, France
Abstract. The objective of Target Physics Program at CEA is the achievement of ignitionon the LMJ, a glass laser facility of 1.8 MJ which will be completed by 2008. It is composedof theoretical work, experimental work and numerical simulations. An important part of ex-perimental studies is made in collaboration with U.S. DOE Laboratories: Lawrence LivermoreNational Laboratory, Los Alamos National Laboratory and the Laboratory for Laser Energeticsat the University of Rochester. Experiments were performed on Phebus, NOVA (LLNL) andOMEGA (LLE) ; they included diagnostics developments. Recent efforts have been focused onLaser Plasma Interaction, hohlraum energetics, symmetry, ablator physics and hydrodynamicinstabilities. Ongoing work prepare the first experiments on the LIL which is a prototype facil-ity of the LMJ (8 of its 240 beams). They will be performed by 2002. Recent progress in ICFtarget physics allows us to precise laser specifications to achieve ignition with reasonable margin.
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438
(PDP/4) Steady Improved Confinement in FTU High FieldPlasmas Sustained by Deep Pellet Injection
D. Frigione1), E. Giovannozzi1), C. Gormezano1), F. Poli1), M. Romanelli2), O. Tudisco1),F. Crisanti1), B. Esposito1), L. Gabellieri1), L. Garzotti3), M. Leigheb1), D. Pacella1) and FTUTeam
Abstract. High density plasmas (no ∼ 8×1020m−3) featuring steady improved core-confinementhave been obtained in FTU at the maximum nominal toroidal field (8 T), and lower, by deepmultiple pellet injection. These plasmas featured also high purity, efficient electron-ion couplingand peaked density profiles sustained for several confinement times. Neutron yields in excess of1×1013n/s are measured, consistent with the reduction of the ion transport to neoclassical levels.
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439
(PDP/5) Plasma Performance Improvement by AdvancedHigh Field Side Pellet Injection in ASDEX Upgrade
A. Lorenz, P. T. Lang, J. C. Fuchs, J. Gafert, O. Gehre, O. Gruber, G. Haas, M. Kaufmann,B. Kurzan, M. Maraschek, V. Mertens, H. W. Muller, H. D. Murmann, J. Neuhauser, W. Schnei-der, ASDEX Upgrade Team
Abstract. Limited available pellet velocities so far restricted the plasma performance of theefficient launch scheme from the tokamak magnetic high field side. Although pellet injectionduring H-mode results in more peaked density profiles and enhanced performance with respectto gas puff refuelling, prompt particle and energy losses induced by pellet induced ELM burstsstill limited the operational area. An improved pellet injection set-up applied at ASDEX Up-grade allowed for the first time significantly higher injection velocities. In first proof-of-principleplasma discharges a successful scan of pellet injection at v = 240–1200 m/s and plasma param-eter studies at v = 560 m/s were completed. They resulted in deeper penetration and particledeposition mitigating fast particle and energy losses. Further extension of tokamak operation inthe high density regime seems feasible.
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440
(PDP/6) Particle Transport in High Power ECH and ECCDDischarges in TCV
H. Weisen1), I. Furno1), T. Goodman1)
1) Centre de Recherches en Physique des Plasmas, Ecole Polytechnique Federale de Lausanne,CH-1015 Lausanne, Switzerland
Abstract. A coupled heat and particle transport phenomenon, leading to particle depletionfrom the plasma core is observed in a variety of plasma conditions with centrally deposited ECHand ECCD in TCV. This phenomenon, which causes inverted sawteeth of the central densityin sawtoothing discharges and leads to stationary hollow profiles in the absence of sawteeth,has been linked to the presence of m/n = 1/1 MHD modes. In particular this phenomenon,known as “density pumpout” can be suppressed by stabilizing the mode by means of operationat high triangularity. The correlation of pumpout with the loss of axisymmetry suggests thatneoclassical transport processes involving locally trapped particles near the helically displacedmagnetic axis, previously believed to be important only in stellarators, may account for thephenomenon in tokamaks as well.
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441
(PDP/7) Synergy between LH and ECH Waves in the FTUTokamak
V. Pericoli-Ridolfini1), Y. Peysson2), R. Dumont2), G. Giruzzi2), G. Granucci3), L. Panaccione1),L. Delpech2), B. Tilia1), FTU Team1), FTU-ECH Team3)
1) Associazione Euratom-ENEA sulla Fusione, CR Frascati, 00044 Frascati, Rome, Italy2) Association Euratom-CEA sur la Fusion DRFC/SCCP, CEA/Cadarache, 13108 St. Paul-lez-Durance, France3) Associazione Euratom-ENEA-CNR sulla Fusione, Istituto di Fisica del Plasma, Via R. Cozzi20125, Milano, Italy
Abstract. The effects of the simultaneous injection of LH, up to 1 MW, and ECH power, upto 0.75 MW, have been studied in FTU. In the absence of the cold EC resonance in the plasma(BT = 7.2T), the suprathermal electron tail generated by LHCD absorbs the ECH power ef-fectively, up to 80%. In a restricted parameter range, additional electron heating due to ECHwaves (up to 1.2 keV for PECH = 0.7MW) is also observed. With the cold resonance at thecentre (BT = 5.3T) and no absorption on the tail, the ECH power injection into a sawteethand MHD free plasma, fully sustained by LHCD, causes a quasi-stationary large increase of theelectron temperature in a region r/a < 0.5, with ∆Te0 > 4keV for PECH = 0.35MW only. Thisstrong heating exceeds the predictions of the mixed Bohm/gyro-Bohm model with magneticshear correction, which, instead reproduces the LH phase well.
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442
(PDP/8) Commissioning of a DT Fusion Reactor withoutExternal Supply of Tritium
Y. Asaoka1), S. Konishi2), S. Nishio2), R. Hiwatari1), K. Okano1), T. Yoshida1), K. Tomabechi1)
1) Central Research Institute of Electric Power Industry, Japan2) Japan Atomic Energy Research Institute, Japan
Abstract. Commissioning of a DT fusion reactor without external supply of tritium is dis-cussed. The DD reactions in a DT-oriented fusion reactor with external power injection byneutral beams produce tritium and neutrons. Tritium produced by the DD reaction togetherwith that produced in the blanket by the 2.45 MeV neutron is re-circulated into the plasma.Then, the DT reaction rate increases gradually, as tritium concentration in plasma builds up to-wards the level of nominal operation. Time required to reach the nominal operational condition,i.e. 50 % tritium in plasma, is estimated with assumptions based on a model of fusion powerplant. As a result, the start-up period of a DT fusion reactor without external supply of tritiumis estimated to be approximately 55 days, with the plasma parameters of CREST having a highperformance blanket and tritium processing systems. Major factors to determine the start-upperiod are DD and DT reaction rates, net tritium breeding gain of the plant and dead inventoryin/on facing materials. Elimination of a constraint for fusion reactor deployment and operationwithout any tritium transportation in and out of plant through its entire life may be possible.
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443
Session S — Summaries
Contents
(S/1) Stability, Current Drive and Heating, Energetic Particles . . . . . . . 445
(S/1) Stability, Current Drive and Heating, Energetic Par-ticles
K. Razumova1)
1) Russian research Centre “Kurchatov Institute”, Moscow, Russia
Abstract. The paper summarizes the results presented at the conference Fusion Energy 2000(FEC 2000) in relation to the following subjects: 1. The possibility of realizing plasma param-eters for ITER needs, advanced regimes in tokamaks and stellarators. 2. Stability of plasmaswith an appreciable component of fast particles. 3. Low aspect ratio tokamaks. 4. New resultswith auxiliary heating and current drive methods. 5. β limit and neoclassical tearing mode(NTM) stabilization. 6. Internal transport barriers.
Abstract. In this paper the contributions presented at the 18th IAEA Fusion Energy Con-ference in the field of transport and boundary physics will be summarised with reference tothe following distinct issues: H-mode physics, Internal Transport Barrier formation, transportstudies, Radiative Improved modes and impurity seeding, divertor and He exhaust, new config-urations.
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446
(S/3) Inertial Fusion Energy
K. Mima1)
1) ILE, Osaka University, Suita, Osaka, Japan
Abstract. Reviewed is the present status of the inertial confinement energy (IFE) research.The highlights of the IFE presentations are as follows. Toward demonstrating ignition andburning of imploded plasmas, ignition facilities of mega jule class blue laser system are underconstruction at Lawrence Livermore National Laboratory and the CEA laboratory of Bordeaux.The central ignition by both indirect drive and direct drive will be explored by the middle of2010’s. A new ignition concept so called “fast ignition” has also been investigated intensively inthe last two years. Peta watt level (1PW∼0.1PW output) CPA lasers have been used for heat-ing solid targets and imploded plasmas. With 50J∼500J/psec pulses, solid targets are found tobe heated up to 300eV. They were measured by X-ray spectroscopy, neutron energy spectrum,and so on. Summarized are also researches on simulation code developments, target design andfabrication, heavy ion beam fusion, Z-pinch based X-ray source, and laser driver technology.
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447
(S/4) Technology and Power Plants
S. Milora1)
1) Oak Ridge National Laboratory, Oak Ridge, TN, U.S.A.
Abdou, M. A., 320Abdullaev, S. S., 335Abe, M., 117Acedo, P., 22Ackers, R., 31Adachi, H., 13Adachi, K., 77Afanasiev, N. M., 319Agullo, O., 352Airoldi, A., 209, 284, 359Aizawa, K., 131Akaishi, K., 77Akasaka, H., 13Akers, R. J., 144, 183Akiba, M., 50, 318Akiduki, T., 130Akimoto, H., 347Akino, N., 13, 85Akiyama, R., 17, 77, 94Akutsu, T., 427Albajar, F., 331Alban, D., 287Albanese, R., 231, 336Alberti, S., 37, 432Alejaldre, C., 34Aleksandrov, S. V., 125Alekseyev, A. G., 153Alexander, D. A., 81Alikaev, V. V., 149, 212, 325Alladio, F., 91Allen, S. L., 16, 60, 75, 126Almagri, A. F., 128Almoguera, L., 34Alon, U., 258Alonso, E., 275Alonso, J., 34Alper, B., 143Ambrosino, G., 231, 336Amemiya, T., 313Amrollahi, R., 330Anderson, D. T., 128Anderson, F. S. B., 128Anderson, H., 228Anderson, J. K., 41Andersson, F., 368
Andiel, U., 265Ando, T., 48Andreiko, M. V., 214Ane, J. M., 331Ang, W. L., 131Angelini, B., 68, 91, 209Angioni, C., 37, 155, 359, 432Aniel, T., 89, 233Annou, K., 13Antar, G., 316Antipenkov, A., 47, 241, 242Anton, M., 33, 101Antoni, V., 69, 205Apicella, M. L., 68, 91, 209Apolloni, L., 69, 205Appel, L. C., 67, 144, 183Appert, K., 37, 381, 432Apruzzese, G., 68, 91, 209Arai, K., 48Arai, T., 13Arakawa, H., 193Arakawa, K., 13Araki, M., 338Araki, T., 346Araya, F., 346Arazi, L., 258Arends, E., 183Ariola, M., 231, 336Arista, N., 424Arkhipov, I., 50Arsenin, V. V., 279Arzhannikov, A. V., 135Asahara, H., 270Asai, T., 181Asakawa, M., 152Asakura, N., 13, 83, 85, 86, 216, 220, 240Asaoka, Y., 348, 443Ascasıbar, E., 34ASDEX Upgrade Team, 21, 76, 100, 162,
182, 221, 222Askinazi, L. G., 214Astrelin, V. T., 135Auluck, S. K. H., 282
458
Aumayr, F., 228Austin, M. E., 75, 90, 107Axon, K. B., 67Aydemir, A., 391Aydin, M., 436Aymar, R., 10, 46Azechi, H., 26, 270, 271Azizov, E. A., 185, 328Azumi, M., 13
Baciero, A., 34Bae, Y. D., 297Bae, Y. S., 297Baumel, S., 33Bagatin, M., 69, 205Bahr, R. E., 28Bak, J. G., 298Bak, P. E., 13Baker, C. C., 341Baker, D. R., 75, 81, 93, 199, 217Baker, K. L., 178Baker, W., 69, 205Bakhtiari, M., 13, 116Bakos, J., 37Balbın, R., 34, 203Baldzuhn, J., 33Balescu, R., 382Baluc, N., 308Bangerter, R. O., 27Bao, Y., 175Baonian, W., 207Barabash, V., 50Barana, O., 69, 205Barbato, E., 68, 91, 209, 336Barry, R. E., 31Bartels, H.-W., 245Barth, C. J., 154Bartiromo, R., 69, 205Bastasz, R., 189Batani, D., 268Batchelor, D. B., 425Bateman, G., 81, 367Baylor, L. R., 17, 81, 90, 151, 182, 199, 200Beer, M. A., 355Beg, F. N., 257Behn, R., 37, 155, 432Beidler, C. D., 54, 419Bell, M., 31, 70, 111, 127Bell, R., 31, 70, 127, 172Belousov, S. I., 324Belov, A. M., 153Belova, E. V., 405Ben-Dor, G., 258Benisti, D., 362Benkadda, S., 352
Bergmann, A., 202Bergsaker, H., 159Bernabei, S., 107, 119, 139Bernal, L., 293Bernardinello, A., 268Berry, L. A., 104, 287, 333, 425Bers, A., 31, 426Bertalot, L., 68, 91, 209Bertocchi, A., 68, 91, 209Bessette, D., 48Besshou, S., 131Betti, R., 28Bettinali, L., 300Bettini, P., 69, 205Beurskens, M. N. A., 154Bevilacqua, G., 48Beyer, P., 352Bialek, J., 31, 70, 146, 161, 364Bian, N., 352Bibet, P., 243Biewer, T. M., 41Bigelow, T., 31Bilbao, L., 293Bitter, M., 31, 70, 172Bittner, R., 299Blackwell, B. D., 124Blanchard, P., 37, 432Blanchard, W., 31Bland, S. N., 257Blaumoser, M., 34Boboc, A., 69, 205Boedo, J. A., 60, 81, 126, 187, 217, 218, 400Boehly, T. R., 28Bogatu, I. N., 364Boissin, J.-C., 242Boivin, R. L., 22, 61, 86, 165Bolshukhin, D., 188Bolzonella, T., 69, 78, 205Bombarda, F., 284Bondeson, A., 409Bonnin, X., 184, 397Bonoli, P. T., 22, 31, 42, 165, 172Boozer, A. H., 104, 146, 161, 333Bora, M. P., 410Borba, D. N., 144, 147Borchardt, M., 397Borg, G. G., 124Borrass, K., 184Borshegovskij, A. A., 149, 212, 213Bosch, H.-S., 184Bosia, G., 243Bosshard, P., 37, 432Boswell, C., 22Botija, J., 34
459
Bouquey, F., 166Bourdelle, C., 89Branas, B., 34Braams, B., 184Bracco, G., 68, 91, 209, 233Brakel, R., 33, 101Brambilla, M., 22, 165Bravenec, R., 22, 355Bray, B., 75Brennan, D., 364Brennen, D. P., 75Bretz, N., 22Brezinsek, S., 82Briguglio, S., 422Brix, M., 84Brooks, A., 104, 333Brooks, J. N., 189Brooks, N. H., 60, 81, 126, 189Brooks, R. D., 112Brower, D., 41Brunsell, P. R., 159Bruschi, A., 68, 91, 209, 336, 359Brzozowski, J. H., 159Bucalossi, J., 133Buceti, G., 68, 91, 209Buchaneauer, D., 316Budny, R., 139, 355Budny, R. V., 84, 89, 215Buerbaumer, H., 184Buffa, A., 69, 205Bulanin, V. V., 214Bulmer, R. H., 231, 285Bunting, C. A., 67Buratti, P., 68, 91, 98, 209, 359Burdakov, A. V., 135Burhenn, R., 33Burrell, K. H., 75, 81, 90, 93, 107, 151, 200,
217Burtseva, T. A., 125Bush, C., 233Busigin, A., 241Buttery, R. J., 67, 147, 150Buzhinskij, O. I., 328Byrom, C., 183
Cabal, H., 349Cairns, R. A., 426Cakare, L., 299Califano, F., 359Callaghan, H., 33Callahan-Miller, D., 255Callen, J. D., 364Callis, R. W., 107Campbell, D. J., 234Candela, G., 336
Canton, A., 69, 78, 205Cao, J. Y., 134, 177, 204Cappa, A., 34Cappello, S., 69, 205, 362Caputti, K., 424Cardella, A., 49, 247Cardinali, A., 68, 91, 209Carlsson, J. A., 425Carlstrom, T. N., 60, 217, 370, 389Carolan, P. G., 67, 183, 223Carraro, L., 69, 205Carrasco, R., 34Carreras, B. A., 22, 203, 377, 383, 406, 411Carter, M. D., 31, 172, 425Cary, J. R., 81Cary, W. P., 107Cascino, S., 91Casper, T. A., 75, 90Castaldo, C., 91Castejon, F., 34Castillo, R., 436Cates, C., 161Cattanei, G., 33Catto, P. J., 408, 409Causey, R., 316Cavazos, T., 286Cavazzana, R., 69, 205Cavinato, M., 231Cecconello, M., 159Cenacchi, G., 284Centioli, C., 68, 91, 209Cepero, J. R., 34Cerdan, G., 252Cesario, R., 68, 91, 209, 336, 425Champeaux, S., 351Chan, V. S., 42, 151Chan, W. K., 436Chance, M. S., 151, 364, 404Chandra, D., 410Chang, C. S., 389Chapman, B. E., 41Charles, C., 124Charlet, M., 84Chatterjee, R., 22Chattopadhyay, P. K., 41Chavan, R., 37, 243, 432Chen, L., 371, 422Chen, Y. F., 208Chen, Y. P., 334Cheng, C. Z., 13, 119, 141Cheng, E. T., 342Cherepnin, Yu. S., 307, 328Chernenko, A., 263Chernov, V., 329
460
Chiba, S., 13Chikaraishi, H., 17, 77, 310Chiocchio, S., 50, 252Chistiakov, V. V., 213Chistyakov, V., 212Chitarin, G., 69, 205Chittenden, J. P., 257Cho, B. H., 399Cho, M. H., 297Cho, S., 296Cho, T., 132, 206Cho, Y. S., 297Choi, B. H., 297Chrzanowski, J., 31Chu, M. S., 75, 126, 146, 151, 364, 404Chudnovskij, A., 233, 235Chugunov, I. N., 125Chuillon, P., 91, 336Chukbar, K., 263Chumanov, A. N., 319Chung, T., 22Chung, Y. S., 298Chuyanov, V., 10, 46Ciattaglia, S., 68, 91, 209, 329Ciazynski, D., 309Cirant, S., 68, 91, 209, 336, 359Claassen, H. A., 394Claesen, R., 336Clark, D., 286Clemente, R. A., 424Coccorese, E., 231Cocilovo, V., 68, 91, 209, 336Coda, S., 37, 432Coelho, R., 68Coffey, I., 84Coffey, S. K., 286Cohen, B. I., 285, 378Cohen, R. H., 81, 370, 396Colchin, R. J., 60, 81, 126Cole, M. J., 287Coletti, A., 336Collins, M., 436Collins, T. J. B., 28Collis, S. M., 124Combs, S. K., 17, 182, 200Comer, K., 364COMPASS-D Team, 223COMPASS-ECRH Team, 223Condrea, I., 37, 432Conn, R. W., 218Connor, J. W., 223, 363, 368Constantinescu, B., 306Conway, G., 76, 202Conway, N. J., 67, 183, 223
Cook, I., 344Cooper, W. A., 104, 157, 360, 381Coppi, B., 284, 390Cordey, G., 84Cordey, J., 233Corre, Y., 186Costa, S., 69, 205Coster, D. P., 184, 187, 239, 240Costley, A. E., 238Cote, A., 233Courtois, L., 166Cowan, T. E., 267Cox, M., 183Cox, S., 228Cozzani, F., 329Craig, D., 41, 160Crawford, E., 112, 288Craxton, R. S., 28Cremy, C., 34Crescenzi, C., 336Crisanti, F., 68, 91, 209, 336, 439Crocker, N. A., 160Crocker, N. C., 41CT-6B Team, 40Cucchiaro, A., 336Cui, Z., 13Cunningham, G., 67
D’Angelo, F., 69, 205, 362D’Azevedo, E. F., 425D’Ippolito, D. A., 370Daenner, W., 49, 247Dahi, H., 161Daido, H., 270Dal Bello, S., 69, 205Damiani, C., 252Dangor, A. E., 257Danko, S., 263Danner, W., 246Darbos, C., 166Darrow, D., 31, 117, 389Das, A., 351Daum, E., 329Davidson, R. C., 27Davis Lee, W., 165Davis, J. W., 189Davis, W., 31Dawson, J. M., 400Day, C., 242Daybelge, U., 394De Angelis, R., 68, 91, 209, 336de Baar, M. R., 154De Benedetti, M., 68, 91, 209de Groot, B., 154de Kloe, J., 154, 210
461
de la Cal, E., 34de la Luna, E., 34, 91de la Pena, A., 34De Lorenzi, A., 69, 205De Luca, F., 227De Marco, F., 91De Michelis, C., 186de Vries, P., 17, 77, 95, 110, 115, 118, 157,
182, 221, 222DeBoo, J. C., 75, 81, 90, 199, 233Dech, A. V., 125Decool, P., 309Decyk, V. K., 400Degeling, A., 37, 432Degnan, J. H., 286deGrassie, J. S., 107, 389del Rio, J. G., 275Delettrez, J. A., 28Delpech, L., 442Den Hartog, D. J., 41, 160, 288Dendy, R. O., 144Deng, Z. C., 134, 204Denisov, G. G., 325Deranian, R. D., 217Desai, T., 268Desideri, D., 69, 205Detragiache, P., 284Dettrick, S. A., 13Devynck, P., 171Di Pietro, E., 230Diamond, P. H., 351, 352DIII-D Team, 16, 60, 81, 146, 200Dimits, A. M., 378Ding, B. J., 175Ding, X. T., 134, 204Dini, F., 330Dittrich, T., 269Dlougach, E. D., 279, 314Dnestrovskij, A., 233Dnestrovskij, Yu., 198, 212, 213Do, C. J., 296Doerner, R., 189, 316Dokouka, V. N., 328Domınguez, E., 275Donaldson, W. R., 28Doncel, J., 34Dong, J. F., 134, 177, 204Dong, J. Q., 385Donne, A. J. H., 154, 238Dorland, B., 289Dorland, W., 22, 355Dorst, D., 33Dowling, J., 67Doyle, E. J., 31, 75, 81, 90, 93, 126, 217
Drake, J. R., 159Dremin, M., 212Dreval, V., 82Drevlak, M., 103, 104, 292Duchateau, J. L., 309Dudek, L., 31Dumbrajs, O., 393Dumont, R., 166, 442Dumortier, P., 82, 84Dunstan, M. R., 67Duquerroy, R., 155Duval, B. P., 37, 187, 432Dux, R., 226, 227Dvorkin, N. Ya., 328
Ebihara, N., 13Ebisawa, K., 238Edenstrasser, J. W., 429Edgell, D. H., 364Edlington, T., 67Edwards, A. W., 143Efthimion, P., 31, 316, 426Egorov, S., 48Eguilior, S., 34Ehmler, H., 33Ehrlich, K., 329Eich, Th., 335Eidmann, K., 265Eisner, E., 22Ejiri, A., 17, 130Ekedahl, A., 133Elbaz, Y., 258Elio, F., 49Eliseev, L., 212Ellis, R., 31Elsner, A., 33Emoto, M., 17, 77, 95, 110, 115, 118, 157,
182, 221, 222Endler, M., 33Enoeda, M., 247Entrop, I., 142Epstein, R., 28Erckmann, V., 33Erents, S. K., 84Eriksson, L.-G., 368Ernst, D. R., 81, 83, 90, 200, 355Escande, D. F., 69, 205, 362Esipchuk, Yu. V., 149, 212, 233Esposito, B., 68, 91, 209, 439Estrada, T., 34Evans, P., 436Evans, R. W., 178Evans, T. E., 81, 126Evtikhin, V. A., 185, 319Ewig, R., 111
462
Ezato, K., 50, 318Ezumi, N., 193
Faehl, R. J., 286Fahrbach, H.-U., 58, 227Falchetto, G. L., 381Falter, H. D., 228Farengo, R., 424Fasel, D., 37, 432Fasoli, A. F., 140, 421Fattorini, L., 69, 205Faulconer, D. W., 335Favez, J.-Y., 37, 432Favre, A., 37Federici, G., 47, 50Fedulov, M., 263Feng, B. B., 177Feng, Y., 33, 192Feng, Z., 177Fenstermacher, M. E., 60, 126Fenzi, C., 81, 90, 93Ferdinand, R., 329Fermani, G., 336Fernandez, A., 34Ferron, J. R., 31, 75, 151, 217, 364Fiedler, S., 33Field, A. R., 67, 223Fielding, S. J., 67, 223, 233Figarella, C., 352Fiksel, G., 41, 160, 379Filatov, O. G., 328Fill, E., 265Finken, K. H., 142, 335Finkenthal, M., 31, 211, 316Fiore, C., 22, 61, 165Fiorentin, P., 69, 205Fisher, P. W., 17, 182Fitzpatrick, R., 379Fogaccia, G., 422Fonck, R. J., 93Fontana, P. W., 160Forest, C. B., 41Fournier, K. B., 211Fowler, R. H., 287Francioni, B., 91Franz, P., 69, 205Franzen, P., 228Fredd, E., 31Fredrickson, E., 31, 70, 104, 111, 127, 139,
146Freeman, R. R., 267Frenje, J., 28Frese, M. H., 286Friconneau, J.-P., 252Frigione, D., 68, 91, 209, 439
182, 221, 222Funahashi, A., 13Funatsu, M., 262Furno, I., 37, 155, 359, 432, 441Furukawa, H., 13, 321, 322Furuya, K., 247
Gabellieri, L., 68, 91, 209, 211, 439Gadelmeier, F., 33Gafert, J., 440Gahn, C., 265Gaio, E., 69, 205Gale, D., 286Galkin, S. A., 364Galkowski, A., 283Gandini, F., 68, 209Gandy, R., 22Gangadhara, S., 22Gantenbein, G., 66Gao, X., 175, 176, 208Garbet, X., 89, 93, 352Garcıa, A., 34Garcıa-Cortes, I., 34Garcia, L., 203Gardener, T. S., 436Garnier, D. T., 408Garofalo, A. M., 75, 146, 364
463
Garzotti, L., 69, 205, 439Gasparotto, M., 336Gates, D., 31, 70, 111, 127, 172Gatti, G., 68, 91, 209Gatto, R., 379Gauvreau, J.-L., 129Gavrilov, G. A., 125Gee, S., 183Gehre, O., 440Geier, A., 188Geiger, J., 33, 101, 168Gelles, D. S., 302Gentle, K., 22Gerstenberg, H., 248Ghendrih, P., 23, 186, 352Giannella, R., 186Giannone, L., 33Gibney, T., 31Gil, C., 133Gil, J. M., 275Gill, R. D., 143Gilmore, M., 129Gimblett, C. G., 363Giovannozzi, E., 68, 91, 209, 439Giruzzi, G., 89, 91, 166, 442Giudicotti, L., 69, 205Gladush, G. G., 328Glass, F. J., 124Glasser, A. H., 70Glebov, V. Yu., 28Glugla, M., 241Glukhikh, V. A., 328Gnesotto, F., 69, 205Godwal, B. K., 274Goetz, J. A., 22, 61, 86, 165Gohil, P., 75, 90, 200, 217Golant, V. E., 125, 214Goldstein, W. H., 211Goldston, R., 31, 104Golingo, R. P., 288Goloborod’ko, V. Ya., 399, 429Golovski, I., 265Golubchikov, L. G., 319Gomez, P., 37, 155, 432Goncharov, V. N., 28Gondhalekar, A., 144Goniche, M., 171Goodman, T., 37, 155, 173, 432, 441Gorbunov, E. P., 314Gordon, C. W., 245Gorelenkov, N. N., 119, 139, 141, 144Gorelenkova, M. V., 144Gorelov, I. A., 75Gorelov, Y., 107
439Gorshkov, A. V., 213Gotchev, O. V., 28Goto, M., 17, 77, 95, 110, 115, 118, 157, 182,
221, 222Goto, S., 181Goto, T., 179Gott, Yu. V., 149, 213Gouge, M. J., 182Gough, A., 124Gourdain, P.-A., 129Grabovsky, E., 263Gram, R. Q., 28Grandgirard, V., 331, 352Grando, L., 69, 205Granetz, R. S., 22, 61, 165Granucci, G., 68, 91, 209, 336, 442Grashin, S., 212Grasso, D., 359Gravanti, F., 91Graves, J. P., 144Gray, D. S., 218Greenfield, C. M., 75, 81, 90, 107, 126, 199,
200Greenwald, M. J., 22, 61, 165, 233, 355Greenwood, D. E., 81Gribov, Y. V., 231Grigull, P., 33, 192Grisham, L. R., 13, 31, 109, 312Grisolia, C., 133, 186Groebner, R. J., 60, 75, 81, 90, 200, 217Grolli, M., 68, 91, 209Grosman, A., 186Gruber, O., 21, 58, 76, 202, 440Gruzinov, I., 351Gryaznevich, M., 146, 183Gu, P., 180Gu, X. M., 208Guasp, J., 34Gude, A., 66, 76, 100, 226Gueits, J. C., 286Gunter, S., 66, 76, 100, 150, 433Guilhem, D., 133Guirlet, R., 186Gulden, W., 245Gunn, J., 133, 171, 186Guo, G. C., 134, 204Guo, H. Y., 112Guo, S. C., 69, 205
464
Gupta, N. K., 274, 282Gusev, V. K., 125
Haan, S., 269Haange, R., 251, 252Haas, G., 440Haas, H., 242Haasz, A. A., 189Haba, K., 17, 77Habara, H., 254, 266Habs, D., 265Hacker, H., 33Hahm, T. S., 351, 353Haines, M. G., 257Hallatschek, K., 372Hallberg, B., 349Hallock, G. A., 22Hamacher, T., 349Hamada, T., 131Hamada, Y., 17, 63, 77, 94, 95, 110, 118,
157, 182, 222Hamaguchi, S., 17, 77Hamamatsu, K., 13, 148, 167Hamano, T., 13Hammel, B. A., 269Hammett, G. W., 355Hanada, K., 39, 169, 190, 219Hanada, M., 313Hanatani, K., 131Hanawa, H., 48Hansen, A. K., 160Hara, E., 48Harada, J., 48Harding, D. R., 28Harmeyer, E., 54Harris, J. H., 124Hartfuß, H. J., 33Hartmann, D., 17, 33, 110Harvey, R. W., 41, 107, 139Hasegawa, M., 39, 219Hashimoto, H., 281Hastie, R. J., 144, 363, 408, 409Hastik, R., 248Hatae, T., 13, 151, 216, 232Hatano, T., 247Hatcher, R., 31, 104, 333Hatchett, S., 255, 267Hattori, A., 13Hattori, N., 193Hatzky, R., 103Hauser, H., 248Hawkes, N., 228Hawkes, N. C., 84Hawryluck, R., 31Hayase, K., 78, 191
Hayashi, K., 315Hayashi, N., 13, 395Hayashi, T., 415, 416Hayashiya, H., 31, 158Hedin, J., 428Hegna, C. C., 160, 379Heidbrink, W. W., 75, 90Heikkinen, J. A., 393Heitzenroeder, P., 341Helander, P., 183, 223, 368Hellsten, T., 428Hender, T. C., 147Henderson, M. A., 37, 155, 432Her, N. I., 296Herranz, J., 34Herrmann, A., 58, 240Herrmann, M., 255Herrmannsfeldt, W. B., 27Herrnegger, F., 54Herz, W., 48Heya, M., 26, 254Hicks, D. G., 28Hidalgo, C., 34, 203Higashijima, S., 13, 83, 85Hikida, S., 13Hildebrandt, D., 33Hill, D. N., 285Hill, K., 13, 31, 83, 119Hillis, D. L., 84Himura, H., 137, 290Hinkel, D., 269Hino, T., 303Hinssen, H. K., 323Hinton, F. L., 351Hiranai, S., 13Hirano, Y., 78Hirata, M., 132, 206Hiratsuka, H., 13Hirohata, Y., 303Hirose, A., 195Hirsch, M., 33, 101Hirshman, S., 104, 287, 333, 343, 360Hiwatari, R., 348, 443Hiyama, M., 251Hiyama, T., 48HL-1M Team, 40, 177, 204Hoang, G. T., 89, 166, 233Hobirk, J., 76, 100Hoek, H. M., 13Hoffman, A. L., 112Hoffman, D., 316Hofmann, F., 37, 155, 432Hogan, J., 81, 84, 186Hogan, W. J., 256
465
Hogeweij, D., 233Hogeweij, G. M. D., 154, 210Hogge, J.-P., 37, 432Hojo, H., 132, 206Holcomb, C. T., 285Holland, C., 351Holly, D. J., 288Holzhauer, E., 33Honda, A., 13Honda, M., 13Honda, T., 245, 251Hong, B. G., 297Hong, G. H., 296Hong, W. Y., 177, 204Hooper, E. B., 285Hopcraft, K. I., 144Hora, H., 436Horacek, J., 187Horiike, H., 281Horton, L. D., 57, 84, 232, 240Horton, R. D., 178Horton, W., 89, 367, 391Hosea, J. C., 22, 31, 107, 111, 139, 172Hoshi, Y., 13Hoshino, K., 170Hosogane, N., 13, 48Hosokawa, M., 395Houlberg, W. A., 81, 104, 199, 200, 341Howard, J., 124Howell, D. F., 147Hsieh, C., 200Hsieh, C.-L., 75HT-7 Team, 40, 123, 175, 176, 194, 207, 208Hu, L., 13Hu, L. Q., 208Hua, H., 134Hua, X., 204Huang, Q. Y., 334Hubbard, A., 22, 61, 165, 232, 233, 409Hudson, S., 104, 407Huget, M., 46Hughes, J., 22, 61, 86Huguet, M., 10, 45, 48Humer, K., 248, 299Humphreys, D. A., 364Hussey, T. W., 286Hutchinson, I. H., 22, 61, 165, 409Huysmans, G. T. A., 147, 150Hwang, C. K., 297Hwang, D. Q., 178Hwang, S. M., 298Hyatt, A. W., 60, 75, 81
Iannone, F., 91Ibbott, C., 47, 50
Ichige, H., 13Ichiguchi, K., 157, 222, 406, 411Ichimura, M., 132, 206Ida, K., 17, 77, 94, 95, 110, 115, 118, 157,
221, 222Ido, T., 63, 170Idomura, Y., 13, 356Igarashi, K., 13Igitkhanov, Yu., 54, 223, 232, 240Iguchi, H., 94Iida, M., 156Iiduka, S., 115Iima, M., 17, 77Iio, S., 17Iiyoshi, A., 77Ijiri, Y., 131Ikeda, K., 17, 77, 95, 110, 115, 118, 157, 182,
221, 222Ikeda, Y., 315Ikeda, Y. I., 13, 109, 131, 148, 167, 250Ikegawa, T., 272Ilieva, Kr. D., 324Im, K. H., 296Imagawa, S., 17, 77, 310Imahashi, K., 48Imai, T., 167, 315Imbeaux, F., 58, 166In, S. R., 296, 297In, Y., 409Inabe, T., 245Inagaki, S., 17, 77, 95, 110, 115, 118, 157,
182, 221, 222Ingesson, L. C., 84, 143Innocente, P., 69, 205Inomoto, M., 102, 181Inoue, A., 13Inoue, N., 17, 77, 95, 110, 115, 118, 157, 221,
222, 281Inoue, T., 230Intrator, T. P., 286Intravaia, A., 69, 205Ioki, K., 49, 243, 246, 247IPP-NBI Group, 101Iqbal, M., 290Irby, J., 22, 61, 165Irving, M., 252Isaev, M., 104Isaka, M., 13Isayama, A., 13, 73, 99, 107, 109, 148, 151,
167, 198, 216
466
Isei, N., 13, 117, 148, 233Iseli, M., 245Ishida, K., 291Ishida, S., 13Ishigooka, T., 48Ishii, K., 13, 132, 179, 206Ishii, Y., 13, 99Ishijima, T., 13, 83Ishikawa, M., 13, 141Ishizawa, A., 13, 414Isler, R. C., 126, 189Isobe, M., 17, 77, 94, 95, 110, 115, 118, 157,
182, 221, 222, 292, 317Isono, T., 48Isoz, P.-F., 37, 432Itagaki, T., 158Itakura, A., 132, 179, 206Itami, K., 13, 83, 85, 220, 240ITER Joint Central Team, 10, 45, 46, 49,
234, 249Ito, A., 290Itoh, K., 17, 77, 94, 95, 110, 115, 118, 222,
292, 392, 412Itoh, S., 39, 169, 190, 219Itoh, S.-I., 392, 412Itoh, T., 13, 109, 312Ivanov, I. A., 135Ivanov, N. V., 138, 213, 234Iwahashi, T., 13Iwamoto, A., 17, 77Iwasaki, K., 13Iwase, M., 13Izawa, Y., 26, 254, 261, 266, 270, 321Izumi, N., 26, 254, 271Izzo, V. A., 180
Jaanimagi, P. A., 28Jacchia, A., 68, 209, 227Jachmich, S., 84, 218Jackson, G. L., 81, 84, 90, 151, 189Jaeger, E. F., 425Jaenicke, R., 33, 101Jakubka, K., 171Jakubowski, L., 278Jameson, R. A., 329Janeschitz, G., 47, 50, 232, 238–240Jarboe, T. R., 31, 111, 180, 285Jardin, S. C., 31, 111, 127, 340, 341, 405Jaspers, R., 82, 142Jaun, A., 140, 421Jayakumar, J., 107Jayakumar, R., 48, 75, 81Jenko, F., 355, 386Jensen, T. H., 146Jeong, S. H., 297, 298
Jergigan, T. C., 81Jernigan, T. C., 90, 200JET Team, 15, 57, 59, 62, 74, 98, 201, 215,
225JET-EFDA Team, 431Jewell, P. E., 180JFT-2M Group, 63, 117, 174Ji, H., 31, 111, 405Jiang, Y., 41Jie, Y. X., 208Jimenez, J. A., 34, 203Jitsukawa, S., 329Jitsuno, T., 26, 254, 270, 321Johner, J., 331Johnson, D., 22, 31Johnson, L. C., 146Johnson, M. F., 143Johnson, T., 428Johzaki, T., 273Jones, B., 316Jones, L., 246, 247Jones, O., 269Jones, R. H., 301Jones, T. T. C., 228Jong, R. A., 81Jost, G., 381Jotaki, E., 39, 169, 190, 219Joye, B., 37, 432JT-60U Team, 83, 85, 92, 99, 109, 116, 141,
182, 221, 222Kaita, R., 31, 111, 316Kajiwara, K., 13, 109, 148, 167, 315Kajiyama, E., 13Kakudate, S., 251Kakurin, A. M., 138, 149, 213Kakuta, T., 305Kalinin, Yu., 263Kalish, M., 31Kallenbach, A., 58, 84, 86, 184, 188, 226,
182, 221Kang, H. S., 297Kanki, T., 181Kanzaki, T., 261Kardaun, O. J., 233, 235Karelse, F. A., 154Karger, F., 33Kariya, T., 315Kartoon, D., 258Karttunen, S., 393Kasahara, H., 130Kasai, S., 238, 305Kashiwabara, T., 13Kashiwagi, M., 313Kasugai, A., 167, 315Kasugai, Y., 250Kasuya, K., 262Kasuya, N., 130, 170Katanuma, I., 132, 179, 206, 369Kato, S., 182, 260Kato, T., 48Kato, Y., 321Katoh, Y., 301Katsuki, Y., 179Katsurai, M., 102, 158Kaufmann, M., 58, 184, 440Kaushik, T. C., 274, 282Kavin, A., 231Kaw, P. K., 354, 410Kawabe, M., 48Kawahata, K., 17, 77, 95, 110, 115, 118, 157,
182, 221, 222Kawai, M., 13, 109, 312Kawamata, Y., 13Kawamura, H., 247Kawamura, N., 291Kawamura, T., 26, 266Kawano, K., 48Kawano, Y., 13, 216Kawasaki, S., 39, 169, 190, 219Kawasaki, T., 48, 270Kawashima, H., 117, 170Kawashima, T., 261Kawazome, Y., 131Kaya, Y., 348Kaye, S., 31, 70, 111, 119, 127, 233Kayruthdinov, R. R., 37Kazarian, F., 133Kazawa, M., 13Keck, R. L., 28Kelly, J., 28Kendl, A., 54, 386
Kesner, J., 408Kessel, C., 104, 341, 360Key, M. H., 267Khayrutdinov, R., 116, 231, 328Khimchenko, L., 212, 213Khlopenkov, K., 17, 77, 95, 110, 115, 118,
157, 182, 221, 222Khorasani, S., 330Khramenkov, A. V., 213Khripunov, B. I., 319Kick, M., 33Kii, T., 281Kikuchi, H., 13Kikuchi, M., 13Kim, B. C., 296, 298Kim, D. L., 296Kim, G. H., 296Kim, J.-W., 184Kim, J. B., 296Kim, J. S., 364Kim, J. Y., 298Kim, K. R., 297Kim, W. C., 296Kim, Y. C., 296Kimura, A. K., 302Kimura, H., 117, 170, 174Kimura, T., 13Kingsep, A., 263Kinsey, J. E., 75, 81, 90, 93, 227, 367Kirkpatrick, R., 286Kirnev, G., 212Kirneva, N., 149, 212, 213Kirov, K., 76Kirpitchev, I., 34Kishimoto, Y., 13, 356, 361, 391, 415Kishony, R., 258Kislov, A., 212Kislov, D. A., 38, 149, 212Kislyakov, A., 238Kissick, M. W., 400Kisslinger, J., 33, 54, 192Kitagawa, S., 17, 77Kitagawa, Y., 26, 254, 266Kitamura, H., 270Kitamura, S., 13Kitano, K., 181Kitsunezaki, A., 13Kiuttu, G. F., 286Kiviniemi, T., 393Kiyama, S., 78, 191Kizu, K., 13Kleiber, R., 103Klein, H., 329Klima, R., 171
468
Klimanov, I. V., 149Klimenko, E. Yu., 314Klose, S., 33, 101Klueh, R. L., 302Knauer, J. P., 28, 33Knight, P. J., 344Kobayashi, M., 335Kobayashi, N., 236, 243Kobayashi, T., 131Kobuchi, T., 17, 77, 95, 110, 115, 118, 157,
182, 221, 222Koch, J. A., 267Koch, R., 243, 335Kochin, V. A., 138Kodaira, J., 17, 77Kodama, K., 13Kodama, R., 26, 254, 266, 271Kodera, F., 181Koenies, A., 103Koenig, R., 33, 192Koguchi, H., 78, 191Kohagura, J., 132, 206Kohyama, A., 301, 302Koidan, V. S., 135Koide, Y., 13, 83, 92, 99, 198, 216Koiwa, M., 13Koizumi, K., 49, 246, 251, 311Koizumi, N., 48Kojima, M., 94Kokusen, S., 13Kolbasov, B., 245Kolesnichenko, Ya. I., 54, 419, 420Kolman, B., 171Komori, A., 17, 77, 95, 110, 115, 118, 157,
182, 221, 222Kondo, K., 131Kondo, T., 329Kondoh, S., 290Kondoh, T., 13, 141Konishi, S., 241, 348, 443Konno, C., 250Konoshima, S., 13, 83Konovalov, S. V., 236Koponen, J. P., 33Korhonen, R., 349Kornev, V. A., 214, 233Korolev, V., 263Korotkov, A. A., 144Korzhavin, V. M., 319Koslowski, H. R., 82, 84Kostsov, Yu. A., 125Kotschenreuther, M., 289, 355Kovaltsova, E., 184Kovan, I. A., 328
Kovrov, P. E., 138Koyama, T., 281Kozaki, Y., 321, 322Kramer, G. J., 13, 119, 141Kramer-Flecken, A., 82Krasheninnikov, S. I., 193, 408Krashennikova, N., 22Krasilnikov, A. V., 17, 110, 115, 238Kreter, A., 82Krieger, K., 188Krikunov, S. V., 125, 214Krilov, S. V., 149, 212Kritz, A. H., 81, 367Krivenski, V., 34, 91Krlin, L., 171Kroegler, H., 68, 91, 209Krommes, J. A., 353Kroupa, F., 171Kruijt, O. G., 154Krupin, V. A., 213Krylov, S. V., 213Krylov, V., 49, 246, 328KSTAR Team, 53, 296KT-5C Team, 40Ku, L. P., 104, 333, 360Ku, S. H., 389Kuang, G. L., 175Kubo, H., 13, 48, 83, 85Kubo, S., 17, 77, 95, 110, 115, 118, 157, 182,
221, 222Kubota, S., 31Kubota, Y., 17, 77Kuehner, G., 33Kugel, H., 31, 70, 111, 127, 316Kuhn, S., 171Kukushkin, A. S., 47, 187, 239, 240Kulcinski, G. L., 320Kulsrud, R. M., 405Kulygin, V. M., 279, 314Kumazawa, R., 17, 77, 95, 110, 115, 118,
157, 182, 221, 222Kuramochi, K., 48Kurbatov, V. I., 325Kurihara, K., 13Kurita, G., 13, 116Kurita, T., 261Kuriyama, M., 13, 109, 141, 312Kurki-Suonio, T., 393Kuroda, T., 247, 261Kuroki, Y., 273Kurzan, B., 58, 76, 440Kusama, Y., 13, 119, 141, 238Kusanagi, N., 13Kuwahara, K., 260
469
Kuyanov, A. Yu., 279Kuzikov, S., 243Kuznetsov, E. A., 125Kuznetsov, S. A., 135Kuzumin, E., 246Kwak, J. G., 297Kwon, M., 296, 298
La Haye, R. J., 60, 75, 81, 107, 126, 146,147, 150, 364
LaBombard, B., 22, 61, 86, 165Labrador, I., 34Lackner, K., 433Ladd, P., 47, 241, 242LaFonteese, D. J., 129Lagniel, J.-M., 329Lako, P., 349Lamarche, P., 31Lang, P. T., 440Lanier, N. E., 41Lao, L. L., 13, 31, 70, 75, 81, 90, 107, 111,
151, 364Lapayese, F., 34Laqua, H., 33, 168Lashmore-Davies, C. N., 426Lasnier, C. J., 60, 75, 81, 126Lavanchy, P., 37, 432Lawson, K. D., 84Laxaback, M., 428Lazarev, V. B., 185Lazarus, E., 75, 104Lazzaro, E., 68, 209, 336, 359Lebedev, S. V., 214, 233, 257LeBlanc, B., 31, 70, 127, 172Leboeuf, J.-N., 400Lechon, Y., 349Ledl, L., 33Lee, G. S., 53, 296Lee, H. G., 298Lee, J. S., 296Lee, K. C., 31Lee, K. W., 297Lee, P., 13Lee, R. W., 267Lee, S., 13, 141Lee, S. G., 298Lee, W., 353, 389Legg, R. A., 107Lehnen, M., 82Lehr, F. M., 286Lei, G. J., 134Leigheb, M., 68, 91, 209, 211, 439Lennholm, M., 166Leonard, A. W., 13, 60, 75, 83, 126, 151, 240Leonov, V. M., 233, 235
Lepicard, S., 349Letterio, J. D., 286Leuterer, F., 17, 58, 66, 76, 227Levin, L. A., 258Levin, L. S., 214Levin, R. G., 125Levinton, F., 31Lewandowski, J., 104, 353Leykin, I. N., 328LHD Group, 337Li, C. K., 28Li, Ding, 388Li, J., 13, 40, 175, 194, 207, 208Li, K. H., 177Li, X. D., 204Liang, Y., 17, 77, 95, 110, 115, 118, 157,
182, 221, 222Libeyre, P., 309Lifschitz, A., 424Likin, K., 34Lin, Y., 22, 61Lin, Z., 104, 351, 353, 371, 389Lin-Liu, Y. R., 75, 107Lindemuth, I., 286Lindl, J., 27, 255, 269Ling, B. L., 207Linhart, J. G., 293Liniers, M., 34Lipschultz, B., 22, 61, 86, 165Lister, J. B., 37, 432Litaudon, X., 166Litvak, A. G., 325Liu, Caigen, 401Liu, J. K., 208Liu, X. P., 334Liu, Y., 40, 131, 134, 177, 204Llobet, X., 37, 432Lloyd, B., 67, 223Loarte, A., 37, 84, 187, 239, 240Lobanov, K., 231LoDestro, L. L., 285, 396Lodi, D., 275Logan, B. G., 27Lohr, J., 75, 107Lok, J., 154Lomas, P. J., 147Lopes Cardozo, N. J., 142, 154, 210Lopez-Bruna, D., 377Lopez-Fraguas, A., 34, 203Lopez-Sanchez, A., 34Lorenz, A., 440Lorenzetto, P., 247Lorenzini, R., 69, 205Lorincik, J., 171
470
Loucks, S. J., 28Luce, T. C., 75, 93, 107, 364Luchetta, A., 69, 205Luckhardt, S., 316Ludwig, G. O., 418Luhmann, N. C., 31Lukash, V., 37, 116, 231Lund, L. D., 28Luo, J. L., 177Luo, J. R., 123, 175, 207, 208Lutsenko, V. V., 54, 420Lynch, V. E., 203, 377, 406Lyon, J. F., 17, 104, 287, 343Lyublinski, I. E., 185, 319
Maassberg, H., 33, 168Macaulay-Newcombe, R. G., 189Maccio, M., 381Mack, A., 242MacKinnon, A., 267Maddaluno, G., 68, 91, 209, 336Maddison, G. P., 84Maegawa, O., 271Maejima, Y., 78, 191Maekawa, F., 250Maekawa, R., 17, 77Maekawa, T., 152Maeno, S., 131Maeyama, M., 191Mafia, G., 91Magne, R., 166Magnin, J.-C., 37, 432Mahajan, S., 380Mahdavi, M. A., 13, 60, 75, 126Maier, H., 188Maingi, R., 31, 60, 70, 111, 126, 127, 316Maisonnier, D., 246, 247, 252Maix, R., 48Majeski, R., 31, 104, 172, 316Makarova, L., 231Makashin, I. N., 153Makhankov, A., 50Makowski, M. A., 75, 81, 90, 107, 151Malaquias, A., 238Malerba, L., 275Malesani, G., 69, 205Malkov, M., 351Malmberg, J.-A., 78, 159Manabe, Y., 131Manahan, R., 48Mancini, R., 265Mandrekas, J., 81, 341Manduchi, G., 69, 205Manfredi, G., 384Manhood, S. J., 67
Manickam, J., 13, 31, 151, 289Manini, A., 37, 432Mank, G., 82, 335Manso, M., 202Mantica, P., 210Mantsinen, M., 144, 147, 393Maqueda, R., 22, 31, 70, 111, 127Maraschek, M., 58, 66, 76, 100, 150, 440Marchenko, V. S., 419Marchetto, C., 362Marchiori, G., 69, 205Marian, J., 275Marinucci, M., 68, 91, 209, 336Markin, A., 50Marletaz, B., 37, 432Marmar, E. S., 22, 61, 165, 409Marmillod, P., 37, 432Marrelli, L., 69, 205Marsala, R., 31Marshall, F. J., 28Martel, P., 275Martin, E., 47, 252Martin, G., 133Martin, P., 69, 205Martın, R., 34Martin, Y. R., 37, 224, 233, 432Martines, E., 69, 205Martinez, A., 48Martınez-Laso, L., 34Martınez-Val, J. M., 275Martini, S., 69, 205Martone, M., 329Martovetsky, N., 48Martynov, A., 37, 147, 155, 359, 432Marushchenko, N. B., 168Masaki, K., 13Masamune, S., 156Maschio, A., 69, 205Mase, A., 179Mashiko, T., 130Masiello, A., 69, 205Maslakowski, J., 246Mast Team, 30Masuda, K., 281Masui, H., 13Masuzaki, S., 17, 77, 95, 110, 115, 118, 157,
182, 221Masuzaki, T., 222Matsuda, S., 329Matsuda, T., 13Matsuda, Y., 291Matsuhiro, K., 317Matsui, H., 261, 329Matsui, K., 48
471
Matsukawa, M., 13, 48Matsumoto, H., 46, 235Matsumoto, T., 13, 99, 361Matsumoto, Y., 251, 260Matsuo, S., 270Matsuoka, C., 272Matsuoka, K., 17, 77, 94, 95, 110, 292Matsuoka, M., 271Matsushima, I., 260Matsuyama, T., 102Matsuzaki, T., 291Matthews, G., 84, 86, 239, 240Mattioli, M., 211Mattor, N., 285Matveev, V. V., 213Mau, T. K., 31, 172Mauel, M. E., 161, 408Maurer, D., 161May, M., 211Mayor, J.-M., 37, 432Mazul, I. V., 50Mazurenko, A., 22, 61, 165Mazzitelli, G., 68, 91, 209, 336Mazzoli, E., 184Mazzucato, E., 31, 70McCarthy, K. J., 34McCarthy, P., 76, 233McClements, K. G., 67, 144, 183McColl, D. R., 195McCollam, K. J., 180Mccormack, B., 31McCormick, K., 33, 192McCracken, G. M., 86McCrory, R. L., 28McCullough, W., 286McDonald, D. C., 59, 233McKee, G. R., 81, 84, 90, 93, 126, 217McKenty, P. W., 28McLean, H. S., 178, 285Meade, D. M., 341Medina, F., 34Medley, S., 31, 183Medrano, M., 34Mehed’kin, A. A., 314Meier, W. R., 27, 320Meigs, A., 84Meister, H., 58, 76, 202, 228Mekler, K. I., 135Melnikov, A., 212Menard, J., 31, 70, 111, 127, 172, 289, 316Mendez, P., 34Mendonca, J. T., 268Menhart, S., 228Menon, M. M., 31
Merkel, P., 103, 104, 292Merola, M., 50, 247Mertens, V., 440Meskat, J., 66Messiaen, A., 81, 82, 84, 400Meulenbroeks, R., 154Meyer, H., 67, 223Meyer-ter-Vehn, J., 265Meyerhofer, D. D., 28Mialton, T. B., 213Michael, C. A., 124Michael, P., 48Micozzi, P., 68, 91, 209Mikhailov, M., 104Mikhailovskii, A. B., 236Mikkelsen, D. R., 13, 22, 104, 216, 355Miklos, M., 409Milani, F., 69, 84, 205Miller, K. E., 112Miller, R., 289Milora, S., 448Milroy, R. D., 112Mima, K., 26, 254, 262, 266, 270–272, 276,
157, 182, 222Moriyama, S., 13, 141, 167, 315Morris, A. W., 67, 223Morse, S. F. B., 28Moses, E., 256Moses, R., 286Moshonas, K., 245Mossessian, D., 22, 61, 86, 165Motojima, O., 17, 77, 95, 110, 115, 118, 157,
182, 221, 222, 310, 337Motoyama, M., 193Moulin, D., 133Moyer, R. A., 81, 93, 126, 217, 370Mueller, D., 31, 70, 111, 127, 172Muller, H. W., 440Mukhin, E. E., 125Mukhovatov, V., 233, 234Mukhovatov, V. S., 236Munsat, T., 316Murai, K., 17, 77
Murakami, M., 26, 75, 81, 84, 90, 151, 200,254, 272
Nakashima, H., 39, 169, 190, 219Nakashima, Y., 132, 206Nakasuga, M., 131Nakatsuka, M., 26, 261, 270, 271, 321Nakayama, T., 117Namkung, W., 297Narihara, K., 17, 77, 95, 110, 115, 118, 157,
182, 221, 222, 337Narushima, Y., 157, 222Naujoks, D., 33Nave, M. F. F., 84, 144Navratil, G. A., 146, 161Nazikian, R., 13, 22, 119, 141Nedoseev, S., 263Neilson, G. H., 104Nelson, B. A., 31, 111, 180, 288Nelson, B. E., 104, 246, 341Nelson, B. N., 287Nelson-Melby, E., 22, 165Nemoto, M., 13Neu, R., 188, 226, 227Neudatchin, S. V., 13, 198Neuhauser, J., 184, 440Neumeyer, C., 31, 341Nevins, W. M., 341, 370, 378Newman, D. E., 377, 383Neyatani, Y., 13, 116, 148, 346Niedermeyer, H., 33Nielsen, P., 69, 205, 238Nightingale, M., 183Nihei, H., 137Niimi, H., 174Nikkola, P., 37, 155, 432Ninomiya, A., 48Ninomiya, H., 13, 234Nishi, M., 241Nishiguchu, A., 276Nishihara, K., 26, 254, 270, 271, 272, 276Nishii, K., 48Nishijima, D., 193Nishijima, G., 48Nishikawa, M., 317Nishikino, M., 271Nishimura, A., 17, 77, 310Nishimura, H., 254, 262, 270, 271Nishimura, K., 17, 77, 95, 110, 115, 118, 157,
182, 221, 222Nomoto, K., 346Nomura, I., 94, 292Noordermeer, E., 154Norimatsu, T., 26, 264, 266, 271, 321, 322Notake, A. T., 110Notake, T., 17, 77, 95, 115, 118, 157, 182,
221, 222Notkin, G., 212Novak, S., 336Novikov, A. Yu., 213Novokhatskii, A. N., 125Novozhilov, S. A., 135Nowak, S., 68, 91, 209Nozato, H., 130NSTX Research Team, 111, 127Nuhrenberg, C., 54, 103, 104, 360Nuhrenberg, J., 103, 292, 397Numata, R., 290Nunoya, Y., 48
O’Brien, M. R., 67, 144O’Connell, R., 41O’Gorman, M., 69, 205Oba, T., 13Obiki, T., 131Obysov, N. A., 328Ochando, M., 34Ochi, Y., 26Odette, G. R., 302Ogando, F., 275Ogawa, H., 174Ogawa, T., 170, 174Ogawa, Y., 137, 290, 348Oh, B. H., 297Ohara, Y., 247Ohdachi, S., 17, 77, 95, 110, 115, 118, 157,
182, 221, 222Ohga, T., 13Ohi, S., 181Ohkubo, K., 17, 77, 95, 110, 118, 222Ohnishi, M., 281Ohnishi, N., 272, 276Ohno, M., 137Ohno, N., 193Ohshima, K., 13Ohta, K., 156Ohtake, I., 17, 77, 95, 310Ohtsu, K., 48Ohtsuka, Y., 317Ohuchi, T., 48Ohyabu, N., 17, 77, 95, 110, 115, 118, 157,
182, 221Ozawa, D., 137Ozeki, T., 13, 99, 119, 141, 148
Paccagnella, R., 69, 205, 362Pacella, D., 68, 91, 209, 211, 439Pacher, G., 233Pacher, H. D., 239Pacios, L., 34Palmer, J., 252Pamela, J., 431Pan, Y. D., 39, 169, 190, 219Panaccione, L., 68, 91, 209, 211, 442Panasenkov, A. A., 314Panella, M., 68, 91, 209Pankin, A., 81, 367Paoletti, F., 31, 70Papastergiou, S., 336Papitto, F., 91Pappas, D., 22Parail, V. V., 201Paris, P. J., 37, 432Park, H., 31Park, Hyoung-Bin, 120Park, W., 119, 405, 407Parker, R. R., 22Parks, P. B., 200Parsells, R., 31Pascal, J. Y., 133Pasqualotto, R., 69, 205Pasternak, A., 283Pastor, I., 34Paul, S., 31, 70, 111, 127, 172Pautasso, G., 162Pavlichenko, R. O., 17, 77, 95, 110, 115, 118,
157, 182, 221, 222Pavlo, P., 171Pavlov, Yu. D., 138, 149, 212, 213Peacock, A., 247Pearson, G., 31Pecquet, A. L., 166Pedersen, T., 22Pedrosa, M. A., 34, 203Peebles, T., 31Peebles, W. A., 90Peeters, A. G., 58, 100, 202, 226Pegoraro, F., 359Pegourie, B., 69Pegurie, B., 205Peng, M., 70, 111Peng, Y.-K. M., 31Penningsfeld, F.-P., 33Pennington, D., 267
475
Pereverzev, G. V., 58, 76, 202, 227Perez, A., 37, 432Pericoli, V., 336Pericoli-Ridolfini, V., 68, 91, 209, 442Perkins, F. W., 42Perkins, L. J., 255Perlado, J. M., 275Peruzzo, S., 69, 205Peterkin, R. E., 286Peterson, B. J., 17, 77, 95, 110, 115, 118,
157, 182, 221, 222Peterson, P. F., 320Petrasso, R. D., 28Petrie, T. W., 13, 60, 81, 126Petrov, V. B., 319Petrov, Yu. V., 125Petrzilka, V., 171Petti, D. A., 320Petty, C. C., 75, 81, 93, 107, 139, 170Peysson, Y., 37, 108, 166, 432, 442Phillips, C. K., 22, 31, 139, 165, 172Phillips, P., 22Piaszczyk, C., 329Piera, M., 275Pieroni, L., 68, 91, 209, 336Pietrzyk, Z. A., 37, 432Piffl, V., 37Pigarov, A., 193Pikuz, S. A., 257Pimanikhin, S., 242Pinches, S. D., 100, 147Pinfold, T., 67Pinsker, R. I., 31, 75, 107, 139, 170, 172Piovan, R., 69, 205Pironti, A., 231, 336Pitcher, C. S., 22, 86Piterskij, V. V., 213Pitts, R. A., 37, 187, 432Pochelon, A., 37, 155, 432Podda, S., 68, 91, 209Podushnikova, K. A., 125Poggianti, A., 252Pogutse, O., 223Polevoi, A., 46, 235Poli, E., 76Poli, F., 91, 439Politzer, P. A., 75Pollaine, S., 269Polman, R. W., 154Polosatkin, S. V., 135Pomaro, N., 69, 205Pomphrey, N., 104, 333Ponce, D., 107Ponno, A., 69, 205
Popovichev, S. V., 213Porcelli, F., 359, 391Porkolab, M., 22, 61, 165, 200Portas, A., 34Porter, G. D., 31, 60, 126, 240, 370, 396Portone, A., 231Pospieszczyk, A., 82Postupesv, V. V., 135Poznyak, V., 212Prager, S. C., 41, 160Prater, R., 75, 90, 107Preti, G., 69, 205Pretzler, G., 265Prieto, J., 275Prokhorov, D. Yu., 153Pronko, S. G. E., 107Proschek, M., 228Prut, V. V., 213Pucella, G., 68, 209Pugno, R., 188Puiatti, M. E., 69, 84, 205Pukhov, A., 265Pulcella, G., 91Punzmann, H., 124Pustovitov, V. D., 236Pustovoit, Yu. M., 314
Qin, H., 353Qin, J., 34Qiu, L. J., 334Quintenz, J., 262
Rachlew, E., 84Radha, P. B., 28Raeder, J., 245Raftopoulos, S., 107Ram, A. K., 31, 426Ramakrishnan, S., 31Raman, R., 31, 111Ramon, R., 180Ramos, J. J., 408, 409Ramponi, G., 68, 209, 336, 359Ran, L. B., 134, 204Rantamaki, K., 393Rao, J., 134, 204Rapp, J., 82, 84Rappaport, H., 289Rathke, J., 329Ravera, G., 91Razumova, K., 445Redd, A. J., 180Redi, M., 104Redi, M. H., 360Regan, S. P., 28Reiersen, W., 104, 333
476
Reiman, A., 103, 104, 360Reimbold, S., 33Reimerdes, H., 37, 155, 432Reinovsky, R. E., 286Reiter, D., 184, 239, 240Renk, T., 262Rensink, M. E., 60, 332, 370, 396Rettig, C. L., 75, 81, 90, 93, 217, 400Rewoldt, G., 13, 95, 353Reyes, S., 275Reznik, S. N., 399Rhodes, T. L., 60, 75, 81, 90, 93, 217, 400Ricci, M., 48Rice, B. W., 75, 81, 364Rice, J., 22, 61, 86, 165, 221Riemann, J., 397Righetti, G. B., 68, 91, 209Righi, E., 225Rikanati, A., 258Rimini, F. G., 62Rix, R., 265Rizzello, C., 300Roberts, S., 28Robinson, J., 31Roccella, M., 336Rodchenkov, B., 247Roderick, N. F., 286Rodgers, E., 247Rodrıguez-Rodrigo, L., 34Rogers, B. N., 355Rogister, A. L., 335, 376, 394Rognlien, T. D., 332, 370, 396Rohde, V., 188Roitershtein, V. S., 214Romanelli, F., 11, 68, 91, 209, 336, 446Romanelli, M., 91, 368, 439Rome, M., 168Romero, J. A., 13Rommers, J. H., 37, 432Roney, P., 31Roquemore, L., 31, 183Rosenberg, A., 31, 172Rosenbluth, M. N., 116, 351Ross, D. R., 93Ross, D. W., 81, 355Rossi, E., 359Rost, J. C., 81, 90Rostagni, G., 69, 205Rothenberg, J., 269Rovenskikh, A. F., 135Rowan, W., 22Rowlands, T. P., 436Roy, I., 212Roy, I. N., 149, 213
Rozhansky, V., 184, 214Rozhdestvensky, V. V., 214Rubbia, C., 8Rubiano, J. G., 275Ruden, E. L., 286Ruhs, N., 33Rumyantsev, E., 125, 231Russo, A., 91Rust, N., 33Rutherford, P., 104, 341Ryan, P., 31, 172Ryter, F., 58, 76, 227, 233Ryutov, D. D., 396
Sa, J. W., 296Saarelma, S., 100, 393Sabbagh, S. A., 31, 70, 111, 127, 172Sadot, O., 258Sadowski, M., 278, 283Saez, R. M., 349Saffert, J., 33Safonova, M. B., 213Sagara, A., 17, 77, 95, 110, 115, 118, 157,
182, 221, 222, 337Saibene, G., 62, 240Saigusa, M., 170Saint-Laurent, F., 133Saito, K., 17, 77, 95, 110, 115, 118, 157, 182,
157, 182, 221, 222Sato, T., 120, 416Sato, Y., 78, 191Satoh, S., 17, 77, 95, 110, 247, 310Satow, S., 222Satow, T., 17, 77, 95, 110, 310Sattin, F., 69, 205Sauter, O., 37, 147, 155, 359, 432Sauthoff, N. R., 341Sauvan, P., 275Savchkov, A. V., 135Savkovic, M., 290Savoldi, L., 48, 327Sawada, K., 276Sawada, T., 437Scarin, P., 69, 205Scavino, E., 37, 432Schade, S., 100Schaffer, M., 31, 111, 126Scheffel, J., 159Schilham, A. M. R., 210Schilling, G., 22, 61, 165Schlegel, T., 265Schleisner, L., 349Schmidt, G. L., 200
Schmidt, J., 104, 341Schmitz, L. W., 129Schnack, D. D., 159Schneider, R., 184, 239, 240, 397Schneider, T., 349Schneider, W., 440Schoenberg, K. F., 286Schoepf, K., 399, 429Schubert, M., 33Schuller, F. C., 154Schultz, J. H., 341Schultz, K. R., 320Schultz, S. D., 426Schunke, B., 186Schweer, B., 82Schweinzer, J., 228Scott, B., 386Scoville, J. T., 146Segre, S. E., 68, 91, 209Segui, J. L., 166Seguin, F., 28Seimiya, M., 13Seka, W., 28Seki, H., 13Seki, M., 13, 167, 315Seki, S., 48Seki, T., 17, 77, 95, 110, 115, 118, 157, 182,
221, 222Sekine, S., 78Semenets, Yu. M., 328Semenov, V. V., 125Semeraro, L., 336Sen, A., 410Sen, S., 195Senda, I., 231, 235, 338Sengoku, S., 303Senju, T., 131Sentoku, Y., 26, 254, 266Sergienko, G., 82Serianni, G., 69, 205Serra, F., 202Sessions, W. D., 200Shamoto, N., 305Shannon, T. E., 329Sharapov, S., 84, 140, 144Sharapov, V. M., 125Shasharina, S., 81Shats, M. G., 124Shelkovenko, T. A., 257Shelukhin, D., 212Shestakov, V., 307, 328Shevchenko, V., 67Shi, M. L., 134Shiba, K., 302
478
Shibaev, S. A., 213Shibano, M., 131Shibanuma, K., 251Shibata, T., 174Shibata, Y., 13Shibayama, N., 290Shidara, H., 131Shigemori, K., 26, 270, 271Shikama, T., 305Shilov, M., 161Shimada, K., 13, 48Shimada, M., 13, 46, 230, 234, 235Shimada, T., 78Shimba, T., 48Shimizu, A., 94, 292Shimizu, K., 13, 83, 85, 311, 395Shimizu, M., 13Shimizu, T., 48Shimomura, Y., 10, 46Shimono, M., 13Shimozuma, T., 17, 77, 95, 110, 115, 118,
157, 182, 221, 222Shinohara, K., 13, 92, 141Shinokhara, K., 119Shinozaki, S., 13Shiraga, H., 26, 270, 271Shirai, H., 13, 83, 92, 198, 216Shirai, Y., 137Shiraiwa, S., 31, 130Shitomi, M., 13Shkolnik, V. S., 328Shlachter, J. S., 286Shoji, C., 222Shoji, M., 17, 77, 95, 110, 115, 118, 157, 182,
221Shoji, T., 231, 338Shoyama, H., 315Shpoliansky, V. N., 319Shumaker, D. E., 378Shumlak, U., 288Shvarts, D., 258Sichta, P., 31Sidorenko, I., 54, 419Sieck, P. E., 180Siemon, R. E., 286Simakov, A. N., 408Simonetto, A., 68, 91, 209Singh, R., 354Sinitsky, S. L., 135Sinman, A., 280Sinman, S., 280Sipila, S., 393Sips, A., 58Sips, A. C. C., 74
Siuko, M., 252Skinner, C., 31, 70, 111, 127Skladnov, K., 247Skovoroda, A. A., 279, 314Skupsky, S., 28Slough, J. T., 112Smalyuk, V. A., 28Smeulders, P., 68, 91, 209Smirnov, A. I., 214Smirnov, A. P., 41Smirnov, P., 139Smirnov, V. A., 314Smirnov, V. P., 263Smith, R. J., 180Smith, S., 244Smolyakov, A., 68, 351Snavely, R., 267Snead, L. L., 301Snipes, J. A., 22, 61, 165, 409Snyder, P. B., 151, 355, 364, 370Soldatov, S., 82, 212Solomon, W. M., 124Sommars, W., 286Sonato, P., 69, 205Song, M., 207Song, X., 13Sorce, C., 28Sorem, M., 256Sotnikov, S. M., 185Sotnikova, G. Yu., 125Soukhanovskii, V., 316Soures, J. M., 28, 256Sovinec, C. R., 285Sozzi, C., 68, 91, 209, 336Spada, E., 69, 205Spaleta, J., 316Spatig, P., 308Speth, E., 33Spineanu, F., 382Spizzo, G., 69, 205Spolaore, M., 69, 205Spong, D. A., 104, 287, 343, 360Srebro, Y., 258St John, H. E., 81Stacey, W. M., 81Stabler, A., 58, 228, 233Staebler, G. M., 81, 84, 90, 93, 126, 200,
367Stallard, B. W., 90, 199Stambaugh, R. D., 342Stammers, K., 67Stamp, M., 84Stangeby, P. G., 126Starace, F., 336
479
Stening, R. J., 436Stephens, R. B., 267Sternberg, A., 299Sternini, E., 68, 209Sternini, S., 91Stevenson, T., 31Stober, J., 58, 227, 233, 240Stoeckel, J., 171Stoeckl, C., 28Stotler, D., 31Strachan, J. D., 84Strait, E. J., 75, 90, 107, 146, 151, 364Strand, P., 81, 104Stratton, B., 31Strauss, H. R., 407Strebkov, Y., 247Strelkov, V., 238Strickler, D., 104, 287, 333, 343Stroth, U., 33Strumberger, E., 54Stutman, D., 31, 70, 127, 172, 316, 407Subbotin, A. A., 149Subbotin, S., 104Sudo, S., 17, 77, 95, 110, 118, 182, 221, 222Sueda, K., 270Sueoka, M., 13Sugama, H., 17, 77, 95Sugawara, A., 13Sugie, T., 13, 83, 238Sugihara, M., 232, 240Sugimoto, M., 48, 329Sugimoto, S., 181Sugisaki, K., 78, 191Sugiyama, L. E., 284, 407Summers, H. P., 228Sunahara, A., 26, 254, 271, 272, 276, 322Sunaoshi, H., 13Sunn Pedersen, T., 61, 408Sushkov, A., 37, 155, 212, 213Suter, L. J., 269Suttrop, W., 58, 66, 227, 232, 233Suzuki, C., 94, 292Suzuki, H., 17, 77, 95, 110, 115, 118, 157,
157, 182, 221, 222, 337Tanaka, K. A., 26, 266Tanaka, S., 132, 206Tang, W. M., 353Tang, X.-Z., 407Taniguchi, M., 313, 318Tarasyan, K., 212Tardini, G., 58, 202Tartoni, N., 68, 91, 209Taruya, K., 281Tassart, J., 438Tataronis, J. A., 171Tatematsu, Y., 132, 206Tatsuno, T., 290, 406, 411Tawara, T., 102Taylor, G., 22, 31, 61, 316, 409, 426Taylor, N., 245Taylor, R. J., 129Taylor, T. S., 75, 81, 151, 364Tazhibayeva, I., 307, 328TCV Team, 173, 187, 224Tebaldi, C., 359Telesca, G., 69, 84, 205Tendler, M., 214Tennfors, E., 428Terakado, M., 13Terakado, T., 13, 48Terranova, D., 69, 205Terry, J. L., 22, 61, 86, 165Terry, P. W., 218, 379Terry, S. D., 178Terumichi, Y., 152Tesini, A., 252Testa, D., 140, 421TEXTOR Team, 82, 218Thomas, D. M., 60, 75, 81, 90, 217Thomas, M. A., 41Thome, R., 48, 341Thomsen, K., 233Thyagaraja, A., 368Tichmann, C., 162Tikhomirov, L. N., 328Tilia, B., 442Tillack, M., 320, 340
Timberlake, J., 316Timofeev, A. V., 279Tivey, R., 47, 50, 252Tobin, S., 112Tobita, K., 13, 109, 141Toda, S., 392Todo, Y., 120Tohyama, Y., 266Toi, K., 17, 77, 94, 95, 110, 115, 118, 157,
182, 221Tomabechi, K., 443Tomiyama, K., 131Tonetti, G., 37, 432Tong, X. D., 208Topilski, L., 245TORE SUPRA Team, 23, 108, 166Torii, Y., 17, 77, 95, 110, 115, 118, 157, 182,
221, 222Toshi, K., 131Tosti, S., 300Totsuka, T., 13Tournianski, M., 67, 183Town, R. P. J., 28Toyokawa, R., 13Tozawa, L. M., 130Tramontin, L., 69, 205Tran, M. Q., 37, 243, 432Tran, T. M., 103, 381Travisanutto, P. E., 91Treutterer, W., 58Tribaldos, V., 34Trukhin, V., 212, 213Trukhina, E., 149, 212Tsakiris, G., 265Tskhakaya, D., 171Tsubakimoto, K., 270Tsuchiya, K., 13, 151, 220, 233, 247Tsuchiya, Y., 48Tsuda, T., 13Tsugita, T., 13Tsuji, H., 48Tsukahara, Y., 13Tsumori, K., 17, 77, 95, 110, 115, 118, 157,
182, 221, 222Tsunematsu, T., 338Tsuneoka, M., 315
481
Tsuru, D., 346Tsuruda, M., 158Tsutsumi, F., 48Tsuzuki, K., 77, 117, 233Tsuzuki, T., 17Tuccillo, A. A., 68, 91, 209, 336Tudisco, O., 68, 91, 209, 336, 359, 439Tukachinsky, A. S., 214Turchi, P. J., 286Turck, B., 309Turman, B., 262Turnbull, A. D., 75, 146, 151, 289, 364Turner, A., 252Tuszweski, M. G., 286Tutt, T., 22Tynan, G., 218, 400Tyunina, M., 299
Uchimoto, E., 41Uckan, N. A., 341Uda, T., 95Ueda, S., 339Ueda, Y., 102, 158, 317Uehara, K., 13, 174Uehara, T., 13Uesugi, Y., 193Ullrich, W., 184Ulrickson, M., 50, 316, 341Umansky, M., 22, 86Umeda, N., 13, 109, 312Unemura, T., 411Uno, Y., 48, 250Unterberg, B., 82, 84Uramoto, Y., 13Urano, H., 13, 83, 216, 232Urushihara, S., 270Ushigusa, K., 13, 109, 167, 339Usui, K., 13Uyama, T., 174, 180Uzlov, V. S., 125
Vaclavik, J., 381, 421Valanju, P., 104, 287, 333, 343Valisa, M., 69, 84, 205Valovic, M., 67, 223, 233van der Meiden, H. J., 154van Milligen, B., 34, 203Van Wassenhove, G., 82Vandenplas, P. E., 400Vasiliev, V., 125, 231Vayakis, G., 235, 238Vdovin, V., 243Vecsey, G., 48Vega, J., 34Velarde, G., 275
Velarde, M., 275Velarde, P., 275Velikhov, E. P., 328Verri, G., 293Versaci, M., 162Vershkov, V., 212, 233Vertkov, A. V., 185, 319Vianello, N., 69, 205Victoria, M., 308Vieider, G., 50Vieira, R., 48Vildjunas, M. I., 125, 214Villard, L., 381, 421Villone, F., 336Viniar, I., 242Violante, V., 300Vitale, V., 68, 91, 209, 336Viterbo, M., 69, 205Vlad, G., 68, 91, 209, 237, 336, 422Vlad, M., 382Volkov, G., 263Volkov, V. V., 138, 149, 213Volpe, F., 33Volponi, F., 290Von Halle, A., 31von Hellermann, M., 84Voskoboynikov, S., 184, 214
W7-AS Team, 101, 168, 192Wada, H., 130Wada, M., 250Wade, M. R., 60, 75, 81, 90, 93, 126, 151,
189, 199Waelbroeck, F., 150, 363Waganer, L. M., 340Wagner, F., 33Wakabayashi, H., 48Wakatani, M., 131, 234, 356, 406, 411, 414Walker, C., 238Walsh, M., 67, 183, 233Waltz, R. E., 81, 90, 93, 367Wampler, W. R., 31, 189Wan, B. N., 175, 208Wan, Y. X., 175Wang, E. Y., 134, 177, 204Wang, L., 40Wang, S., 13, 297Wang, Z., 285Ward, D., 143, 344, 349Ware, A., 287, 343, 360Warner, B., 256Warr, G. B., 124Warrick, C. D., 67Watanabe, K., 313
221, 222Wateri, T., 17Watkins, J. G., 60, 75, 126Watts, C., 22Weber, H. W., 248, 299Wecht, B., 351Weiland, J. G., 354, 380Weisen, H., 37, 232, 432, 441Weitzner, H., 389Weller, A., 33, 101Wen, Y. Z., 40Wendland, C., 33, 168Wenzel, U., 193Werner, A., 33, 101Wesley, J. C., 341, 342West, W. P., 60, 75, 81, 126, 189, 217Westerhof, E., 154Weynants, R. R., 82Weyssow, B., 401White, R., 42, 104, 139, 353, 371, 389, 420Whyte, D. G., 60, 126, 189, 316Widdershoven, H. L. M., 142Wiffen, F. W., 329Wilgen, J., 31, 111, 172Williams, M., 31Williamson, D. E., 343Williamson, E. E., 287Wilson, H. R., 67, 150, 183, 223, 363, 364Wilson, J. R., 22, 31, 139, 172Winter, H. P., 228Witte, K. J., 265Wittman, M. D., 28Wobig, H., 54, 398, 419Wolf, N. S., 60, 126Wolf, R. C., 76, 100, 202Wolfe, S., 22, 61, 165, 409Wong, C. P. C., 189, 342Wong, K. L., 75, 81, 107Woolley, R. W., 341Wu, B., 334Wu, C. H., 50, 323Wu, Y. C., 334Wursching, E., 33Wukitch, S., 22, 61, 165Wurden, G., 31, 70, 285, 286Wysocki, F., 286
Xantopoulos, P., 184Xia, C. Y., 208Xiao, C., 195
Xiao, Z. G., 134Xie, J. K., 40, 175Xu, D. M., 134, 204Xu, G. S., 207Xu, W. B., 204Xu, W. N., 334Xu, X. Q., 31, 370, 396Xu, Y., 221
Yaakobi, B., 28Yagi, K., 290Yagi, M., 392, 412Yagi, Y., 78Yagnov, V. A., 125, 328Yaguchi, K., 131Yagyu, J., 13Yakovenko, Yu. V., 54, 420Yamada, G., 158Yamada, H., 17, 77, 95, 110, 115, 118, 130,
271, 321, 322Yamashita, M., 48Yamashita, Y., 13Yamauchi, K., 17Yamauchi, Y., 303Yamazaki, H., 13Yamazaki, K., 17, 77, 95, 110, 118, 157, 182,
221, 222, 337Yan, J. C., 134Yan, L. W., 134, 177, 204Yanagi, N., 17, 77, 310Yang, H. L., 298Yang, J. G., 31Yang, Y., 175, 208Yao, L., 177, 204
483
Yarim, C., 394Yashiro, H., 260Yasuike, K., 267Yatsu, K., 132, 206Yavorskij, V. A., 399, 429Ybema, J. R., 349Yedvab, Y., 258Yoda, M., 320Yokokura, K., 13Yokota, E., 427Yokoyama, M., 17, 77, 95, 110, 115, 118,
157, 221, 222, 292, 392Yonekawa, I., 13Yonezu, H., 17Yoo, S. J., 298Yoon, B. J., 296Yoon, J. S., 297Yosef-Hai, A., 258Yoshida, H., 13, 241, 254, 266, 270Yoshida, N., 17Yoshida, T., 348, 443Yoshida, Z., 137, 290Yoshikawa, K., 281Yoshikawa, M., 132, 206Yoshimi, T., 251Yoshimura, S., 152, 181Yoshimura, Y., 17, 77, 94, 95, 110, 115, 118,
157, 182, 221, 222Yoshino, R., 13, 92, 116, 167Young, K. M., 341Yu, I. K., 296Yu, Q., 66, 433Yuh, H., 22Yutani, T., 329
Zabeo, L., 69, 205Zabiego, M., 147, 166Zabolotsky, A., 37Zaccaria, P., 69, 205Zacek, F., 171Zahn, G., 48Zaitsev, F. S., 144Zanca, P., 69, 205Zanino, R., 48, 327Zaniol, B., 69, 205Zankl, G., 349Zannelli, L., 336Zanotto, L., 69, 205Zanza, V., 68, 91, 209Zapevalov, V. E., 325Zarnstorff, M. C., 104, 343Zarrabian, M., 188Zastrow, K.-D., 84Zaveriaev, V., 213, 238Zebrowski, J., 278
Zeiler, A., 372Zeng, L., 31, 81, 90Zerbini, M., 68, 91, 209, 211Zhang, Cheng, 423Zhang, N. M., 204Zhang, S. Y., 208Zhang, X. D., 387Zhao, Y., 17, 110, 175, 194, 207, 208Zheng, L.-J., 289, 422Zheng, Y. J., 134Zhil’tsov, V. A., 279, 314Zhogolev, V., 239Zhou, Deng, 423Zhou, Y., 134, 177, 204Zhu, P., 89, 367, 391Zhu, Sizheng, 423Zhu, W., 31, 70Zhubr, N. A., 214Zille, R., 292Zilli, E., 69, 205Zinkle, S., 329Zohm, H., 58, 66Zollino, G., 69, 205Zonca, F., 68, 91, 209, 237, 371, 422Zou, X. L., 166Zubarev, V. F., 314Zuegel, J. D., 28Zurin, M., 263Zurro, B., 34Zushi, H., 39, 169, 190, 219, 238Zvonkov, A. V., 236, 279Zweben, S., 22, 31, 111Zyrmpas, A., 288
