The cover picture shows JT-60, one of the largest operational tokamaks.
By courtesy of the Japan Atomic Energy Research Institute, Tokyo.

PLASMA PHYSICSAND CONTROLLED
NUCLEAR FUSION RESEARCH
1986
VOLUME 1

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©

IAEA, 1987

Permission to reproduce or translate the information contained in this publication may be obtainedby writing to the International Atomic Energy Agency, Wagramerstrasse 5, P.O. Box 100, A-1400 Vienna,
Austria.

Printed by the IAEA in Austria
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NUCLEAR FUSION, SUPPLEMENT 1987

PLASMA PHYSICSAND CONTROLLED
NUCLEAR FUSION RESEARCH
1986
PROCEEDINGS OF THE
ELEVENTH INTERNATIONAL CONFERENCE ON PLASMA PHYSICSAND CONTROLLED NUCLEAR FUSION RESEARCHHELD BY THE
INTERNATIONAL ATOMIC ENERGY AGENCYIN KYOTO, 13-20 NOVEMBER 1986

In three volumes

VOLUME 1

INTERNATIONAL ATOMIC ENERGY AGENCY
VIENNA, 1987

PLASMA PHYSICS AND CONTROLLED NUCLEAR FUSION RESEARCH 1986
IAEA, VIENNA, 1987
STI/PUB/723
ISBN 92-0-130087-5

FOREWORD
The IAEA Conferences on Plasma Physics and Controlled Nuclear Fusion
Research, organized biennially, are the largest and most important meetings in thefield. They are a scientific forum not only for the presentation of the best results
achieved in all laboratories performing fusion research but also for the co-ordinationof international fusion efforts.
The Eleventh Conference, which was organized in co-operation with theJapanese Atomic Energy Research Institute, was attended by about 650 participants
and observers representing 32 countries and two international organizations. About
200 papers were accepted for oral and poster presentations. The Conference opened
with the traditional Artsimovich Memorial Lecture.
The Eleventh Conference became an important event for fusion scientists andengineers all over the world, for two special reasons. First, on the basis of the results
achieved on several experimental devices, the plans for reaching plasma ignition
within the next two years were presented. Second, the Conference not only marked
25 years of international co-operation under the auspices of the IAEA but also
represented the starting point for a new international venture — an international
thermonuclear experimental reactor.
These Proceedings, which include all the technical papers and five Conference
summaries, are published in English as a supplement to the IAEA journal Nuclear
Fusion.
The Agency promotes close international co-operation among plasma and fusion
physicists and engineers of all countries by organizing these regular conferences oncontrolled nuclear fusion and by holding seminars, workshops and specialists
meetings on appropriate topics. It is hoped that the present publication, as part ofthese activities, will contribute to the rapid demonstration of fusion power as one of
the world's future energy resources.

EDITORIAL NOTE
The Proceedings have been edited by the editorial staff of the IAEA to the extent considerednecessary for the reader's assistance. The views expressed remain, however, the responsibility of thenamed authors or participants. In addition, the views are not necessarily those of the governments of
the nominating Member States or of the nominating organizations.
Although great care has been taken to maintain the accuracy of information contained in thispublication, neither the IAEA nor its Member States assume any responsibility for consequences which
may arise from its use.
The use of particular designations of countries or territories does not imply any judgement by the
publisher, the IAEA, as to the legal status of such countries or territories, of their authorities and institutions or of the delimitation of their boundaries.
The mention of names of specific companies or products (whether or not indicated as registered)does not imply any intention to infringe proprietary rights, nor should it be construed as an endorsement
or recommendation on the part of the IAEA.
The authors are responsible for having obtained the necessary permission for the IAEA toreproduce, translate or use material from sources already protected by copyrights.
Material prepared by authors who are in contractual relation with governments is copyrighted
by the IAEA, as publisher, only to the extent permitted by the appropriate national regulations.

CONTENTS OF VOLUME 1
ARTSIMOVICH MEMORIAL LECTURE and TOKAMAK EXPERIMENTS(Session A)
Artsimovich Memorial Lecture (IAEA-CN-47/A-0)
J.F. Clarke
Recent experiments in JT-60 (IAEA-CN-47/A-I-1)
JT-60 Team: T. Abe, H. Aikawa, H. Akaoka, H. Akasaka, M. Akiba,
N. Akino, T. Akiyama, T. Ando, K. Annoh, N. Aoyagi, K. Arakawa,
M. Araki, K. Arimoto, M. Azumi, S. Chiba, M. Dairaku, N. Ebiswa,
T. Fujii, T. Fukuda, H. Furukawa, K. Hamamatsu, K. Hayashi,
M. Hara, K. Haraguchi, H. Hiratsuka, T. Hirayama, S. Hiroki,
K. Hiruta, M. Honda, H. Horiike, R. Hosoda, N. Hosogane, Y. Iida,
T. Iijima, K. Ikeda, Y. Ikeda, T. Imai, T. Inoue, N. Isaji, M. Isaka,
S. Ishida, N. Itige, T. Ito, Y. Ito, A. Kaminaga, M. Kawai,
Y. Kawamata, K. Kawasaki, K. Kikuchi, M. Kikuchi, H. Kimura,
T. Kimura, H. Kishimoto, K. Kitahara, S. Kitamura, A. Kitsunezaki,
K. Kiyono, N. Kobayashi, K. Kodama, Y. Koide, T. Koike,
M. Komata, I. Kondo, S. Konoshima, H. Kubo, S. Kunieda,
S. Kurakata, K. Kurihara, M. Kuriyama, T. Kuroda, M. Kusaka,
Y. Kusama, S. Maehara, K. Maeno, S. Mase, S. Matsuda,
M. Matsukawa, T. Matsukawa, M. Matsuoka, N. Miya, K. Miyati,
Y. Miyo, K. Mizuhashi, M. Mizuno, R. Murai, Y. Murakami, M. Muto,
M. Nagami, A. Nagashima, K. Nagashima, T. Nagashima, S. Nagaya,
H. Nakamura, Y. Nakamura, M. Nemoto, Y. Neyatani, S. Niikura,
H. Ninomiya, T. Nishitani, H. Nomata, K. Obara, N. Ogiwara,
Y. Ohara, K. Ohasa, T. Ohga, H. Ohhara, M. Ohkubo, T. Ohshima,
K. Ohta, M. Ohta, M. Ohtaka, Y. Ohuchi, A. Oikawa, H. Okumura,Y. Okumura, K. Omori, S. Omori, Y. Omori, T. Ozeki, M. Saigusa,
K. Sakamoto, A. Sakasai, S. Sakata, M. Satou, M. Sawahata,
M. Seimiya, M. Seki, S. Seki, K. Shibanuma, R. Shimada, K. Shimizu,
M. Shimizu, Y. Shimomura, S. Shinozaki, H. Shirai, H. Shirakata,
M. Shitomi, K. Suganuma, T. Sugie, T. Sugiyama, H. Sunaoshi,
K. Suzuki, M. Suzuki, M. Suzuki, N. Suzuki, S. Suzuki, Y. Suzuki,
H. Takahashi, M. Takahashi, S. Takahashi, T. Takahashi,
M. Takasaki, M. Takatsu, A. Takeshita, H. Takeuchi, S. Tamura,
S. Tanaka, K. Tani, M. Terakado, T. Terakado, K. Tobita,

3
11

T. Tokutake, T. Totsuka, N. Toyoshima, H. Tsuda, T. Tsugita, S. Tsuji,
Y. Tsukahara, M. Tsuneoka, K. Uehara, M. Umehara, Y. Uramoto,H. Usami, K. Ushigusa, K. Usui, J. Yagyu, K. Yamada, M. Yamamoto,
0. Yamashita, Y. Yamashita, K. Yano, T. Yasukawa, K. Yokokura,
H. Yokomizo, H. Yoshida, M. Yoshikawa, Y. Yoshinari, R. Yoshino,
1. Yonekawa, K. Watanabe
Discussion
JET latest results and future prospects (IAEA-CN-47/A-I-2)
JET Team: H. Altmann, R.J. Anderson, J. Arbez, W. Bailey,
D.V. Bartlett, B. Beaumont, G. Beaumont, K. Behringer, P. Bertoldi,
E. Bertolini, C.H. Best, V. Bhatnagar, R.J. Bickerton, A. Boileau,
F. Bombarda, T. Bonicelli, S. Booth, A. Boschi, G. Bosia, M. Botman,
H. Brelen, H. Brinkschulte, M.L. Browne, M. Brusati, T. Budd,
M. Bures, P. Butcher, H. Buttgereit, D. Cacaut, C. Caldwell-Nichols,
J.D. Callen, D.J. Campbell, J. Carwardine, G. Celentano,
CD. Challis, A. Cheetham, J.P. Christiansen, C. Christodoulopoulos,
P. Chuilon, R. Claesen, J.P. Coad, S.A. Cohen, M. Cooke, J.G. Cordey,
W. Core, S. Corti, A.E. Costley, G.A. Cottrell, M. Cox, C. David,
J. Dean, L. de Kock, E. Deksnis, G.B. Denne, G. Deschamps,
K.J. Dietz, J. Dobbing, S.E. Dorling, D.F. Diichs, G. Duesing,
P. A. Duperrex, H. Duquenoy, A.W. Edwards, J. Ehrenberg,
W. Engelhardt, S.K. Erents, F. Erhorn, B.T. Eriksson, M. Evrard,
H. Falter, N. Foden, M.J. Forrest, C. Froger, K. Fullard,
G. Fussmann, M. Gadeberg, A. Galetsas, A. Gallacher, D. Gambier,R. Giannella, A. Giannelli, A. Gibson, R.D. Gill, A. Goede,
A. Gondhalekar, C. Gordon, C. Gormezano, N.A. Gottardi,
C.W. Gowers, R. Granetz, B. Green, S. Gregoli, F.S. Griph,
R. Haange, J.H. Hamnen, C.J. Hancock, P.J. Harbour, N.C. Hawkes,
P. Haynes, T. Hellsten, J.L. Hemmerich, R. Hemsworth, F. Hendriks,
R.F. Herzog, L. Horton, J. How, M. Huart, A. Hubbard, J. Hugill,
M. Hugon, M. Huguet, B. Ingram, H. Jacket, J. Jacquinot,
Z. Jankowicz, O.N. Jarvis, E.M. Jones, P. Jones, T.T.C. Jones,
C. Jupen, E. Kdllne, J. Kallne, O. Kaneko, A. Kaye, B.E. Keen,
M. Keilhacker, G. Kinahan, S. Kissel, A. Konstantellos, M. Kovanen,
U. Kuphnapfel, P. Kupschus, P. Lallia, J.R. Last, L. Lauro-Taroni,
K.D. Lawson, E. Lazzaro, R.C. Lobel, P.J. Lomas,
N. Lopes-Cardozo, M. Lorenz-Gottardi, C. Lowry, G. Magyar,
D. Maisonnier, M. Malacarne, V. Marchese, P. Massmann,
G.M. McCracken, P. McCullen, M.J. Mead, P. Meriguet, V. Merlo,
V. Mertens, S. Mills, P. Millward, A. Moissonnier, P.L. Mondino,
D. Moreau, P.D. Morgan, E.R. Mutter, D. Murmann, G. Murphy,
M.F. Nave, L. Nickesson, P. Nielsen, P. Noll, S. Nowak, W. Obert,
M. Olsson, J. O'Rourke, M.G. Pacco, J. Paillere, S. Papastergiou,

29
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J. Partridge, D. Pasini, N.J. Peacock, M. Pescatore, J. Plancoulaine,
J.-P. Poffe, R. Prentice, T. Raimondi, J. Ramette, C. Raymond,
P.H. Rebut, J. Removille, W. Riediker, R. Roberts, D.C. Robinson,
A. Rolfe, R. T. Ross, G. Rupprecht, R. Rushton, H. C. Sack,
G. Sadler, J. Sqffert, N. Salmon, F. Sand, A. Santagiustina,
R. Saunders, M. Schmid, F.C. Schuller, K. Selin, R. Shaw, A. Sibley,
D. Sigournay, R. Sillen, F. Simonet, R. Simonini, P. Smeulders,
J.A. Snipes, K. Sonnenberg, L. Sonnerup, M.F. Stamp, C.A. Steed,
D. Stork, P.E. Stott, T.E. Stringer, D.R. Summers, H.P. Summers,
A.J. Tagle, G. Tallents, A. Tonga, A. Taroni, A. Terrington, A. Tesini,
P.R. Thomas, E. Thompson, K. Thomsen, F. Tibone, R. Tivey,
T.N. Todd, P. Trevalion, M. Tschudin, B.J. Tubbing, P. Twynam,
E. Usselmann, H. Van der Beken, J.E. Van Montfoort,
M. von Hellermann, J. von Seggern, T. Wade, C. Walker,
B.A. Wallander, M. Walravens, K. Walter, M.L. Watkins,
M. Watson, D. Webberley, A. Weller, J.A. Wesson, J. Wilks,
T. Winkel, C. Woodward, M. Wykes, D. Young, L. Zannelli, J. W. Zwart
Dicussion
TFTR plasma regimes (IAEA-CN-47/A-I-3)
R.J. Hawryluk, V. Arunasalam, M.G. Bell, M. Bitter, W.R. Blanchard,
N.L. Bretz, R. Budny, C.E. Bush, J.D. Callen, S.A. Cohen,
S.K. Combs, S.L. Davis, D.L. Dimock, H.F. Dylla, P.C. Efthimion,
L.C. Emerson, A.C. England, H.P. Eubank, R.J. Fonck,
E. Fredrickson, H.P. Furth, G. Gammel, R.J. Goldston, B. Grek,
L.R. Grisham, G. Hammett, W.W. Heidbrink, H.W. Hendel,
K.W. Hill, E. Hinnov, S. Hiroe, H. Hsuan, R.A. Hulse, K.P. Jaehnig,
D. Jassby, F.C. Jobes, D.W. Johnson, L.C. Johnson, R. Kaita,
R. Kamperschroer, S.M. Kaye, S.J. Kilpatrick, R.J. Knize, H. Kugel,
P.H. LaMarche, B. LeBlanc, R. Little, C.H. Ma, D.M. Manos,
D.K. Mansfield, R.T. McCann, M.P. McCarthy, D.C. McCune,
K. McGuire, D.H. McNeill, D.M. Meade, S.S. Medley,
D.R. Mikkelsen, S.L. Milora, W. Morris, D. Mueller, V. Mukhovatov,
E.B. Nieschmidt, J. O'Rourke, D.K. Owens, H. Park, N. Pomphrey,
B. Prichard, A.T. Ramsey, M.H. Redi, A.L. Roquemore,
P.H. Rutherford, N.R. Sauthoff, G. Schilling, J. Schivell,
G.L. Schmidt, S.D. Scott, S. Sesnic, J.C. Sinnis, F.J. Stauffer,
B.C. Stratton, G.D. Tait, G. Taylor, J.R. Timberlake, H.H. Towner,
M. Ulrickson, V. Vershkov, S. Von Goeler, F. Wagner, R. Wieland,
J.B. Wilgen, M. Williams, K.L. Wong, S. Yoshikawa, R. Yoshino,
K.M. Young, M.C. Zarnstorff, V.S. Zaveryaev, S.J. Zweben
Discussion

48
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Experimental studies in JET with magnetic separatrix(IAEA-CN-47/A-I-5)
A. Tanga, D.V. Bartlett, K. Behringer, R.J. Bickerton, A. Cheetham,
J.G. Cordey, A. Gibson, N.A. Gottardi, A. Gondhalekar, P.J. Harbour,
H. Jaeckel, M. Keilhacker, E. Lazzaro, V. Martens, P. Noll, S. Nowak,
F.C. Schuller, A.J. Tagle, A. Taroni, P.R. Thomas, F. Tibone
Discussion
Energy confinement and profile consistency in TFTR
(IAEA-CN-47/A-H-1)
R.J. Goldston, V. Arunasalam, M.G. Bell, M. Bitter, W.R. Blanchard,
N.L. Bretz, R- Budny, C.E. Bush, J.D. Callen, S.A. Cohen,
S.K. Combs, S.L. Davis, D.L. Dimock, H.F. Dylla, P.C. Efthimion,
L.C. Emerson, A.C. England, H.P. Eubank, R.J. Fonck, E. Fredrickson,
H.P. Furth, G. Gammel, B. Grek, L.R. Grisham, G. Hammett,
R.J. Hawryluk, W.W. Heidbrink, D.B. Heifetz, H.W. Hendel, K.W. Hill,
E. Hinnov, 5. Hiroe, H. Hsuan, R.A. Hulse, K.P. Jaehnig, D. Jassby,
F.C. Jobes, D.W. Johnson, L.C. Johnson, R. Kaita, R. Kamperschroer,
S.M. Kaye, S.J. Kilpatrick, R.J. Knize, H. Kugel, P.H. LaMarche,
B. LeBlanc, R. Little, C.H. Ma, D.M. Manos, D.K. Mansfield,
R.T. McCann, M.P. McCarthy, D.C. McCune, K. McGuire,
D.H. McNeill, D.M. Meade, 5.5. Medley, D.R. Mikkelsen, S.L. Milora,
W. Morris, D. Mueller, V. Mukhovatov, E.B. Nieschmidt, J. O'Rourke,
D.K. Owens, H. Park, N. Pomphrey, B. Prichard, A.T. Ramsey,
M.H. Redi, A.L. Roquemore, P.H. Rutherford, N.R. Sauthoff,
G. Schilling, J. Schivell, G.L. Schmidt, S.D. Scott, S. Sesnic,
J.C. Sinnis, F.J. Stauffer, B.C. Stratton, G.D. Tait, G. Taylor,
J.R. Timberlake, H.H. Towner, M. Ulrickson, V. Vershkov,
5. Von Goeler, F. Wagner, R. Wieland, J.B. Wilgen, M. Williams,
K.L. Wong, S. Yoshikawa, R. Yoshino, K.M. Young, M.C. Zarnstorff,
V.S. Zaveryaev, S.J. Zweben
Discussion
Energy and particle confinement of JT-60 plasmas with high power
heating (IAEA-CN-47/A-II-2)
JT-60 Team: T. Abe, H. Aikawa, H. Akaoka, H. Akasaka, M. Akiba,
N. Akino, T. Akiyama, T. Ando, K. Annoh, N. Aoyagi, K. Arakawa,
M. Araki, K. Arimoto, M. Azumi, 5. Chiba, M. Dairaku, N. Ebiswa,
T. Fujii, T. Fukuda, H. Furukawa, K. Hamamatsu, K. Hayashi,
M. Hara, K. Haraguchi, H. Hiratsuka, T. Hirayama, S. Hiroki,
K. Hiruta, M. Honda, H. Horiike, R. Hosoda, N. Hosogane, Y. lida,
T. Iijima, K. Ikeda, Y. Ikeda, T. Imai, T. Inoue, N. Isaji, M. Isaka,
5. Ishida, N. Itige, T. Ito, Y. Ito, A. Kaminaga, M. Kawai,
Y. Kawamata, K. Kawasaki, K. Kikuchi, M. Kikuchi, H. Kimura,
T. Kimura, H. Kishimoto, K. Kitahara, 5. Kitamura, A. Kitsunezaki,

65

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75

86
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K. Kiyono, N. Kobayashi, K. Kodama, Y. Koide, T. Koike,
M. Komata, I. Kondo, S. Konoshima, H. Kubo, S. Kunieda,
S. Kurakata, K. Kurihara, M. Kuriyama, T. Kuroda, M. Kusaka,
Y. Kusama, S. Maehara, K. Maeno, S. Mase, S. Matsuda,
M. Matsukawa, T. Matsukawa, M. Matsuoka, N. Miya, K. Miyati,
Y. Miyo, K. Mizuhashi, M. Mizuno, R. Murai, Y. Murakami, M. Muto,
M. Nagami, A. Nagashima, K. Nagashima, T. Nagashima, S. Nagaya,
H. Nakamura, Y. Nakamura, M. Nemoto, Y. Neyatani, S. Niikura,
H. Ninomiya, T. Nishitani, H. Nomata, K. Obara, N. Ogiwara,
Y. Ohara, K. Ohasa, T. Ohga, H. Ohhara, M. Ohkubo, T. Ohshima,
K. Ohta, M. Ohta, M. Ohtaka, Y. Ohuchi, A. Oikawa, H. Okumura,Y. Okumura, K. Omori, S. Omori, Y. Omori, T. Ozeki, M. Saigusa,
K. Sakamoto, A. Sakasai, S. Sakata, M. Satou, M. Sawahata,
M. Seimiya, M. Seki, S. Seki, K. Shibanuma, R. Shimada, K. Shimizu,
M. Shimizu, Y. Shimomura, S. Shinozaki, H. Shirai, H. Shirakata,
M. Shitomi, K. Suganuma, T. Sugie, T. Sugiyama, H. Sunaoshi,
K. Suzuki, M. Suzuki, M. Suzuki, N. Suzuki, S. Suzuki, Y. Suzuki,
H. Takahashi, M. Takahashi, S. Takahashi, T. Takahashi,
M. Takasaki, M. Takatsu, A. Takeshita, H. Takeuchi, S. Tamura,
S. Tanaka, K. Tani, M. Terakado, T. Terakado, K. Tobita,
T. Tokutake, T. Totsuka, N. Toyoshima, H. Tsuda, T. Tsugita, S. Tsuji,
Y. Tsukahara, M. Tsuneoka, K. Uehara, M. Umehara, Y. Uramoto,H. Usami, K. Ushigusa, K. Usui, J. Yagyu, K. Yamada, M. Yamamoto,
0. Yamashita, Y. Yamashita, K. Yano, T. Yasukawa, K. Yokokura,
H. Yokomizo, H. Yoshida, M. Yoshikawa, Y. Yoshinari, R. Yoshino,
1. Yonekawa, K. Watanabe
Discussion
Energy confinement in JET with Ohmic heating and strong
auxiliary heating (IAEA-CN-47/A-II-3)
J.G. Cordey, D.V. Bartlett, V. Bhatnagar, R.J. Bickerton, M. Bures,
J.D. Callen, D.J. Campbell, CD. Challis, J.P. Christiansen, S. Corti,
A.E. Costley, G.A. Cornell, G. Duesing, J. Fessey, M. Gadeberg,
A. Gibson, A. Gondhalekar, N.A. Gottardi, C.W. Gowers,
M. von Hellermann, F. Hendriks, L. Horton, H. Jdckel, J. Jacquinot,
O.N. Jarvis, T.T.C. Jones, E. Kallne, J. Kallne, M. Keilhacker,
S. Kissel, L. de Kock, P. Lallia, E. Lazzaro, P.J. Lomas,
N. Lopes-Cardozo, P.D. Morgan, P. Nielsen, J. O'Rourke,
R. Prentice, R. T. Ross, G. Sadler, F. C. Schuller, A. Stabler,
P. Smeulders, M.F. Stamp, D. Stork, P.E. Stott, D.R. Summers,
A. Tonga, P.R. Thomas, E. Thompson, K. Thomsen, G. Tonetti,
B.J. Tubbing, M.L. Watkins
Discussion

98
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Effect of p and j profiles on energy confinement in ECH experiments
on T-10 (IAEA-CN-47/A-II-4)
V. V. Alikaev, A.A. Bagdasarov, E.L. Berezovskij, A.B. Berlizov,
G.A. Bobrovskij, A.A. Borshchegovskij, N.L. Vasin, A.N.I. Vertiporokh,
V.V. Volkov, E.P. Gorbunov, Yu.A. Gorelov, Yu.V. Esipchuk,
S.L. Efremov, V.S. Zaveryaev, V.I. Win, A.Ya. Kislov, P.E. Kovrov,
S.Yu. Luk'yanov, Yu.S. Maksimov, A.A. Medvedev, G.E. Notkin,
A.B. Pimenov, V.I. Poznyak, I.A. Popov, S.V. Popovichev,
K.A. Razumova, P.V. Savrukhin, E.V. Sitar, A.M. Solntsev,
M.M. Stepanenko, V.S. Strelkov, A.V. Tarakanov, K.N. Tarasyan,
K.B. Titishov, V.M. Trukhin, G.S. Yaramyshev, S.N. Vlasov,
A.L. Gol'denberg, V.A. Markelov, V.A. Flyagin, B.V. Shishkin
Discussion

Ill

123

Achievement of regimes with high density, low qa and good confinement
on ASDEX (IAEA-CN-47/A-II-5)
125
H. Niedermeyer, F. Wagner, G. Becker, K. Bttchl, A. Eberhagen,G. Fussmann, O. Gehre, G. von Gierke, O. Gruber, G. Janeschitz,
M. Kaufrnann, K. Lackner, R. Lang, V. Mertens, H. Murmann,
W. Sandmann, E. Speth, K. Steinmetz, G. Vlases, H.S. Bosch,
H. Brocken, J. Gerhardt, E. Glock, G. Haas, J. Hofrnann,A. Izvozchikov, F. Karger, M. Keilhacker, O. Klttber, M. Kornherr,
M. Lenoci, G. Lisitano, F. Mast, KM. Mayer, K. McCormick,D. Meisel, E.R. Muller, Z.A. Pietrzyk, W. Poschenrieder, H. Rapp,
H. Riedler, H. Rohr, J. Roth, F. Ryter, F. Schneider, C. Setzensack,G. Siller, P. Smeulders, F.X. Soldner, K.-H. Steuer, O. Vollmer, D. Zasche
Discussion
136
Studies of the regime of improved particle and energy confinement
following pellet injection into Alcator C (IAEA-CN-47/A-III-l)
M. Greenwald, M. Besen, F. Camacho, C. Fiore, M. Foord,
R. Gandy, C. Gomez, R. Granetz, D. Gwinn, S. Knowlton,
B. LaBombard, B. Lipschultz, H. Manning, E. Marmar, S.C. McCool,
J. Parker, R. Parker, R. Petrasso, P. Pribyl, J. Rice, D. Sigtnar,
Y. Takase, J. Terry, R. Watterson, S. Wolfe
Discussion
Confinement studies of additionally heated plasma in the
JFT-2M tokamak (IAEA-CN-47/A-III-2)
K. Odajima, A. Funahashi, K. Hoshino, S. Kasai, T. Kawakami,
H. Kawashima, T. Matoba, T. Matsuda, H. Matsumoto, Y. Miura,
M. Mori, H. Ogawa, T. Ogawa, H. Ohtsuka, S. Sengoku, T. Shoji,
N. Suzuki, H. Tamai, Y. Uesugi, T. Yamamoto, T. Yamauchi,
M. Hasegawa, S. Takada, K. Hasegawa, A. Honda, I. Ishibori,T. Kashimura, Y. Kashiwa, M. Kazawa, K. Kikuchi, Y. Matsuzaki,

139

149
151

K. Ohuchi, H. Okano, T. Shibata, T. Shibuya, T. Shiina, K. Suzuki,
T. Tani, K. Yokoyama
Discussion
157
Initial results from the DIII-D tokamak (IAEA-CN-47/A-III-3)
159
J. Luxon, P. Anderson, F. Batty, C. Baxi, G. Bramson, N. Brooks,
B. Brown, B. Burley, K.H. Burrell, R. Callis, G. Campbell, T. Carlstrom,
A. Colleraine, J. Cummings, J. Davis, J. DeBoo, S. Ejima,
R. Evanko, H. Fukumoto, R. Gallix, J. Gilleland, T. Glad, P. Gohil,
A. Gootgeld, R.J. Groebner, S. Hanai, J. Haskovec, E. Heckman,
M. Heiberger, F.J. Helton, P. Henline, D. Hill, D. Hoffman,
E. Hoffmann, R. Hong, N. Hosogane, C. Hsieh, G.L. Jackson, G. Jahns,
G. Janeschitz, E. Johnson, A. Kellman, J.S. Kim, J. Kohli, A. Langhorn,
L. Lao, P. Lee, S. Lightner, J. Lohr, M. Mahdavi, M. Mayberry,
B. McHarg, T. McKelvey, R. Miller, C.P. Moeller, D. Moore, A. Nerem,
P. Noll, T. Ohkawa, N. Ohyabu, T. Osborne, D. Overskei, P. Petersen,
T. Petrie, J. Phillips, R. Prater, J. Rawls, E. Reis, D. Remsen,
P. Riedy, P. Rock, K. Schaubel, D. Schissel, J. Scoville, R. Seraydarian,
M. Shimada, T. Shoji, B. Sleaford, J. Smith, Jr., P. Smith, T. Smith,
R.T. Snider, R.D. Stambaugh, R. Stav, H. St. John, R. Stockdale,
E.J. Strait, R. Street, T.S. Taylor, J. Tooker, M. Tupper, S.K. Wong,
S. Yamaguchi
Discussion
168
Pellet injection results during TFTR Ohmic and neutral beam heating
experiments (IAEA-CN-47/A-III-4)
171
G.L. Schmidt, S.L. Milora, V. Arunasalam, M.G. Bell, M. Bitter,
C.E. Bush, S.K. Combs, H.F. Dylla, P.C. Efthimion, E. Fredrickson,
R.J. Goldston, B. Grek, L.R. Grisham, G. Hammett, R.J. Hawryluk,
W.W. Heidbrink, H.W. Hendel, K.W. Hill, H. Hsuan, D.W. Johnson,
L.C. Johnson, P.H. LaMarche, R. Little, D.K. Mansfield, D.C. McCune,
K. McGuire, D.M. Meade, S.S. Medley, D. Mueller, W. Morris,
E.B. Nieschmidt, J. O'Rourke, D.K. Owens, H. Park, A.T. Ramsey,
M.H. Redi, J. Schivell, D. Schuresko, T. Senko, S. Sesnic,
F.J. Stauffer, B.C. Stratton, G.D. Tait, G. Taylor, J.R. Timberlake,
H.H. Towner, M. Ulrickson, R. Wieland, J.B. Wilgen, M. Williams,
M.C. Zarnstorff, S.J. Zweben
Discussion
177
Fast ion losses and plasma confinement in TFR during neutralbeam injection (IAEA-CN-47/A-III-5)
179
Equipe TFR
Discussion
185
Particle and thermal transport, and resonant field experiments in TEXT(IAEA-CN-47/A-III-6)
187
A.J. Wootton, R.D. Bengtson, J.A. Boedo, R.V. Bravenec, K.L. Empson,

K.W. Gentle, W.L. Hodge, C.C. Klepper, J. Porter, A. Macmahon,
S.C. McCool, W.H. Miner, D.M. Patterson, P.E. Phillips, B. Richards,
C.P. Ritz, T.L Rhodes, D.W. Ross, W.L. Rowan, J.C. Wiley,
D.L. Brower, S.K. Kim, N.C. Luhmann, W.A. Peebles,
J.S. DeGrassie, T.E. Evans, G.L. Jackson, N. Ohyabu, P.H. Diamond,
T.S. Hahm, P.W. Terry, Shaping Fan, Yexi He, J.C. Forster,
R.L. Hickok, P.M. Schoch, G. Haas, F. Karger, Xuehua Yu,
Shaobai Zheng, R. C. Isler
Discussion

194

Impurity production mechanisms and behaviour during additional heatingin JET (IAEA-CN-47/A-IV-l)
197
K. Behringer, A. Boileau, F. Bombarda, G.B. Denne, W. Engelhardt,
M.J. Forrest, G. Fussmann, R. Giannella, N.A. Gottardi,
M. von Hellermann, L. Horton, H. Jacket, C. Jupen, E. Kdllne,
K.D. Lawson, G. Magyar, G.M. McCracken, P.D. Morgan, E.R. Miiller,
N.J. Peacock, J. Ramette, B. Saoutic, M.F. Stamp, H.P. Summers,
G. Tallents, A. Weller
Discussion
205
Impurity and particle transport and control in TFTR
(IAEA-CN-47/A-IV-2)
207
K.W. Hill, V. Arunasalam, M.G. Bell, M. Bitter, W.R. Blanchard,
N.L Bretz, R. Budny, C.E. Bush, S.A. Cohen, S.K. Combs,
S.L. Davis, D.L. Dimock, H.F. Dylla, P.C. Efthimion, L.C. Emerson,
A.C. England, H.P. Eubank, R.J. Fonck, E. Fredrickson, H.P. Furth,
G. Gammel, R.J. Goldston, B. Grek, L.R. Grisham, G. Hamrnett,
R.J. Hawryluk, W. W. Heidbrink, H. W. Hendel, E. Hinnov, S. Hiroe,
H. Hsuan, R.A. Hulse, K.P. Jaehnig, D. Jassby, F.C. Jobes,
D.W. Johnson, L.C. Johnson, R. Kaita, R. Kamperschroer, S.M. Kaye,
S.J. Kilpatrick, R.J. Knize, H. Kugel, P.H. LaMarche, B. LeBlanc,
C.H. Ma, D.M. Manos, D.K. Mansfield, R.T. McCann, M.P. McCarthy,
D.C. McCune, K. McGuire, D.H. McNeill, D.M. Meade, S.S. Medley,
D.R. Mikkelsen, S.L. Milora, W. Morris, D. Mueller, V. Mukhovatov,
E.B. Nieschmidt, J. O'Rourke, D.K. Owens, H. Park, N. Pomphrey,
B. Prichard, A.T. Ramsey, M.H. Redi, A.L. Roquemore,
P.H. Rutherford, N.R. Sauthoff, G. Schilling, J. Schivell,
G.L. Schmidt, S.D. Scott, S. Sesnic, M. Shimada, J.C. Sinnis,
F.J. Stauffer, J. Strachan, B.C. Stratton, G.D. Tait, G. Taylor,
J.R. Timberlake, H.H. Towner, M. Ulrickson, V. Vershkov,
S. Von Goeler, F. Wagner, R. Wieland, J.B. Wilgen, M. Williams,
K.L. Wong, S. Yoshikawa, R. Yoshino, K.M. Young, M.C. Zarnstorff,
V.S. Zaveryaev, S.J. Zweben
Discussion
216

Impurity and MHD behaviour in JT-60 divertor discharges(IAEA-CN-47/A-IV-3)
JT-60 Team: T. Abe, H. Aikawa, H. Akaoka, H. Akasaka, M. Akiba,
N. Akino, T. Akiyama, T. Ando, K. Annoh, N. Aoyagi, K. Arakawa,
M. Araki, K. Arimoto, M. Azumi, S. Chiba, M. Dairaku, N. Ebiswa,
T. Fujii, T. Fukuda, H. Furukawa, K. Hamamatsu, K. Hayashi,
M. Hara, K. Haraguchi, H. Hiratsuka, T. Hirayama, S. Hiroki,
K. Hiruta, M. Honda, H. Horiike, R. Hosoda, N. Hosogane, Y. Iida,
T. Iijima, K. Ikeda, Y. Ikeda, T. Imai, T. Inoue, N. Isaji, M. Isaka,
S. Ishida, N. Itige, T. Ito, Y. Ito, A. Kaminaga, M. Kawai,
Y. Kawamata, K. Kawasaki, K. Kikuchi, M. Kikuchi, H. Kimura,
T. Kimura, H. Kishimoto, K. Kitahara, S. Kitamura, A. Kitsunezaki,
K. Kiyono, N. Kobayashi, K. Kodama, Y. Koide, T. Koike,
M. Komata, I. Kondo, S. Konoshima, H. Kubo, S. Kunieda,
S. Kurakata, K. Kurihara, M. Kuriyama, T. Kuroda, M. Kusaka,
Y. Kusama, S. Maehara, K. Maeno, S. Mase, S. Matsuda,
M. Matsukawa, T. Matsukawa, M. Matsuoka, N. Miya, K. Miyati,
Y. Miyo, K. Mizuhashi, M. Mizuno, R. Murai, Y. Murakami, M. Muto,
M. Nagami, A. Nagashima, K. Nagashima, T. Nagashima, S. Nagaya,
H. Nakamura, Y. Nakamura, M. Nemoto, Y. Neyatani, S. Niikura,
H. Ninomiya, T. Nishitani, H. Nomata, K. Obara, N. Ogiwara,
Y. Ohara, K. Ohasa, T. Ohga, H. Ohhara, M. Ohkubo, T. Ohshima,
K. Ohta, M. Ohta, M. Ohtaka, Y. Ohuchi, A. Oikawa, H. Okumura,Y. Okumura, K. Omori, S. Omori, Y. Omori, T. Ozeki, M. Saigusa,
K. Sakamoto, A. Sakasai, S. Sakata, M. Satou, M. Sawahata,
M. Seimiya, M. Seki, S. Seki, K. Shibanuma, R. Shimada, K. Shimizu,
M. Shimizu, Y. Shimomura, S. Shinozaki, H. Shirai, H. Shirakata,
M. Shitomi, K. Suganuma, T. Sugie, T. Sugiyama, H. Sunaoshi,
K. Suzuki, M. Suzuki, M. Suzuki, N. Suzuki, S. Suzuki, Y. Suzuki,
H. Takahashi, M. Takahashi, S. Takahashi, T. Takahashi,
M. Takasaki, M. Takatsu, A. Takeshita, H. Takeuchi, S. Tamura,
S. Tanaka, K. Tani, M. Terakado, T. Terakado, K. Tobita,
T. Tokutake, T. Totsuka, N. Toyoshima, H. Tsuda, T. Tsugita, S. Tsuji,
Y. Tsukahara, M. Tsuneoka, K. Uehara, M. Umehara, Y. Vramoto,H. Usami, K. Ushigusa, K. Usui, J. Yagyu, K. Yamada, M. Yamamoto,
0. Yamashita, Y. Yamashita, K. Yano, T. Yasukawa, K. Yokokura,
H. Yokomizo, H. Yoshida, M. Yoshikawa, Y. Yoshinari, R. Yoshino,
1. Yonekawa, K. Watanabe
Discussion
Density limit and impurity transport investigations in
DITE tokamak (IAEA-CN-47/A-IV-4)
J. Allen, G.E. Austin, K.B. Axon, R. Barnsley, M. Dunstan,
D.N. Edwards, K.D. Evans, S.J. Fielding, D.H.J. Goodall,

217

224
227

W. Han, N.C. Hawkes, J. Hugill, P.C. Johnson,
G.M. McCracken, G.F. Matthews, W. Millar, N.J. Peacock,
C.S. Pitcher, J. Pritchard, A. Sykes, M.F. Turner, J.P. Webber
Discussion
Studies of edge plasma phenomena in T-10
(IAEA-CNW/A-IV-5)
K.F. Alexander, L. Dietrich, H. Grote, K. Gunther,J. Lingertat, E.L. Berezovskij, A. V. Chankin, V.M. Chicherov,
S.E. Lysenko, V.A. Medvedev, A.Yu. Pigarov, V.I. Pistunovich,S. V. Popovichev, V. Vershkov
Discussion
ALT-I pump limiter behaviour and edge plasma flows during biasing and
ICRF heating in the TEXTOR tokamak (IAEA-CN-47/A-IV-6)
R.W. Conn, D.M. Goebel, W.K. Leung, G.A. Campbell, B. LaBombard,
A.K. Prinja, W.L. Hsu, G.J. Thomas, K.H. Dippel, K.H. Finken,
R. Reiter, G.H. Wolf, H.L. Bay, G. Bertschinger, W. Bieger, P. Bogen,
W. Brussau, Y. Cao, H.G. Esser, G. Fuchs, B. Giesen, E. Graffinann,
H. Hartwig, E. Hintz, F. Hoenen, K. Hothker, A. Kaleck, H. Kever,L. Kdnen, M. Korten, L. Li, Y.T. Lie, A. Pospieszczyk, A. Rogister,G. Ross, D. Rusbttldt, U. Samm, J. Schluter, B. Schweer, H. Soltwisch,
F. Waelbroeck, G. Waidmann, P. Wienhold, J. Winter, R. Yamauchi,
T. Delvigne, P. Descamps, F. Durodie, M. Jadoul, R. Koch,
D. Lebeau, A.M. Messiaen, D.I.C. Pearson, P.E. Vandenplas,
R. Van Nieuwenhove, G. Van Oost, G. Van Wassenhove, R.R. Weynants
Discussion
Current density profiles in the TEXTOR tokamak (IAEA-CN-47/A-V-1) ...
H. Soltwisch, W. Stodiek, J. Manickam, J. Schluter
Discussion
Stability and confinement studies in the Princeton Beta Experiment (PBX)(IAEA-CN47/A-V-2)
M. Okabayashi, K. Bol, M. Chance, P. Couture, H. Fishman,
R.J. Fonck, G. Gammel, W.W. Heidbrink, K. Ida, K.P. Jaehnig,
G. Johns, R. Kaita, S.M. Kaye, H. Kugel, B. LeBlanc, J. Manickam,
W. Morris, G.A. Navratil, N. Ohyabu, S. Paul, E. Powell, M. Reusch,
S. Sesnic, H. Takahashi
Discussion
Stabilization of the tearing mode (IAEA-CN-47/A-V-3)
T.C. Hender, D.C. Robinson, J.A. Snipes
Discussion
Experimental and theoretical studies of circular cross-section high beta
tokamaks (IAEA-CN-47/A-V-4)
G.A. Navratil, Y. Baransky, A. Bhattacharjee, C.K. Chu, A.V. Deniz,
A.A. Grossman, A. Holland, T. Ivers, X.L. Li, T.C. Marshall,

235
237

246
249

261
263
272
275

290
291
298
299

M.E. Mauel, S. Sabbagh, A.K. Sen, J. Van Dam, X.-H. Wang,
M. Hughes, M. Phillips, A.M.M. Todd
Discussion
Disruption control experiments using local modular
multipole field coils (IAEA-CN-47/A-V-5)
K. Yamazaid, K. Kawahata, Y. Abe, R. Akiyama, T. Amano, R. Ando,
Y. Hamada, S. Hirokura, K. Ida, E. Kako, O. Kaneko, Y. Kawasumi,
S. Kitagawa, T. Kuroda, K. Masai, K. Matsuoka, A. Mohri,
M. Mugishima, N. Noda, I. Ogawa, Y. Ogawa, K. Ohkubo, Y. Oka,
K. Sakurai, M. Sasao, K. Sato, K.N. Sato, S. Tanahashi, Y. Taniguchi,
K. Toi, T. Watari
Discussion
Studies of MHD modes and high frequency fluctuations on the
HT-6B and HT-6M tokamaks (IAEA-CN-47/A-V-6)
Jikang Xie, Jiayu Chen, Wenkang Guo, Yuping Huo, HT-6B Group,
HT-6M Group
Anomalous transport in JET plasmas (IAEA-CN-47/A-VI-1-1)
D.F. Diichs, T.E. Stringer, A. Taroni, M. Brusati, N.A. Gottardi,
T. Hells ten, F. Tibone
Microinstability based models for confinement properties and ignition
criteria in tokamaks (IAEA-CN-47/A-VI-1-2)
W.M. Tang, CM. Bishop, B. Coppi, S.M. Kaye, F.W. Perkins,
M.H. Redi, G. Rewoldt
Drift wave models of tokamak confinement (IAEA-CN-47/A-VI-1-3)
R.E. Waltz, R.R. Dominguez, S.K. Wong, P.H. Diamond, G.S. Lee,
T.S. Hahm, N. Mattor
Discussion on papers IAEA-CN-47/A-VI-1 -1 to A-VI-1-3
MHD stability and transport of beam heated ASDEX discharges
in the vicinity of the beta limit (IAEA-CN-47/A-VI-2)
O. Gruber, G. Becker, G. von Gierke, K. Grassie, O. Kliiber,
M. Kornherr, K. Lackner, J.K. Lee, H.P. Zehrfeld, H.S. Bosch,
H. Brocken, A. Eberhagen, G. Fussmann, O. Gehre, J. Gernhardt,
E. Glock, G. Haas, J. Hofinann, A. Izvozchikov, G. Janeschitz,
F. Karger, M. Kaufinann, M. Keilhacker, M. Lenoci, G. Lisitano,
F. Mast, H.M. Mayer, K. McCormick, D. Meisel, V. Mertens,
E.R. Miiller, H. Murmann, H. Niedermeyer, W. Poschenrieder,
H. Rapp, H. Riedler, H. Rdhr, J. Roth, F. Ryter, F. Schneider,
C. Setzensack, G. Siller, P. Smeulders, F.X. Soldner, E. Speth,
K.-H. Steuer, O. Vollmer, F, Wagner, D. Zasche
Discussion
Beta limit disruptions (IAEA-CN-47/A-VI-3)
Mingsheng Chu, E.J. Strait, J.K. Lee, L. Lao, R.D. Stambaugh,
K.H. Burrell, R.J. Groebner, J.M. Greene, F.J. Helton, J.Y. Hsu,

308
309

316
317

325

337

345

355
357

371
373

J.S. Kim, H. St. John, R.T. Snider, T.S. Taylor
Discussion
Sawteeth and temperature profiles in tokamaks (IAEA-CN-47/A-VI-4)
D.A. Boyd, R.E. Denton, J.Q. Dong, J.F. Drake, P.N. Guzdar,
A.B. Hassam, R.G. Kleva, Y.C. Lee, C.S. Liu, F.J. Stauffer
Discussion

384
387

398

First results from the HL-1 tokamak (IAEA-CN-47/A-VII-l)
399
HL-1 Tokamak Group
Tokamak current drive via AC magnetic helicity injection(IAEA-CN-47/A-VII-2)
407
P.M. Bellan, P.C. Liewer, J.M. Finn, T.M. Antonsen, Jr.
Study of enhanced confinement and q = 2 disruptive modes with fastrising current in the STOR-1M tokamak (IAEA-CN-47/A-VII-3)
413
H. Kuwahara, O. Mitarai, E.J.M. Van Heesch, M. Emaami,
A.H. Sarkissian, S. W. Wolfe, A. Hirose, KM. Skarsgard
Coherent and turbulent fluctuations in TFTR (IAEA-CN-47/A-VII-4)
421
K. McGuire, V. Arunasalam, M.G. Bell, M. Bitter, W.R. Blanchard,
N.L. Bretz, R. Budny, C.E. Bush, J.D. Callen, M. Chance, S.A. Cohen,
S.K. Combs, S.L. Davis, D.L. Dimock, H.F. Dylla, P.C. Efthimion,
L.C. Emerson, A.C. England, H.P. Eubank, R.J. Fonck, E. Fredrickson,
H.P. Furth, G. Gammel, R.J. Goldston, B. Grek, L.R. Grisham, G. Hammett,
R.J. Hawryluk, W.W. Heidbrink, H.W. Hendel, K.W. Hill, E. Hinnov,
S. Hiroe, H. Hsuan, R.A. Hulse, K.P. Jaehnig, D. Jassby, F.C. Jobes,
D.W. Johnson, L.C. Johnson, R. Kaita, R. Kamperschroer, S.M. Kaye,
S.J. Kilpatrick, R.J. Knize, H. Kugel, P.H. LaMarche, B. LeBlanc,
R. Little, C.H. Ma, J. Manickam, D.M. Manos, D.K. Mansfield,
R.T. McCann, M.P. McCarthy, D.C. McCune, D.H. McNeill,
D.M. Meade, S.S. Medley, D.R. Mikkelsen, S.L Milora, D. Monticello,
W. Morris, D. Mueller, E.B. Nieschmidt, J. O'Rourke, D.K. Owens,
H. Park, W. Park, N. Pomphrey, B. Prichard, A.T. Ramsey,
M.H. Redi, A.L. Roquemore, P.H. Rutherford, N.R. Sauthoff,
G. Schilling, J. Schivell, G.L. Schmidt, S.D. Scott, S. Sesnic,
J.C. Sinnis, F.J. Stauffer, B.C. Stratton, G.D. Tail, G. Taylor,
J.R. Timberlake, H.H. Towner, M. Ulrickson, S. Von Goeler,
F. Wagner, R. Wieland, J.B. Wilgen, M. Williams, K.L. Wong,
S. Yoshikawa, KM. Young, M.C. Zarnstorff, S.J. Zweben
Sawteeth and disruptions in JET (IAEA-CN^7/A-VII-5)
433
D.J. Campbell, P. A. Duperrex, A.W. Edwards, R.D. Gill,
C.W. Gowers, R. Granetz, M. Hugon, P.J. Lomas, N. Lopes-Cardozo,
M. Malacarne, M.F. Nave, D.C. Robinson, F.C. Schuller,
P. Smeulders, J.A. Snipes, P.E. Stott, G. Tonetti, B.J. Tubbing,
A. Welter, J.A. Wesson

PLASMA HEATING AND CURRENT DRIVE(Session F)
Radiofrequency heating on JET (IAEA-CN-47/F-I-1)
449
JET Team: H. Altmann, R.J. Anderson, J. Arbez, W. Bailey,
D.V. Bartlett, B. Beaumont, G. Beaumont, K. Behringer, P. Bertoldi,
E. Bertolini, C.H. Best, V. Bhatnagar, R.J. Bickerton, A. Boileau,
F. Bombarda, T. Bonicelli, S. Booth, A. Boschi, G. Bosia, M. Botman,
H. Brelen, H. Brinkschulte, M.L. Browne, M. Brusati, T. Budd,
M. Bures, P. Butcher, H. Buttgereit, D. Cacaut, C. Caldwell-Nichols,
J.D. Callen, D.J. Campbell, J. Carwardine, G. Celentano,
CD. Challis, A. Cheetham, J.P. Christiansen, C. Christodoulopoulos,
P. Chuilon, R. Claesen, J.P. Coad, S.A. Cohen, M. Cooke, J.G. Cordey,
W. Core, S. Corti, A.E. Costley, G.A. Cornell, M. Cox, C. David,
J. Dean, L. de Kock, E. Deksnis, G.B. Denne, G. Deschamps,
K.J. Dietz, J. Dobbing, S.E. Dorling, D.F. Duchs, G. Duesing,
P.A. Duperrex, H. Duquenoy, A.W. Edwards, J. Ehrenberg,
W. Engelhardt, S.K. Erents, F. Erhorn, B.T. Eriksson, M. Evrard,
H. Falter, N. Foden, M.J. Forrest, C. Froger, K. Fullard,
G. Fussmann, M. Gadeberg, A. Galetsas, A. Gallacher, D. Gambier,R. Giannella, A. Giannelli, A. Gibson, R.D. Gill, A. Goede,
A. Gondhalekar, C. Gordon, C. Gormezano, N.A. Gottardi,
C.W. Gowers, R. Granetz, B. Green, S. Gregoli, F.S. Griph,
R. Haange, J.H. Hamnen, C.J. Hancock, P.J. Harbour, N.C. Hawkes,
P. Haynes, T. Hellsten, J.L. Hemmerich, R. Hemsworth, F. Hendriks,
R.F. Herzog, L. Horton, J. How, M. Huart, A. Hubbard, J. Hugill,
M. Hugon, M. Huguet, B. Ingram, H. Jackel, J. Jacquinot,
Z. Jankowicz, O.N. Jarvis, E.M. Jones, P. Jones, T.T.C. Jones,
C. Jupen, E. Kdllne, J. Kallne, O. Kaneko, A. Kaye, B.E. Keen,
M. Keilhacker, G. Kinahan, S. Kissel, A. Konstantellos, M. Kovanen,
U. Kuphnapfel, P. Kupschus, P. Lallia, J.R. Last, L. Lauro-Taroni,
K.D. Lawson, E. Lazzaro, R.C. Lobel, P.J. Lomas,
N. Lopes-Cardozo, M. Lorenz-Gottardi, C. Lowry, G. Magyar,
D. Maisonnier, M. Malacarne, V. Marchese, P. Massmann,
G.M. McCracken, P. McCullen, M.J. Mead, P. Meriguet, V. Merlo,
V. Mertens, S. Mills, P. Millward, A. Moissonnier, P.L. Mondino,
D. Moreau, P.D. Morgan, E.R. Mailer, D. Murmann, G. Murphy,
M.F. Nave, L. Nickesson, P. Nielsen, P. Noll, S. Nowak, W. Obert,
M. Olsson, J. O'Rourke, M.G. Pacco, J. Paillere, S. Papastergiou,
J. Partridge, D. Pasini, N.J. Peacock, M. Pescatore, J. Plancoulaine,
J.-P. Pqffe, R. Prentice, T. Raimondi, J. Ramette, C. Raymond,
P.H. Rebut, J. Removille, W. Riediker, R. Roberts, D.C. Robinson,
A. Rolfe, R.T. Ross, G. Rupprecht, R. Rushton, H.C. Sack,

G. Sadler, J. Saffert, N. Salmon, F. Sand, A. Santagiustina,
R. Saunders, M. Schmid, F. C. Schuller, K. Selin, R. Shaw, A. Sibley,
D. Sigournay, R. Sillen, F. Simonet, R. Simonini, P. Smeulders,
J.A. Snipes, K. Sonnenberg, L. Sonnerup, M.F. Stamp, C.A. Steed,
D. Stork, P.E. Stott, T.E. Stringer, D.R. Summers, H.P. Summers,
A.J. Tagle, G. Tallents, A. Tanga, A. Taroni, A. Terrington, A. Tesini,
P.R. Thomas, E. Thompson, K. Thomsen, F. Tibone, R. Tivey,
T.N. Todd, P. Trevalion, M. Tschudin, B.J. Tubbing, P. Twynam,
E. Usselmann, H. Van der Beken, J.E. Van Montfoort,
M. von Hellermann, J. von Seggern, T. Wade, C. Walker,
B.A, Wallander, M. Walravens, K. Walter, M.L. Watkins,
M. Watson, D. Webberley, A. Wetter, J.A. Wesson, J. Wilks,
T. Winkel, C. Woodward, M. Wykes, D. Young, L. Zannelli, J. W. Zwart
Discussion459
Ion cyclotron heating and lower hybrid experiments on ASDEX(IAEA-CN-47/F-I-2)
461
K. Steinmetz, F.X. Soldner, D. Eckhartt, G. Janeschitz,
F. Leuterer, F. Mast, K, McCormick, J.-M. Noterdaeme, C. Setzensack,F. Wagner, F. Wesner, I. Baumler, G. Becker, H.S. Bosch,
M. Brambilla, F. Braun, H. Brocken, H. Derfler, A. Eberhagen,H.-U. Fahrbach, G. Fussmann, R. Fritsch, O. Gehre, J. Gernhardt,
G. von Gierke, E. Glock, O. Gruber, G. Haas, J. Hoftnann,F. Hofineister, K. Hubner, A. Izvozchikov, F. Karger, M. Keilhacker,
O. Kluber, M. Kornherr, K. Lackner, M. Lenoci, G. Lisitano,
E. Van Mark, H.M. Mayer, D. Meisel, V. Mertens, E.R. Mutter,
E. Munich, H. Murmann, H. Niedermeyer, Z.A. Pietrzyk, W. Poschenrieder,
S. Puri, H. Rapp, H. Riedler, H. Rohr, J. Roth, F. Ryter, K.H. Schmitter,
F. Schneider, G. Siller, P. Smeulders, E. Speth, K.-H. Steuer, T. Vien,
0. Vollmer, H. Wedler, D. Zasche, M. Zouhar
Discussion
474
Ion Bernstein wave heating experiments on PLT (IAEA-CN-47/F-I-3)
477
M. Ono, P. Beiersdorfer, R. Bell, S. Bernabei, A. Cavallo, A. Chmyga,
S.A. Cohen, P. Colestock, G. Gammel, G. Greene, J. Hosea, R. Kaita,
1. Lehrman, G. Mazzitelli, E. Mazzucato, D.H. McNeil!, M. Mori,
K. Sato, J. Stevens, S. Suckewer, J.R. Timberlake, V. Vershkov,
J.R. Wilson, A. Wouters
Discussion
484
Confinement studies on TEXTOR with high power, long pulse ICRH
(IAEA-CN-47/F-I-4)
485
P.E. Vandenplas, T. Delvigne, P. Descamps, F. Durodie, M. Jadoul,
R. Koch, D. Lebeau, A.M. Messiaen, D.I.C. Pearson,
R. Van Nieuwenhove, G. Van Oost, G. Van Wassenhove, R.R. Weynants,
H.L. Bay, G. Bertschinger, W. Bieger, P. Bogen, W. Brussau,

G.A. Campbell, Y. Cao, R.W. Conn, H. Conrads, K.H. Dippel,
H.G. Esser, K.H. Finken, G. Fuchs, B. Giesen, D.M. Goebel,
E. Graffmann, H. Hartwig, E. Hintz, F. Hoenen, K. Hothker,
W.L. Hsu, A. Kaleck, H. Kever, L. Konen, M. Korten, W.K. Leung, L. Li,
Y. T. Lie, A. Pospieszczyk, D. Reiter, A. Rogister, G. Ross, D. Rusbuldt,
U. Samm, J. Schluter, B. Schweer, H. Soltwisch, G.J. Thomas,
F. Waelbroeck, G. Waidmann, P. Wienhold, J. Winter, G.H. Wolf
Experiments on ICRF heating and fast wave current drive in JIPP T-IIU(IAEA-CN-47/F-I-5)
T. Watari, K. Ohkubo, R. Akiyama, R. Ando, D. Eckhartt, Y. Hatnada,
S. Hirokura, K. Ida, E. Kako, O. Kaneko, K. Kawahata, Y. Kawasumi,
S. Kitagawa, T. Kuroda, K. Masai, K. Matsuoka, A. Mohri, S. Morita,
M. Mugishima, N. Noda, I. Ogawa, Y. Ogawa, Y. Oka, M. Sakamoto,
K. Sakurai, M. Sasao, K. Sato, K.N. Sato, S. Tanahashi, Y. Taniguchi,
T. Tetsuka, K. Toi, K. Yamazaki
Discussion

495

501

Electron heating and MHD modes during lower hybrid current driveon PLT (IAEA-CN-47/F-II-l)
503
S. Bernabei, R. Bell, A. Cavallo, T.K. Chu, P. Colestock, J. Felt,
W. Hooke, J. Hosea, F.C. Jobes, E. Mazzucato, E. Meservey, R. Motley,
J.A. Murphy, M. Ono, R. Pinsker, J. Stevens, S. Von Goeler, J.R. Wilson
Radiofrequency heating and current drive experiments on Alcator C
and Versator II (IAEA-CN-47/F-II-2)
M. Porkolab, P. Bonoli, Kuoin Chen, C. Fiore, R. Granetz, D. Griffin,
D. Gwinn, S. Knowlton, B. Lipschultz, S.C. Luckhardt, E. Marmar,
M. Mayberry, F.S. McDermott, J. Moody, R. Parker, J. Rice,Y. Takase, J. Terry, S. Texter, S. Wolfe
Discussion
Current drive by lower hybrid waves at 3.7 GHz in Petula B(IAEA-CN-47/F-II-3)
F. Parlange, D. Van Houtte, H. Bottollier-Curtet, G. Briffod,
P. Chabert, M. Clement, C. David, A. Girard, C. Gormezano, W. Hess,
M. Hesse, G.T. Hoang, G. Ichtchenko, J.M. Rax, F. Ryter, J.C. Vallet
Discussion
Experiments on current drive and plasma heating by LH and EC waves
in the T-7 tokamak (IAEA-CN-47/F-II-4)
V.V. Alikaev, A.A. Borshchegovskij, V.V. Chist'akov, Yu.A. Gorelov,
V.I. Win, D.P. Ivanov, N.V. Ivanov, A.M. Kakurin, P.P. Khvostenko,
A.Ya. Kislov, V.A. Kochin, P.E. Kovrov, K.I. Likin, Yu.S. Maksimov,
I.B. Semenov, A.F. Shigaev, Yu.A. Sokolov, N.L. Vasin, V.V. Volkov,
J. batlov, V. Kopecky, L. KrySka

509

523
525

531
533

Electron cyclotron heating of lower hybrid current driven plasma in the
JFT-2M tokamak (IAEA-CN-47/F-II-5)
T. Yamamoto, A. Funahashi, K. Hoshino, S. Kasai, T. Kawakami,
H. Kawashima, T. Matoba, T. Matsuda, H. Matsumoto, Y. Miura,
M. Mori, K. Odajima, H. Ogawa, T. Ogawa, H. Ohtsuka, S. Sengoku,
T. Shoji, N. Suzuki, H. Tamai, Y. Uesugi, T. Yamauchi, M. Hasegawa,S. Takada, K. Hasegawa, A. Honda, I. Ishibori, T. Kashimura,Y. Kashiwa, M. Kazawa, K. Kikuchi, Y. Matsuzaki, K. Ohuchi,
H. Okano, T. Shibata, T. Shibuya, T. Shiina, K. Suzuki,
T. Tani, K. Yokoyama
Discussion
Electron cyclotron and lower hybrid current drive experiments in the
WT-2 and WT-3 tokamaks (IAEA-CN-47/F-II-6)
S. Tanaka, Y. Terumichi, T. Maekawa, M. Nakamura, T. Cho, A. Ando,
K. Ogura, H. Tanaka, M. Iida, S. Ide, K. Oho, S. Ozaki, R. Itatani,
H. Abe, Y. Yasaka, M. Fukao, K. Ishii, A. Mohri, S. Kubo
Discussion
Improvement of current drive and confinement by combination of
LHCD and NBI heating in JT-60 (IAEA-CN-47/F-II-7)
JT-60 Team: T. Abe, H. Aikawa, H. Akaoka, H. Akasaka, M. Akiba,
N. Akino, T. Akiyama, T. Ando, K. Annoh, N. Aoyagi, K. Arakawa,
M. Araki, K. Arimoto, M. Azumi, S. Chiba, M. Dairaku, N. Ebiswa,
T. Fujii, T. Fukuda, H. Furukawa, K. Hamamatsu, K. Hayashi,
M. Hara, K. Haraguchi, H. Hiratsuka, T. Hirayama, S. Hiroki,
K. Hiruta, M. Honda, H. Horiike, R. Hosoda, N. Hosogane, Y. Iida,
T. Iijima, K. Ikeda, Y. Ikeda, T. Imai, T. Inoue, N. Isaji, M. Isaka,
S. Ishida, N. Itige, T. ho, Y. Ito, A. Kaminaga, M. Kawai,
Y. Kawamata, K. Kawasaki, K. Kikuchi, M. Kikuchi, H. Kimura,
T. Kimura, H. Kishimoto, K. Kitahara, S. Kitamura, A. Kitsunezaki,
K. Kiyono, N. Kobayashi, K. Kodama, Y. Koide, T. Koike,
M. Komata, I. Kondo, S. Konoshima, H. Kubo, S. Kunieda,
S. Kurakata, K. Kurihara, M. Kuriyama, T. Kuroda, M. Kusaka,
Y. Kusama, S. Maehara, K. Maeno, S. Mase, S. Matsuda,
M. Matsukawa, T. Matsukawa, M. Matsuoka, N. Miya, K. Miyati,
Y. Miyo, K. Mizuhashi, M. Mizuno, R, Murai, Y. Murakami, M. Muto,
M. Nagami, A. Nagashima, K. Nagashima, T. Nagashima, S. Nagaya,
H. Nakamura, Y. Nakamura, M. Nemoto, Y. Neyatani, S. Niikura,
H. Ninomiya, T. Nishitani, H. Nomata, K. Obara, N. Ogiwara,
Y. Ohara, K. Ohasa, T. Ohga, H. Ohhara, M. Ohkubo, T. Ohshima,
K. Ohta, M. Ohta, M. Ohtaka, Y. Ohuchi, A. Oikawa, H. Okumura,Y. Okumura, K. Omori, S. Omori, Y. Omori, T. Ozeki, M. Saigusa,
K. Sakamoto, A. Sakasai, S. Sakata, M. Satou, M. Sawahata,
M. Seimiya, M. Seki, S. Seki, K. Shibanuma, R. Shimada, K. Shimizu,

545

551
553

562
563

M. Shimizu, Y. Shimomura, S. Shinozaki, H. Shirai, H. Shirakata,
M. Shitomi, K. Suganuma, T. Sugie, T. Sugiyama, H. Sunaoshi,
K. Suzuki, M. Suzuki, M. Suzuki, N. Suzuki, S. Suzuki, Y. Suzuki,
H. Takahashi, M. Takahashi, S. Takahashi, T. Takahashi,
M. Takasaki, M. Takatsu, A. Takeshita, H. Takeuchi, S. Tamura,
S. Tanaka, K. Tani, M. Terakado, T. Terakado, K. Tobita,
T. Tokutake, T. Totsuka, N. Toyoshima, H. Tsuda, T. Tsugita, S. Tsuji,
Y. Tsukahara, M. Tsuneoka, K. Uehara, M. Umehara, Y. Uramoto,H. Usami, K. Ushigusa, K. Usui, J. Yagyu, K. Yamada, M. Yamamoto,
0. Yamashita, Y. Yamashita, K. Yano, T. Yasukawa, K. Yokokura,
H. Yokomizo, H. Yoshida, M. Yoshikawa, Y. Yoshinari, R. Yoshino,
1. Yonekawa, K. Watanabe
Discussion
573
Current drive, confinement and instability control experiments with ECRH
on the CLEO tokamak (IAEA-CN-47/F-III-2)
575
D.C. Robinson, N.R. Ainsworth, P.R. Collins, A.N. Dellis, T. Edlington,
P. Haynes, S. Iyengar, I. Johnson, B. Lloyd, M. O'Brien,
B.J. Parham, A.C. Rivtere, D.F.H. Start, T.N. Todd
Discussion
585
Confinement in electron cyclotron heating experiments on Doublet III(IAEA-CN-47/F-IH-3)
587
R. Prater, S. Ejima, R.W. Harvey, A.J. Lieber, K. Matsuda,
C.P. Moeller, and the Doublet III Physics Group
Discussion
597
Electron cyclotron heating on TFR (IAEA-CN-47/F-III-4)
599
H.P.L. de Esch, J.A. Hoekzema, W.J. Schroder, E. Westerhof,
W. Kooijman, R. W. Polman, T.J. Schep, and Equipe TFR
Discussion
606
Investigation of plasma formation regimes for ICRF heating experimentsin the TUMAN-3 tokamak with magnetic compression(IAEA-CN-47/F-III-5)
607
L.G. Askinasi, N.E. Bogdanova, V.E. Golant, S.G. Goncharov,
M.P. Gryaznevich, V.V. D'yachenko, A. Izvozchikov, E.R. Its,
S.V. Krikunov, S.V. Lebedev, B.M. Lipin, I.P. Pavfov, G.T. Razdobarin,
V. V. Rozhdestvenskij, N. V. Sakharov, M.A. Khalilov, F. V. Chernyshev,
K.G. Shakhovets, O.N. Shcherbinin, S.E. Bender, Yu.A. Kostsov,
R.N. LitunovsMj, O.A. Minyaev, E.N. Rumyantsev, P.P. Teplov
Discussion
615
Alfven wave heating and current profile modification in TCA(IAEA-CN-47/F-III-6)
617
G. Besson, G.A. Collins, B.P. Duval, C. Hollenstein, A.A. Howling,
B. Joye, J.B. Lister, J.-M. Moret, F. Ryter, C.W. Simm, H. Weisen
Discussion625

Analysis of ICRF heating and current drive in tokamaks(IAEA-CN-47/F-IV-l)
627
A. Fukuyama, T. Morishita, Y. Kishimoto, K. Hamamatsu,
S.-I. Itoh, K. Itoh
Double helix (f X 6) current drive in the RF tokamak device RYTHMAC(IAEA-CN^7/F-IV-2)
635
M.J. Dutch, A.L McCarthy
Dependence of power absorbed in a plasma on Alfve'n heating antenna
parameters (IAEA-CN-47/F-IV-3)
645
A.G. Kirov, V.P. Sidorov, A.G. Elfimov, S.N. Lozovskij, L.F. Ruchko,
K. G. Komoshvili, V. V. Onishchenko, V. V. Dorokhov, A. V. Sukachev,
V.D. Samushia, L.Ya. Malykh, N.L. Gulasaryan, D.A. Vojtenko
Recent advances in the theory and modelling of RF heating ininhomogeneous plasmas (IAEA-CN-47/F-IV-4)
653
D.G. Swanson, S. Cho, C.K. Phillips, D.Q. Hwang, W. Houlberg,
L. Hively
Electron cyclotron current drive and supra-thermal generation inlarge tokamaks (IAEA-CN-47/F-IV-5)
659
T.J. Schep, R. W. Polman, E. Westerhof, L. Klieb
Chairmen of Sessions and Secretariat of the Conference669

Session A
ARTSIMOVICH MEMORIAL LECTURE
AND
TOKAMAK EXPERIMENTS
Chairmen of Sessions
Session A-I

S. MORI

Japan

Session A-II

D.M. MEADE

USA

Session A-III

J.B. LISTERSwitzerland

Session A-IV

M. KEILHACKER

JET

Session A-V

R.A. GROSS

USA

Session A-VI

L.E. ZAKHAROVUSSRPapers A-VI-1-1, A-VI-1-2 and A-VI-1-3
were presented by D.F. Diichs as Rapporteur
Session A-VII (Posters)

IAEA-CN-47/A-0

ARTSIMOVICH MEMORIAL LECTURE
J.F. CLARKEAssociate Director for Fusion Energy,
Office of Energy Research,
United States Department of Energy,
Washington, D.C.,
United States of America

Mr. Chairman, Ladies and Gentlemen,
I want to thank the International Atomic Energy Agency (IAEA) for asking me
to give this address. I consider it an honour to be here, to be speaking to you —scientists of many races and nationalities united by a common effort to provide fusion
power to the world. I am especially pleased to join you in honouring the memory of
Lev Artsimovich, a man who contributed so much to the achievement of our goal.
This occasion is particularly meaningful to me, because it marks the 25th
anniversary of my own involvement in fusion research. In 1961, as a young doctoralcandidate in physics at MIT, I was fortunate enough to be offered the opportunity
to choose my field of study under masters of almost all of the disciplines of physics.
For a young man, the choice was bewildering. After much thought, I finally settled
on a field for a most unscientific reason. That field was the relatively undeveloped— and at the time unprestigeous — field of plasma physics. The reason for my choice
was a simple one. I learned that plasma physics might hold the key to unlimited
energy for the people of the world. That possibility engaged the enthusiasm of ayoung man. I have never regretted that decision. Over the years, the science ofplasma physics has grown into a major scientific field and has provided all of theprofessional satisfaction and excitement a scientist could desire. Moreover, my
primary motivation for choosing plasma physics — the benefit that might one day
flow from the science — has been strengthened as the reality of the promised
contribution to the future well-being of humanity has become more certain.
I believe that this motivation is a distinguishing characteristic that unites fusion
scientists. It is widely recognized that all scientists are united by their common
professional endeavours. This is certainly true in fusion research. However, inworking with fusion scientists and engineers from many countries, I have found astronger bond than merely sharing scientific skills — a bond that comes fromapplying those skills in the service of humanity.
This lecture honours the memory of one of our fraternity. A man who exemplified dedication to this higher purpose. Lev Artsimovich will always be remembered
as a giant in the development of fusion power. His scientific contributions werecertainly important, and yet, his stature rests on more than this. He was a giant

4

CLARKE

because of his wisdom, courage, skill, and, most of all, his dedication to our common
purpose.
Fortunately, he was not alone in these qualities. Throughout the history offusion development there have been a number of scientist/statesmen who shared the
dream of harnessing fusion power, but who also saw the task was beyond thecapability of their own laboratory. More importantly, these men had the wisdom to
see that — in practical if not in absolute sense — the task was possibly beyond thecapability of their own nation.
Over the years, men like Igor Kurchatov, Lev Artsimovich, Evgenij Velikhov,
Edward Teller, David Rose, Hans Otto Wiister, Donato Palumbo, Koji Husimi,
Tatsuoki Miyajima, and others have advocated international co-operation in thedevelopment of fusion power. Moreover, they had the courage to act to makeco-operation possible. In honouring these men for their wisdom, we must not ignore
their courage. Advocating co-operation takes wisdom. Acting upon it takes courage.
I really began to appreciate the degree of courage required of them as I prepared
for this address. Since Artsimovich is identified with international co-operation infusion, I began to reflect on the very first international meeting on fusion. It took
place at Harwell in Great Britain, in 1956 — thirty years ago. The date is important.
It was a time of great political tension and rivalry in the world. I have only to mention
the places: Indochina, Korea, Formosa, Berlin, Hungary, and Suez, to recall theoccasions of confrontation. To heighten the tension — if that were possible — themajor world powers had developed and were racing to deploy the new militarytechnologies of the hydrogen bomb and the ICBMs, and the world was on the vergeof the space race with the Soviet preparations to launch Sputnik. In a small way,
fusion research mirrored these tensions and rivalries. National programmes worked
in rivalry and secrecy to achieve the first fusion reaction and thereby obtain for theircountry a significant technological advantage — unlimited, economical energy or,
perhaps, even some military advantage. On an even more human scale, key figures
within national programmes struggled to expand their own laboratory's share in a
programme that they believed would surely make them Nobel Laureates.
Fortunately, then as now, the leaders of the major nations were seeking ways
to improve relations and reduce tensions, and scientists/statesmen were at theforefront of the initiatives. In 1953, President Eisenhower made his Atoms for Peace
proposal at the United Nations. By 1954, the United States and the Soviet Union had
agreed to consider exchanging atomic energy information, and by 1955, the first
Geneva Conference on the Peaceful Uses of Atomic Energy was held. At this Conference, Homi Bhabha of India first alerted the world at large to the potential of fusion
research. In 1956, Igor Kurchatov took the next step by openly discussing Soviet
fusion work while visiting Great Britain. The result was that by 1958, the time of
the Second Geneva Conference on the Peaceful Uses of Atomic Energy, the situationof secrecy and rivalry in fusion had changed. The papers of Artsimovich, Teller,
Biermann, Thonemann, and Alfven summarized previously hidden research infusion. As a result of these events, the European Countries, the United States, andthe Soviet Union agreed to open magnetic fusion research.

IAEA-CN-47/A-0

5Historians of the period record many political factors leading to the change.
However, I believe it was mainly a result of the wisdom, courage, and dedicationof both national political leaders and the scientific leaders in the national fusion
programmes. The scientific leaders realized that fusion would be achieved mostdirectly by uniting the scientific talents of the world. The political leaders recognizedthat co-operation on fusion — an effort in the interest of all mankind — could help
to reduce world tension and rivalry. They had the courage to take the risk and try
to establish international co-operation in fusion in spite of political tension and voicesof caution in their own countries. They had the wisdom to create useful co-operative
mechanisms, such as the IAEA, and the dedication to live up to these commitments
for almost three decades.
It is well to reflect on our beginnings and to realize that our scientific success
today rests on a remarkable foundation. It is simply that in fusion research, national
political leaders and fusion scientific leaders have been able to transcend their own
parochial interests in an effort to achieve a plentiful supply of energy for future
generations.
Today, we have a situation similar in many ways to that which existed in 1958.
Political tensions are high and the power and number of modern weapons has raised
the premium on reducing those tensions. On a political level, world leaders are
struggling with major issues, such as nuclear disarmament. As in the 1950s, they arealso looking toward co-operation on fusion as a means to reduce tensions and increase
understanding. Recently, Andronik Petrosyants, Chairman of the Soviet State
Committee on the Utilization of Atomic Energy, acknowledge this when he said that,in the view of the Soviet Union, scientific co-operation has great political significancein that it helps to strengthen peace and lessens the danger of new global war.
On the scientific level, fusion leaders are beginning to realize that in order toexploit our successes and complete the development of fusion power, a new degreeof collaboration may be necessary that will allow us to build facilities that go beyond
the resources available in any one national programme.
These factors led to the discussion of fusion at the Geneva Summit and at the
recent Reykjavik meeting. The Reagan-Gorbachev initiative to expand fusion
co-operation and the positive reactions of other world leaders to the initiative haveprovided the political mandate and opportunity to accelerate progress in our field. Iturges us to combine our resources in new ways for mutual benefit. The issue for us
to decide is "How should we, the international fusion community, respond to thisopportunity?"
I would like to discuss this question in the remainder of this lecture. First, it
is clear that there is a technical imperative for closer co-operation. This was clearlyrecognized by Artsimovich from the very beginning. At Geneva, in 1958, he said:' 'This problem (of fusion) seems to have been created especially for the purpose ofdeveloping close co-operation between scientists and engineers of various countries,
working at this problem according to a common plan, and continuously exchanging
the results of their calculations, experiments, and engineering developments." In the

6

CLARKE

years since these words were uttered, the nations of the world have developed aremarkable degree of co-operation in fusion. The most outstanding example is, ofcourse, the common fusion programme developed by the European Communities.
Within the European Atomic Energy Community, the words of Artsimovich havebeen borne out to the letter, and the obvious benefits are clearly evident in theachievements of the JET facility.
On the wider international scene, numerous co-operative projects andexchanges have also advanced the field. We have created effective mechanisms for
co-operation through the International Atomic Energy Agency, the International
Energy Agency (IEA), the US A/ Japan and the USA/USSR bilateral agreements infusion, and, most recently, the Fusion Working Group of the Economic Summit
countries. Although fusion co-operation outside of the European Summit countries.
Although fusion co-operation outside of the European Communities has not reached
the stage of working to a common plan described by Artsimovich, there has beensignificant movement in that direction over the last few years.
The science of fusion is by no means finished. The remaining scientific steps
will require new, expensive facilities and equally expensive operation of thesefacilities. The relatively new technology research will require perhaps more
expensive facilities. If it existed, the common plan envisioned by Artsimovich would
allow this work to be shared, duplication to be avoided, and the available technicalexpertise in the world to be used to best advantage. The practical benefit of this wasrecognized by Donato Palumbo in his 1980 Artsimovich Lecture. He questioned
whether each of the four big programmes could proceed individually to address next
generation fusion issues, and at the same time continue other necessary activities
along a wider front. A common plan would allow a broader attack on these difficult
problems within available resources. With this in mind, the Heads of State of the
Economic Summit nations set up the Fusion Working Group (FWG) to examine thefusion development plans of participating countries, in an effort to identify opportunities for working to a common plan.
The Fusion Working Group has agreed that the remaining problems of fusion
can be summarized by four critical issues. These issues are: improving magnetic
confinement concepts to make them more amenable to practical application,
investigating the scientific problems associated with a burning plasma, developing
materials suitable for fusion reactors, and solving the problems of fusion nuclear
technology. The work of the FWG in joint planning to resolve the four critical issues
is not finished, but already its practical effects are evident.
The FWG has examined national plans for addressing the burning plasma issue.
They concluded that a machine like the European NET, or the Japanese FER, isneeded in the late 1990s to complete the scientific base and to address the technicalfeasibility of fusion. They concluded that it was valuable and feasible to attempt a
short pulse ignition experiment in the early 1990s to gain realistic operating
experience with a burning fusion plasma. The United States of America has taken on
this latter task and has produced a conceptual design for a Compact Ignition Device

IAEA-CN-47/A-0

7(CIT) which, depending on details of funding, will operate as early as 1993. The
Europeans and Japanese have continued to develop designs for a facility with thebroader technical aim that might be ready for construction decisions at about the time
CIT is completed.
The FWG has also examined the national plans in fusion nuclear technology andmaterials. In the latter area, their deliberations were aided greatly by a detailed studyof the IEA members' fusion materials programmes by an IEA Blue Ribbon Panel,chaired by Professor Amelinckx of Belgium. In both areas, it was found that aco-ordinated international programme was possible and desirable.
Finally, in a recent meeting of the Co-ordinating Committee of the USA/USSR
Co-operative Agreement on the Peaceful Uses of Atomic Energy, it was agreed to
begin the process of joint planning in order to improve the co-ordination of the US
and the USSR fusion programmes.
So, in small steps, using all the mechanisms of co-operation established over
the years since Artsimovich made his statement in 1958, we appear to be moving
toward his common plan. However, the movement is not without difficulty. I havealready remarked on the courage of Artsimovich in making this proposal, considering
the time at which it was made. The courage demanded of us is becoming moreevident as we face the difficulty of actually implementing his proposal on aworld scale.
The difficulty is not technical. The above examples show that technical agreement can be reached if we wish to do so. The real difficulty was well described byHans-Otto WUster in testimony to the US Congress in 1985 on the process ofachieving a unified European fusion programme. He noted that international collaboration on the scale achieved in Europe is the result of the general insight that, becauseof resource limitations, research in separate nations could not be as successful as a
joint effort. As Wiister so pungently remarked, "After all, if you have a common
aim and want to do something, youd will find a way to give in to reason." If theworld fusion programme has lagged behind the European Communities in achieving
a common plan, it is probably due to lack of agreement on the necessity to work moreclosely together to achieve a common aim within our total financial and humanresources. Reaching agreement on the necessity of working together is difficult,because it involves many non-technical factors. It involves the personal ambitions ofscientists, as well as institutional ambitions of laboratories. The dedication to a common aim must be very strong to motivate an appropriate balance between such ambitions and more efficient pursuit of our common goal.
In considering how the fusion community should respond to the opportunity
presented by the Geneva Summit, we need to look more closely at this question. Whataim do we have in common and is it necessary that it be achieved jointly? The
heightened necessity for co-operation was perceived as long ago as 1977. David Rose
— a man who can justly be called the father of fusion technology — became
concerned about the possibility of reaching the goal of fusion power. His worry was
not primarily technical, although he recognized — perhaps more clearly than anyone

8

CLARKE

at the time — the real technical problems that would have to be overcome todemonstrate the scientific and technical feasibility of fusion. No, he foresaw that thesolution of these problems would require large, expensive facilities. He knew thatthese would also take a long time to plan, and operate. Finally, he knew that thepractical benefit of fusion would be in doubt until the end of this expensive and longterm programme. His worry was that national governments would be reluctant tospend such large sums for these facilities — let alone even more for larger, butessential support programmes — without clearer indication that fusion will work andbe more attractive than fission breeders or bulk solar power. His conclusion was thata world fusion programme conducted according to a common plan was needed if
these facilities were to be built.
His arguments defining the necessity of intensified co-operation led Director
General Sigvard Eklund, of the IAEA, to request the views of member governmentson the possibility of greater collaboration, and, in particular, their consideration of
how to take the major step after JET, JT-60, T-15, and TFTR — the step that would
complete the scientific base and demonstrate the technical feasibility of fusion. In
response, very much in the tradition of Artsimovich, Academician Velikhov
proposed that a joint group with participants from the European Communities, Japan,
the United States, and the USSR should be formed to investigate a project to build
a tokamak reactor experiment.
This proposal led, in 1978, to the organization of the INTOR Workshop. Overthe last eight years, this Workshop, under the leadership of Sigeru Mori, Giinter,
Grieger, Weston Stacey and Boris Kadomtsev, has clarified the nature and magnitudeof a facility that could complete the scientific base and address the technicalfeasibility of fusion. Through their pioneering efforts and the steadfast support of the
IAEA, we have developed a worldwide appreciation of the key role in fusion
development and the technical requirements of such a facility. The INTOR work has
been the essential ingredient in creating a world consensus on a major step in fusion.
During the past year, discussions based largely on the INTOR work have takenplace among the Economic Summit nations within the framework of the Fusion
Working Group, between the USA and the USSR in bilateral meetings, and among
key figures from all the world's major fusion programmes in the Director General'sInternational Fusion Research Council. All parties have reached a consensus that adevice of the INTOR type, generally known as an Experimental or Engineering Test
Reactor (ETR), is the common midterm goal of the world's fusion programmes. As
a result, we finally seem to have arrived at agreement on the common aim that Wiisterreferred to.
However, it remains to be seen whether we will find a way to give in to reason.
During the development of the ETR concept through the INTOR Workshop, all theparticipating nations put a considerable, if not principal, effort on defining their own
national versions of this next step — NET in Europe, FER in Japan, OTR in theSoviet Union, and FED, or more recently TIBER, in the United States. Having comeso far and having surmounted so many problems in international collaboration, one

IAEA-CN-47/A-0

9could take comfort in the thought that the passage of time will allow a coalescenceof these technical approaches. And yet, I cannot ignore the sense that history ispressing upon us to make a decision on how we wish to take this next critical stepin fusion development. We should all recall Professor Husimi's gentle reminder in
his 1982 Artsimovich memorial lecture, "...the public has for a long time been
eagerly waiting to see the production of fusion energy". I am also spurred in a notso gentle way by the knowledge that the world has spent between the equivalent of
US $15 billion and US $20 billion dollars to date in the quest for that production offusion energy. Once we have the technical capability, every year that we avoid adecision to move forward to the next step adds roughly another billion dollars to that
total. For this reason, I believe we must take every opportunity to move forward assoon as the technical capability becomes available.
Based on recent technical progress, I suspect the technical capability to build
an ETR can be attained within a few years. Therefore, it seems reasonable that weshould use the opportunity presented by the Geneva initiative to move as far as
possible in preparing for an ETR. If the existence of five different technicalapproaches prevents near term agreement on a common project, we should consider
taking the steps necessary to coalesce those technical approaches. If we lack thecomplete technical base to support commitment to construction, let us consider
co-ordinating the research of our programmes to provide the missing information.
Preparing one design and co-operating to prepare a sound technical base would notguarantee that an ETR would be built in common. However, it should help guarantee
both that it would be built and that whatever nation or group of nations undertook
the project would have a maximum chance at success.
Political leaders seem to recognize the need for collaboration in such tasks. In
the recent meeting in Reykjavik, President Reagan and General Secretary Gorbachev
agreed on a renewed effort to reach agreement on specific actions in expanding fusion
co-operation. I am very pleased that on October 30, the United States transmitted aformal proposal to the Economic Summit Nations and the Soviet Union. The proposal
was to join together to prepare a conceptual design report (CDR) for an ETR. It
recommended that we adopt the suggestion of Director General Hans Blix that webuild on the INTOR process to carry out the work. It also advocated that the researchand development necessary to support the CDR be undertaken co-operatively throughexisting agreements. The proposal indicated that the United States was prepared tocommit sufficient resources to accomplish this task within three years. If accepted,
this proposal could lead to a major focusing on world resources on preparing for an
ETR — the most widely accepted element on a common fusion plan.
Lev Artsimovich offered us the vision of scientists in all nations working
together on the fusion problem according to a common plan. Just as the vision of aworld served by fusion power engaged the enthusiasm of a young man twenty-fiveyears ago, the vision of a world united in a common fusion effort engages theenthusiasm of an older man today. However — perhaps because of age — it is anenthusiasm tempered by realism. Without in the least minimizing the difficulty of the

10

CLARKE

remaining problems of fusion science and technology, I would suggest that
attempting to merge our futures in a common plan may be a more difficult task. I
do not know if we can succeed. But then Artsimovich did not know that we would
ever overcome the difficult scientific problems to reach reactor-grade plasma conditions in a tokamak. He persevered in spite of the scientific difficulty and uncertainty.
I believe we, too, should persevere in spite of the human difficulty and uncertainty.
Lev Artsimovich and Hans-Otto Wiister would surely have recognized the difficulties
that must be overcome. Remembering what they attempted in their lives, I am
tempted to think they might have advised us in words similar to those of Professor
Husimi in his 1982 Artsimovich lecture: "Difficulty does not cause despair;difficulty invites courage, particularly when the goal is splendid."

IAEA-CN-47/A-M

RECENT EXPERIMENTS IN JT-60
JT-60 TEAM
(Presented by M. Yoshikawa)
T. Abe, H. Aikawa, H. Akaoka, H. Akasaka, M. Akiba, N. Akino, T. Akiyama,
T. Ando, K. Annoh, N. Aoyagi, K. Arakawa, M. Araki, K. Arimoto, M. Azumi,
S. Chiba, M. Dairaku, N. Ebiswa, T. Fujii, T. Fukuda, H. Furukawa,
K. Hamamatsu, K. Hayashi, M. Hara, K. Haraguchi, H. Hiratsuka, T. Hirayama,
S. Hiroki, K. Hiruta, M. Honda, H. Horiike, R. Hosoda, N. Hosogane, Y. Iida,
T. Iijima, K. Ikeda, Y. Ikeda, T. Imai, T. Inoue, N. Isaji, M. Isaka, S. Ishida,
N. Itige, T. Ito, Y. Ito, A. Kaminaga, M. Kawai, Y. Kawamata, K. Kawasaki,
K. Kikuchi, M. Kikuchi, H. Kimura, T. Kimura, H. Kishimoto, K. Kitahara,
S. Kitamura, A. Kitsunezaki, K. Kiyono, N. Kobayashi, K. Kodama, Y. Koide,
T. Koike, M. Komata, I. Kondo, S. Konoshima, H. Kubo, S. Kunieda,
S. Kurakata, K. Kurihara, M. Kuriyama, T. Kuroda, M. Kusaka, Y. Kusama,
S. Maehara, K. Maeno, S. Mase, S. Matsuda, M. Matsukawa, T. Matsukawa,
M. Matsuoka, N. Miya, K. Miyati, Y. Miyo, K. Mizuhashi, M. Mizuno, R. Murai,
Y. Murakami,
M. Muto,M. Nagami,
A. Nagashima,
K. Nagashima,
T. Nagashima, S. Nagaya, H. Nakamura, Y. Nakamura, M. Nemoto, Y. Neyatani,
S. Niikura, H. Ninomiya, T. Nishitani, H. Nomata, K. Obara, N. Ogiwara,
Y. Ohara, K. Ohasa, T. Ohga, H. Ohhara, M. Ohkubo, T. Ohshima, K. Ohta,
M. Ohta, M. Ohtaka, Y. Ohuchi, A. Oikawa, H. Okumura, Y. Okumura,
K. Omori, S. Omori, Y. Omori, T. Ozeki, M. Saigusa, K. Sakamoto, A. Sakasai,
S. Sakata, M. Satou, M. Sawahata, M. Seimiya, M. Seki, S. Seki, K. Shibanuma,
R. Shimada, K. Shimizu, M. Shimizu, Y. Shimomura, S. Shinozaki, H. Shirai,
H. Shirakata, M. Shitomi, K. Suganuma, T. Sugie, T. Sugiyama, H. Sunaoshi,
K. Suzuki,
M. Suzuki,
M. Suzuki,
N. Suzuki,
S. Suzuki,
Y. Suzuki,
H. Takahashi', M. Takahashi, S. Takahashi, T. Takahashi, M. Takasaki,
M. Takatsu, A. Takeshita, H. Takeuchi, S. Tamura, S. Tanaka, K. Tani,
M. Terakado, T. Terakado, K. Tobita, T. Tokutake, T. Totsuka, N. Toyoshima,
H. Tsuda, T. Tsugita, S. Tsuji, Y. Tsukahara, M. Tsuneoka, K. Uehara,
M. Umehara, Y. Uramoto, H. Usami, K. Ushigusa, K. Usui, J. Yagyu,
K. Yamada, M. Yamamoto, O. Yamashita, Y. Yamashita, K. Yano, T. Yasukawa,
K. Yokokura,
H. Yokomizo,
H. Yoshida,
M. Yoshikawa,Y. Yoshinari,
R. Yoshino, I. Yonekawa, K. Watanabe

Naka Fusion Energy Research Establishment,
Japan Atomic Energy Research Institute,Naka-machi, Naka-gun, Ibaraki-ken,
Japan

Abstract
RECENT EXPERIMENTS IN IT-60.
The paper describes the Joule heating experiment and more recent NBI and ICRF heating and
LHRF current drive experiments made in JT-60 in the past three months. JT-60 is a large tokamak witha poloidal divertor using TiC as first wall material. It is currently operated at a level of Ip < 2.0 MA,
1

Princeton Plasma Physics Laboratory, Princeton University, NJ, USA.

11

12

JT-60 TEAMP NB < 20 MW, PIC < 1.4 MW, P LH < 1.2 MW. The plasma is characterized by a wide range of
operational parameters: 0.1 x 1019 cm" 3 < ne < 9.7 X 1019 m " 3 and qeff > 2.2 and by a very
low impurity content, Zrff a 1.5, and Prad > 5 % . The energy confinement time is about 0.5 s for Jouleheated plasmas at high densities and 0.12 s for NB heated plasmas at high heating power. In the LHRF
experiment an LH driven current up to 1.5 MA is obtained, and a synergetic effect of LH current drive
and NB heating to enhance current drive and heating efficiency is observed. In the ICRF experiment,
the heating has an efficiency of 0.18 J-W" 1 and is found to reduce the hydrogen recycling.

1.

INTRODUCTION

JT-60 is a long pulse tokamak device with a plasma of circular cross-section anda poloidal divertor placed radially outside. Three means of intense heating, NBI,
LHRF, and ICRF are provided, the latter two of which can also be used for studying
RF current drive. The main parameters of JT-60 are listed in Table I. The operationof JT-60 was started in April 1985; it was shut down in July 1985 for the installationof heating systems. After the Joule heating experiment, which took about fivemonths, the heating and current drive experiments were started in August 1986.
The present paper reviews the present status of the JT-60 experiment on Joule
heating [1-4], NBI and ICRF heating, and LHRF current drive. The status of JT-60
will be described in the next section, and experimental results will be presented anddiscussed in later sections.

2.

STATUS OF JT-60

In this section, we present technical features of JT-60, in particular, those
relevant to the description of experimental results. Detailed technical descriptions of
JT-60 are provided in Refs [3-5]. Table I compares the designed and the achieved
values of the main parameters and shows that JT-60 is now being operated near its
maximum performance although more conditioning is required to raise the power
handling capability of the RF launchers.
The cross-section of the vacuum vessel is shown in Fig. 1. The toroidal limiters,
divertor plates, in-board armour plates, and the protective plates of the divertor coils
are made of molybdenum coated with TiC of 20 jum thickness. Protective plates
elsewhere are made of Inconel 625, coated with TiC of the same thickness. Since nolarge horizontal ports are available on the vacuum vessel, profile measurements rely
heavily on the four large ports at an angle of ±4° with the midplane as well as on
the array of vertical ports.
To block the backflow of discharge gas compressed in the divertor chamber,the gaps between the divertor coils and the vacuum vessel are sealed, except at thepumping ports where deflecting plates are placed to guide the backflow into thepumping ports. Additional Zr-AI getter pumps of an effective total pumping speed

IAEA-CN-47/A-I-1

TABLE I. DESIGNED AND ACHIEVED PARAMETERS OF JT-6O(a) JT-60 Parameters
Achieved(Divertor) (Limiter)
3.10 m3.04 mMajor Radius
3.0 m
0.90 m0.90 mMiner Radius
0.95 m
4.5 T
Toroidal Field* 4.5 T * * 4.5 T
2.0 MA
2.0 MA
Plasma Current
2.7 MA
10 s
10 s
Pulse Length
10 s* At a major radius of 3 m, ** 2.1 MA with divertor
Designed(b) JT-60 NBI Parameters
Achieved(Test)** (With plasma)
Total power into torus 20 MW 20 MW
20 MW
Number of beamlines
14
14
14
Number of sources
28
28
28
Beam energy
75 keV -100 keV£75 keV
Species power ratio*
80:13:7
75:16:9
Pulse length
10 s
10 sec2 s* H°(E): H°(E/2):H°(E/3) at 75 keV,** Each beamline tested separately in a test standDesigned(c) JT-60 RF Parameters
Designed

Achieved(Test)*(With plasma)
Lower Hybrid
Frequency
1.7-2.23GH2 1.7-2.23GHZ
2.0 GHz
Number of units33
1Power tubes per unit 8 klystrons 8
8Power at sources
24 MW
24 MW
2.0 MW
Power into torus
15 MW
1.2
_
Power into plasma
7.5 MW
Pulse length
10 s
10 s
0.75 sIon Cyclotron
Frequency
110-130MH2 110-130MHZ
120 MHz
Number' of units111Power tubes
8 Tetrodes 8
8Power at source
6 MW
1.6 MW
6 MW
5 MW
Power into torus
1.4 MW
Power into plasma
2.5MW
Pulse length
10 s
10 s
1.2 s
* Each unit tested separately13

JT-60 TEAM

14

DIAGNOSTICS PORTPOLOIDAL FIELD COIL
VACUUM VESSELDIVERTOR COILS
DIVERTOR PLATE
PUMPING PORT
DEFLECTION PLATE

FIG. I.

Cross-section of vacuum vessel.

of 5 m3 • s ' (H2) are connected at the divertor chamber to pump the hydrogen gas
built up in the chamber.
Nearly the entire diagnostic equipment manufactured is now in working order.
Its availability in the experiments is quite high. A list of diagnostic equipment is given
in Table II.

3.

OPERATIONAL CHARACTERISTICS

3.1. Machine operation
Figure 2 shows the time behaviour of typical divertor and limiter discharges in
JT-60. In most discharges, hydrogen prefill is used since helium prefill tends to leadto a disruptive discharge during the ramp-up phase. Later in the discharge, gas feedof hydrogen or helium (as in Fig. 2) is performed through gas valves which are placedat the main vacuum vessel or at the divertor chamber. The content of hydrogen in
the 'helium' discharge is not precisely known although an estimate from neutral beamshinethrough indicates that the helium ion content is higher than the hydrogen content
during the NB heating. In some discharges, the plasma density is feedback controlled
so as to evolve according to the instructions before the discharge.

IAEA-CN-47/A-I-1

15

TABLE II. DIAGNOSTICS IN JT-6O

PlasmaSub-systems
parameter
3ch, CH30H Laser (118.8 *nn)
FIR Interferometer
Electron
2-mmInterferometerlch
Density
Fourier Transform
Electron
ECE
Spectrometer
Temperature
Thomson Scattering
Multi-pulse Ruby/YAG Laser(6ch)
System
E=0.1 -llOkeV, E//B
CX Particle Analyser
ActiveBeam
Scattering
200keV-3,5A, He, e =7 g
Ion
SystemTemperature CX Recombination(on Active Beam)
Spectrometer
E =1.5-4 MeV
Neutron Counter
0.2 -500 A, Holographic Grating
VUV Spectrometer (2)5 -9000 A
Doppler Spectrometer
0.6-2.7 A
Crystal Spectrometer
2000-7000A,
Spectrograph,
VisibleSpectrometer
I
Impurity
Photo-Multiplier
Visible Spectrometer 11 The same as above, Rotating
MirrorGrazing Incidence
10-1300 A
Spectrometer
Radiation
Flux
Peripheral
PlasmaDivertor
PlasmaSX High Speed PHA
PIN Diode Array
Bolometer Array
HX Detector
Visible TV
Magnetic Probes,
Thermocouples
Ha Diode ArrayFIR Interferometer
Visible Spectrometer
VUV Spectrometer
Ho Diode Array
Bolometer Array
Thermocouple Array
IR & Visible TV

lch
17ch
16ch
3ch
61ch
4ch, 3ch, lch
2ch, CH3OH Laser
2ch
2 -1200 A, Holographic Grating
4ch
4ch(on Divertor Plate)2chJT-60 TEAM

16

LIMITER ( E 2 I 8 7 I

DIVERTOR IE2229)

0
>

3

^

2

I
,~0.5
il
O-

_ 0
i 4
° 2
x°
0

0

5

10

0

t (s)

FIG. 2. Divertor and limiter discharges with NB heating.

The discharge is usually initiated as a limiter discharge. In the case of divertor
experiments, it is then switched over to a divertor discharge by exciting the divertor
coil current with a typical delay of a few hundred milliseconds. Feedback control isapplied, with a sampling time of 1 ms, to the plasma current, and the vertical andthe horizontal field coil currents. The quadrupole and the divertor coil currents areeither preprogrammed (as in Fig. 2) or feedback controlled with special attention toprecise positioning of the separatrix magnetic surface with respect to the closest
limiter and the divertor plates [1]. In some of the discharges, the intersection of theseparatrix magnetic surfaces is swept over a width of several centimetres bymodulating the divertor coil current to reduce the peak heat load at the divertor plates.
The supplementary heating power can be applied at any desired time during thedischarge. The discharge is terminated by turning off the heating power and the gas
feed, and then the plasma current. In hydrogen discharges, the plasma density canbe reduced simultaneously with the plasma current, while in helium discharges the

IAEA-CN-47/A-M

17

density stays high until the end of the discharge because of the longer effective
particle confinement time. The discharge ends as a limiter discharge such as it has
been initiated.
3.2. Operational regime
JT-60 is rather routinely operated near its maximum rating (Table I), as
required by the experimental plan. Statistically, a large fraction of experiments ismade near the maximum toroidal magnetic field. The plasma current is usually variedwith steps of 0.5 MA up to 2 MA.
The operational regime of JT-60 divertor plasmas extends over a wide range ofplasma parameters. Figures 3(a) and 3(b) show the regime for NB heateddivertor (D) and limiter (L) discharges as well as for Joule heated hydrogen divertor
discharges. The lower limit of qeff is about 2.2 in both NB heated and Joule heateddischarges. Attempts to reduce qeff closer to 2.0 result in disruptive discharges.
In Joule heated discharges efforts are made to explore the upper limitsof He and the Murakami parameter neRp/BT, which are 9.7 X 1019 m~ 3 and6.6 x 1019 m ^ - T " 1 , respective. The lowest limit of ne achieved is
0.1 x 10 1 9 m" 3 . On the other hand, in NB heated discharges, no particularefforts are made to raise the upper limits of these parameters. In this sense, therespective limits of n,, and the Murakami parameter, of 8 X 10 l9 m" 3 and5.2 x 10~19 m " 2 - T " ' , should not be regarded as unsurpassable limits. The lower
end of fie is limited by the shinethrough power allowed by the armour plates of thevacuum vessel.
Figure 3(c) is a plot of NB heated discharges in the diagram of ne versuspabs = p jouie + Pinj - Pshinethrough for both divertor (D) and limiter (L) discharges. It
shows that the experiments have covered rather evenly the operational regime shownin this diagram.
3.3. Global effects of the divertor
A comparison of radiation losses from the main plasma as shown in Fig. 4
demonstrates the dramatic effect of the divertor in reducing the radiation loss in the
NB heated and the Joule heated plasmas. The radiation loss is typically 5-10% of
Pabs in divertor discharges and 30-90% in limiter discharges. It shows that thedivertor is particularly effective in reducing the radiation loss of the NB heatedplasmas and keeps Prad/Pabs below the Joule heating value.
Figure 2 shows a pronounced increase, during NB heating, of hydrogen radiation from the divertor chamber and a small change in the radiation in the main
vacuum vessel. This observation indicates that the particle loss in the plasma isenhanced during NB heating and also that the lost particles are efficiently transported
to the divertor chamber through the divertor throat similar to the observation made

JT-60 TEAM

18

OL H

1

i

i

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I

BOY CHOKE)

O Y =2-0MR

i

I

o

i K = l .SMfl•
xx=l-0MR

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OL HE

H
. * = l .8Mfl

m m R H HOOT

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. « = l .SMfl^** » - -- -i

" " X JOULE-HEATED

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3.0=OIHER

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(a)

<e.o

i

i i i i i i

BEZXZ S Z-

, 0 = 1 .OMA

I
9.0

6.0

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12.0

n e (10l9m-3)

1

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i

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DL HE

^x

0 4 < =2.0Mfl

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;

+

i

I
I

- x x x--H4-++MMI- +>

-

- x

-

•-

1

' * " '^JOULE-HEATED

(b)
i
2.0

i

1

4.0

1

1

1

6-0

M = ncRp/Bt(1Ol9m-2-T-1)

8.0

10.0

« = l .8Mfl

19

IAEA-CN-47/A-I-1

<s.o

2.0

4.0

6-0

(10l9rn-3)
FIG. 3. Operational regime of Joule heated and NB heated plasmas.
(a) iie <1019 m~3) versus Ip (MA);(b) nfiJB, (10l9/m2T) versus l/qep(c) ne (1019 m~3) versus P^ (MWj.

100

I P = 1.5 MA
Helium

*o

•

•8

a.

JouleHeated
Divertor
Limiter•50

NBHeated

A

A.

O

•

a.
\

7MW)

0

FIG. 4. ne versus

1 2

3

4

5

6

7

in Joule heated and NB heated plasmas.

20

JT-60 TEAM

2.5r
Kinetic Stored Energy
2.0-

H
2MA

0

He
•

I.5MA

D

•

1MA

A

A

2 MA

1.5(MJ)
1.0

Joule

(a)
20

15

0.5
\joule(b)
0.4Energy Confinement Timer

H

_ We +W|

0.3

He
•

2MA
I.5MA

0

•

•

IMA

A

A

(S)

0.2

O.I
(He, Llmiter, t5MA) A
10
Pabs

15
(MW)

20

21

IAEA-CN-47/A.-I-l
0.2

1

1

r

IP

Energy Conflnemenl Time
H
O

He
•

H

He I.5MA

D

(S)

•

Pabs

2 MA I4H7MW

9-IIMWO.I

ne(IO l9 rrf 3 )FIG. 5. Confinement in Joule heated and NB heated plasmas:(a) P^ versus W";(b) Pabs versus TE;(c) nt versus TE.

in Joule heated plasmas [1, 2, 4]. With respect to heat transport through the throat
to the divertor chamber, the temperature rise AT of the divertor plates is measured
during NB heating by a set of thermocouples embedded in the plates. The half-widthof the AT distribution across the separatrix is less than 2 cm, which corresponds toa half-width of less than about 1 cm at the divertor throat, small enough to assure
unobstructed transport of heat to the divertor plates.

4.

ENERGY CONFINEMENT AND TRANSPORT

Figure 5(a) shows the total kinetic stored energy W = We + W; + Wbeam
versus Pabs in the NB heated divertor and limiter discharges, as the density is varied.
It shows that the stored energy increases as an offset, linear function of Pabs up toa maximum of 2 MJ and that the energy confinement time is reduced by NB heating.
This L-mode-like behaviour can also be seen in Figs 5(b) and (c), where thedependence of the energy confinement time rE on other parameters is shown. rEdecreases from Joule heating values of 0.3-0.5 s, depending on the density [4], to0.1-0.12 s with NB heating and approaches an asymptotic value. It increases withfie and Ip, although the dependence is rather weak. No clear dependence on the ion
mass number was observed within the experimental ambiguity, which can be understood in view of the substantial hydrogen content in the helium discharges (seeSubsection 3.1). The degradation in the energy confinement in the core region isinterpreted as due to the enhancement in the electron conduction since none of theJT-60 TEAM

22

Diverior
P 0ts >10MW

100•

Missing

(a)

or

i

5 0 --

-

P 7/C

—]

'rod
r

---

,

moin
rad

10
fie

(xl0l9m3)
p = 1.5MA, He
p = 2.0MA,H s
p = 2.0MA, H

1

(b)

1

i
j

o

4-

1

1

1

1

[

,

|-<V=

ro
:^A
|O

o o
N

CM

1.0

A Bremsstrahlung
O Active Beam
I.5MA

i

— nV

A Bremsstrahlung• Active Beam

1

© Jv

• Resistivity

.

2.0MA

• Resistivity

i

A lo

Helium — Dlvertor
h- -

jo
1
1

OA®
~

A

k ) d
1 •
K

^ ®op"

•

^

A

••

• er

1

1
—J

1

1
1
1

1
1
1
1

0.0

1 [VIM

1
1®

i ®

o

Hydrogen — Dlvertor
1 MA

'

H

j

•

D
D

IMA
v Bremsstrahlung
1.5MA

D

•

•

Bremsstrahlung-

O -with NB heating

1

i

1

1

1.0

2.0

3.0

4.0

5.0

3

<10"m )
FIG. 6. (a) Energy accounting in the NB heated divertor plasma.
P™^ and PjJJU are the radiation loss from the main and the divertor plasmas, respectively.
energy loss to the divertor plates measured by thermocouples;(b) Zeff versus ne for Joule heated and NB heated plasmas.

is theIAEA-CN-47/A-M

23

reasonably enhanced neoclassical ion, convection, and radiation losses is large
enough to account for the experimentally observed values. More detailed discussions
on this subject are given in a companion paper [6].
A typical set of plasma parameters obtained in the NB heated plasmas is Ti(O)
6 keV, Te(0) 4 keV, n, 3 x 1019 m" 3 at Pabs 14 MW. In an attempt to improve theconfinement, a limited number of discharges with H-mode characteristics are
obtained. They are, however, not reproducible and need further study.
In Fig. 6(a) the power losses observed in the main vacuum vessel and in thedivertor chamber are compared with Pat,s. There is a good accountability between
them in the NB heated plasmas as was observed earlier in the Joule heateddischarges [1].

5.

IMPURITIES

Major species of impurities that are observed spectroscopically in hydrogen
divertor discharges are oxygen, carbon, titanium (from TiC coating) with and without
NB heating. In some of the NB heated plasmas a molybdenum burst is observed
during the NB heating, which can, in most cases, be eliminated by sweeping theseparatrix magnetic surface over the divertor plates (see Subsection 3.1). The
titanium ion content near the axis, estimated from the dominant Ti XXI intensity
observed along the viewline through the vessel axis, decreases roughly in inverse
proportion to the plasma density although the shot to shot fluctuation is fairly large.
The titanium content is about 1 x 10 ~5 in Joule heated hydrogen divertor plasmas
at iie 2 x 1019 m" 3 and Ip 2 MA and increases to about 2 X 10 ~5 in NB heatedplasmas with Pabs > 10 MW. These values are too small to contribute to Zeff andPrad. In Joule heated hydrogen limiter plasmas, on the other hand, the titanium
content is much larger, 3 X 10 ~4, which is still too small to affect Zeff and may
have a small contribution to Prad.
The oxygen and carbon contents are not quantified by spectroscopic measurements. However, their combined effects are estimated from the energy distributionof helium scattered near the axis from the helium active beam. The amount thus
estimated is equivalent to 1.2% in oxygen or, alternatively, to 2.4% in carbon in
Joule heated hydrogen divertor plasmas at n,. 3.4 X 1019 m~3. With NB heating, it
increases up to 1.8% and 3.6%, respectively. These values are naturally consistent
with Zeff(0) as measured by the helium scattering method to be described below. No
direct information is as yet available on the spatial distributions of the impurities.
Zeff(0) is estimated by three methods: visible bremsstrahlung, helium active
beam scattering, and resistivity analyses (assuming Spitzer resistivity; theneoclassical resistivity yields Zeff(0) below unity in hydrogen discharges).
Figure 6(b) shows Zeff(0) in various kinds of discharges with and without NB
heating. These methods consistently give Zeff(0) 1.5-2 for h^ > 3 X 1019 m~3hydrogen divertor discharges with and without NB heating.

JT-60 TEAM

24
2.5
•g

Gas puff

Joule plasma

Main 1 Div.
0

2.0

1

•

(2-4)x1
Hydroge n
Divertor

1
1.5

1.0
i

*

0

0

0.5

5

15

10

20

Pobs
FIG. 7. Effective particle confinement time r * versus

0.1- - 0 85 MW

10

(a)

o

9

A

(0 ,0)(7T , 0 )(0 , * )

A

(TT

•

8

,*)

7

_____ 6
a
a. 5
o
or
4
3
2
Vacuum- 1

30

35

40

d (mm)

45

50

IAEA-CN-47/A-M

Q925

Hydrogen
Divertor

0.8

ICRF I.1MW

0.75 MW

Q7

fE 4

2

2
«

3

F/G. & /CKF heating experiment:(a) Antenna loading;(b) Plasma behaviour.

6.

PARTICLE CONFINEMENT AND TRANSPORT

Particle confinement in the NB heated hydrogen divertor plasmas is evaluatedfrom the density profile and influx of fast (NBI) and slow hydrogen atoms. The
diffusion coefficient D in the plasma core is about 0.5 m 2 -s"' atfie = (4.5-5) X 1019 m" 3 and Ip = 2 MA for Pabs = 20 MW and increases inproportion to Pabs, when a shaping parameter of C v = 0 is assumed. In contrast, thediffusion coefficient is, at least, an order of magnitude smaller in Joule heatedplasmas. The diffusion coefficient of titanium impurity in Joule heated plasmas isevaluated from the temporal evolution of titanium lines following an accidental
penetration of a minute titanium piece into the plasma and is about 1 m 2 -s"' with
Cv~0.
Recycling behaviour in Joule heated hydrogen plasmas has been reported forboth divertor and limiter plasmas [1, 3]. The effective particle confinement time,JT-60 TEAM

26

3.0

4.0

6.0

5.0
Time

(s)

2.0
• 0H + LHCDnCD=o NBI HHCD

/

/

/

(b)

3 .0

n
/

•

/

>/

/

1.0 o

/ °/

/

V

1

0

1

2

3
19

4

3

P L n(MW)/n e (x!0 m" )FIG. 9. LHRF current drive experiment:(a) Plasma behaviour with and without LH power shown by full and broken tine, respectively;(b) Efficiency.

IAEA-CN-47/A-I-1

27

m~ 3
and is proportional to the plasma density, T* in NB heated hydrogen divertor
plasmas is plotted in Fig. 7, as a function of PNB = Pinj - Pshinethrough for a density
range of (2-4) X 1019 m~ 3 . It shows that T*V decreases with PNB and is independentof the plasma density. The hydrogen pressure, Pdiv, in the divertor chamberincreases in proportion to fie in the NB heated plasmas as in the Joule heatedplasmas [2, 3], indicating compression of the hydrogen gas in the divertor chamber.
The inverse normalized pressure in the divertor chamber, njr/P, decreases with PNB
and then tends to an asymptotic value. This observation and the aforementioned
increase of D with Pabs consistently indicate the enhancement of recycling during
NB heating, due to degradation of particle confinement.
T* = T P /(1-R), is about 2 s and 5.5 s, respectively, at a density of 3 X 107.

19

ICRF HEATINGThe ICRF heating experiment with a 2coci resonance on the axis is conductedwith an injected power of P IC < 1.4 MW, injected from a 2 x 2 loop antenna array.
Figure 8(a) shows the antenna loading resistance R^, as a function of the Faradayshield-to-separatrix distance d for various choices of the phase difference (Ap, Ad)of the loops. The plasma is a hydrogen divertor plasma with r^ = 3 X 1019 m" 3
and Ip = 1.5 MA at 4 T. R,.p as high as 10 Q is obtained with (0, 0) phasedifference. This value of R^ is larger than the values evaluated from a simplified
theoretical model [8].
The increase in the kinetic stored energy AW at PIC = 1.2 MW is 0.20 MJ,
corresponding to AW/PIC of 0.17 J- W" 1 . Figure 8(b) shows the plasma behaviour
with and without ICRF heating, in particular, an increase in stored energy and areduction of the hydrogen radiation from the main and divertor plasmas.
8.

LHRF CURRENT DRIVE

The current drive experiment is performed by using a 4 x 8 waveguidelauncher fed by one of the LHRF power units [9]. Experiments are performed at an
injected power PLH of up to 1.2 MW at Ip = 0.5-1.5 MA into a hydrogen divertor
plasma of H,. = (0.1-2) X 1019 m~3. With a phase difference A<p of ir/2
corresponding to waves peaked at an n|| value of about 1.7, evidence is obtained
for a current drive as shown in Fig. 9(a) for two cases with and without LH power.
The NB heating is timed so that its effect can be seen clearly. It is interesting tonote that NB heating and LH current drive are synergetic. In Fig. 9(b), the driven
current IRF, estimated as I p |AV,|/V f from the loop voltage V,, is plotted as afunction of PLH/ne. The current drive efficiency ?;CD = ^ R I ^ / P R F
is (1-1.7) x 1019 A-m- 2 .W-' without NB heating and increases to(2-2.8) x 1019 A-m" 2 -W-' with NB heating. The stored energy during the NB
heating is improved by 30% throughout the current drive.

28

JT-60 TEAM

9.

CONCLUSIONS

JT-60, as well as its heating and diagnostics systems, has been operated at near
maximum capabilities. The main results obtained in recent experiments are:(1)

The experiments are carried out over a wide operational regime:
0.1 x 1019 m~3 < n < 9.7 x 1019 m" 3 , qeff > 2.2, and Pabs < 20 MW.
A Wtot of up to 2.0 MJ is obtained.
(2) The energy confinement time is up to 0.5 s in Joule heated plasmas, dependingon the plasma density. With NB heating, it is reduced to 0.12 s at high heating
power; it is weakly dependent on density and plasma current. This behaviouris accompanied by a reduction of the particle confinement time.
(3) The divertor is effective in reducing impurities with Prad/Pabs ^ 5 % andZeff > 1.5 in the NB heated plasmas. This indicates that a first wall material
with medium Z can be used at reactor grade plasma parameters. Compressionof hydrogen gas in the divertor chamber is observed.
(4) In the LH current drive experiment a current equivalent to 1.5 MA is driven.
The current drive efficiency riCD is as high as (1-1.7) X 1019 A-m~ 2 W-'. The
efficiency is further increased to (2-2.8) x 1019 A-m~ 2 W~' by NB heating,which is accompanied by the enhancement of the NB heating efficiency by the
LH current drive. In the ICRF heating experiment an effective heating of
AW*/P]C = 0.17 J-W" 1 is observed and is accompanied by the reduction ofhydrogen recycling.
(5) Attempts to improve confinement are being made.

ACKNOWLEDGEMENTS
The authors wish to thank the members of JAERI who have contributed
to the JT-60 project throughout its progress. They also wish to express gratitude
to Drs S. Mori, Y. Iso, and K. Tomabechi for continuous encouragement andsupport.

REFERENCES
[1]
[2]
[3]
[4]
[5]

JT-60 TEAM (presented by M. Yoshikawa), Plasma Phys. Controll. Fusion 28 (1986) 165.
JT-60 TEAM (presented by S. Tsuji), in Controlled Fusion and Plasma Physics (Proc. 12th
Europ. Conf. Budapest, 1985), Part I, European Physical Society (1985) 375.
JT-60 TEAM (presented by S. Tamura), Plasma Phys. Controll. Fusion 28 (1986) 1377.
JT-60 TEAM (presented by H. Kishimoto), in Plasma-Surface Interactions in Controlled Fusion
Devices (Proc. 7th Int. Conf. Princeton, NJ, 1986).
Nucl. Eng. Des./Fusion (Special Issue), to be published.

IAEA-CN-47/A-I-1

[6]
[7]
[8]
[9]

29

JT-60 TEAM (presented by M. Nagami), Paper IAEA-CN-47/A-II-2, these Proceedings, Vol. 1.
JT-60 TEAM (presented by H. Takeuchi), Paper IAEA-CN-47/A-IV-3, these Proceedings,
Vol. 1.
KIMURA, H., et al., in Heating in Toroidal Plasmas (Proc. 4th Int. Symp. Rome, 1984), Vol. 2,
ENEA, Frascati (1984) 1128.
JT-60 TEAM (presented by T. Imai), Paper IAEA-CN-47/F-II-7, these Proceedings, Vol. 1.

DISCUSSION
R.J. GOLDSTON: I am interested in your measurements of H a emisssion in
the main chamber. You showed that it rose very little during neutral injection into
divertor discharges. Whereabouts in the main chamber do you measure the H aemission and how does the emission near the divertor throat behave?
M. YOSHIKAWA: We have an array of H a detectors to measure the Ha
intensity. The signal from the detector in the direction of the throat is much larger
than the other signals.
M. KEILHACKER: Since H-mode discharges have been achieved recently inboth JET and DIII-D, could you comment on the progress made in H-mode researchon JT-60?
M. YOSHIKAWA: We have spent a considerable amount of experimental time
investigating the H-mode. Although a small number of shots exhibit characteristicsof the H-mode type, we are not convinced that H-mode operation is achieved, sincethese shots are not reproducible and limited in number.
J.G. CORDEY: Do you have a physical explanation for the improvement in theefficiency of the lower hybrid current drive with the addition of neutral beam
injection?
M. YOSHIKAWA: No, we are not yet able to interpret the improvement
clearly. One possible explanation is that the electron energy distribution is broadened
at the upper end by neutral beam heating and improves the efficiency of the lower
hybrid current drive, whereas the ion energy distribution is not disturbed and does
not affect the wave power.
B. COPPI: What type of density profile did you observe with Ohmic discharges
and auxiliary heated discharges?
M. YOSHIKAWA: The density profile in JT-60 is generally wide. One
interesting observation in association with recent TFTR experiments is that thedensity profile broadens when the confinement of the neutral beam heated plasma is
improved by lower hybrid current drive.

IAEA-CN-47/A-I-2

JET LATEST RESULTS
AND FUTURE PROSPECTS
JET TEAM(Presented by P.H. Rebut)
H. Altmann, R.J. Anderson, J. Arbez, W. Bailey, D.V. Bartlett, B. Beaumont1,
G. Beaumont, K. Behringer, P. Bertoldi, E. Bertolini, C.H. Best, V. Bhatnagar2,
R.J. Bickerton, A. Boileau3, F. Bombarda4, T. Bonicelli, S. Booth, A. Boschi,
G. Bosia, M. Botman, H. Brelen, H. Brinkschulte, M.L.Browne, M. Brusati,
T. Budd, M. Bures, P. Butcher, H. Buttgereit, D. Cacaut, C. Caldwell-Nichols,
J.D. Callen5, D.J. Campbell, J. Carwardine, G. Celentano, C D . Challis,
A. Cheetham, J.P. Christiansen, C. Christodoulopoulos, P. Chuilon, R. Claesen,
J.P. Coad, S.A. Cohen6,
M. Cooke, J.G. Cordey,
W. Core, S. Corti,
A.E. Costley, G.A. Cottrell, M. Cox 7 , C. David, J. Dean, L. de Kock,
E. Deksnis, G.B. Denne, G. Deschamps, K.J. Dietz, J. Dobbing, S.E. Dorling,
D.F. Diichs, G. Duesing, P.A. Duperrex8, H. Duquenoy, A.W. Edwards,
J. Ehrenberg9,
W. Engelhardt,
S.K. Erents 7 ,F. Erhorn,
B.T. Eriksson,
M. Evrard2, H. Falter, N. Foden, M.J. Forrest7, C. Froger, K. Fullard,
G. Fussmann9,
M. Gadeberg10, A. Galetsas, A. Gallacher,
D. Gambier',
R. Giannella", A. Giannelli4, A. Gibson, R.D. Gill, A. Goede, A. Gondhalekar,
C. Gordon, C. Gormezano, N.A. Gottardi, C.W. Gowers, R. Granetz, B. Green,
S. Gregoli, F.S. Griph, R. Haange, J.H. HamneV 1 , C.J. Hancock, P.J. Harbour,
N.C. Hawkes7, P. Haynes7, T. Hellsten, J.L. Hemmerich, R. Hemsworth,
F. Hendriks, R.F. Herzog, L. Horton12, J. How, M. Huart, A. Hubbard13,
J. Hugill7, M. Hugon, M. Huguet, B. Ingram, H. Jackel9, J. Jacquinot,
Z. Jankowicz' 4 , O.N. Jarvis, E.M. Jones, P. Jones, T.T.C. Jones, C. Jup6n15,
E. Kallne, J. Kallne, O. Kaneko16, A. Kaye, B.E. Keen, M. Keilhacker,
G. Kinahan, S. Kissel, A. Konstantellos, M. Kovanen17, U. Kiiphnapfel9,
P. Kupschus, P. Lallia, J.R. Last, L. Lauro-Taroni, K.D. Lawson7, E. Lazzaro,
R.C. Lobel, P.J. Lomas, N. Lopes-Cardozo18, M. Lorenz-Gottardi, C. Lowry13,
G. Magyar,
D. Maisonnier,
M. Malacarne,
V. Marchese,
P. Massmann,
G.M. McCracken7,
P. McCullen,
M.J. Mead,
P. Meriguet,
V. Merlo,
V. Mertens9, S. Mills, P. Millward, A. Moissonnier, P.L. Mondino, D. Moreau',
P.D. Morgan,
E.R. Miiller9,
D. Murmann9,
G. Murphy,
M.F. Nave,L. Nickesson, P. Nielsen, P. Noll, S. Nowak, W. Obert, M. Olsson, J. O'Rourke,
M.G. Pacco, J. Paillere, S. Papastergiou, J. Partridge7, D. Pasini12, N.J. Peacock7,
M. Pescatore, J. Plancoulaine, J.-P. PofK, R. Prentice, T. Raimondi, J. Ramette1,
C. Raymond,
P.H. Rebut,
J. Removille,
W. Riediker,
R. Roberts,
D.C. Robinson7, A. Rolfe, R.T. Ross, G. Rupprecht9, R. Rushton, H.C. Sack,
G. Sadler, J. Saffert, N. Salmon13, F. Sand, A. Santagiustina, R. Saunders,M. Schmid, F.C. Schuller, K. Selin, R. Shaw, A. Sibley, D. Sigournay,
R. Sillen18, F. Simonet1, R. Simonini, P. Smeulders, J.A. Snipes7, K. Sonnenberg,
L. Sonnerup, ,M.F. Stamp, C.A. Steed, D. Stork, P.E. Stott, T.E. Stringer,
D.R. Summers, H.P. Summers19, A.J. Tagle, G. Tallents20, A. Tanga, A. Taroni,
A. Terrington, A. Tesini, P.R. Thomas, E. Thompson, K. Thomsen10, F. Tibone,
R. Tivey, T.N. Todd7, P. Trevalion, M. Tschudin, B.J. Tubbing18, P. Twynam,
E. Usselmann, H. Van der Beken, J.E. Van Montfoort, M. von Hellermann,
J. von Seggern21,
T. Wade, C. Walker,
B.A. Wallander,
M. Walravens,

31

32

JET TEAM
K. Walter, M.L. Watkins, M. Watson, D. Webberley, A. Weller9, J.A. Wesson,
J. Wilks, T. Winkel, C. Woodward, M. Wykes, D. Young, L. Zannelli,
J.W. Zwart

JET Joint Undertaking,
Abingdon, Oxfordshire,
United Kingdom

Abstract
J E T L A T E S T RESULTS A N D F U T U R E P R O S P E C T S .
During J E T ' s first three years of operation, experiments have been undertaken with up t o~ 7 M W of I C R F , ~ 10 M W of neutral beam and ~ 18 M W of total input power during combined
operation. The latest J E T results show many advances made during this period. In particular, the especially good confinement ( T E up to 0.8 s), high temperatures ( T e up to 5 keV) and fusion product({ftiTETj) = 10 2 0 m " 3 ' S - k e V ) achieved with ohmic heating alone have exceeded expectations. Also,very efficient ion heating has been observed (Tj > 12.5 ± 1.5 keV) at moderate neutral beam injection power levels (P N B < 8 M W ) and at low average densities (flj ~ (1-1.5) x 10 1 9 m ~ 3 ) . In addition, a fusion product of ((hj^tj— 2 x 10 2 0 m ~ 3 - S ' k e V has been achieved with a separatrix configuration in JET. However, limitations have been encountered in plasma density and electron and ion
temperatures because of disruptions and sawtooth oscillations. With additional heating, confinement isdegraded, as the electron temperature outside the sawtooth region exhibits only a weak dependence on
the input power per particle. As a consequence, additional heating has little impact on the fusion
product. T h e consequent developments to provide significant alpha heating in J E T are described.

1

CEA, Centre d'etades nuclfiaires de Fontenay-aux-Roses, France.
Plasma Physics Laboratory, Association "Euratom-Belgian State" E R M / K M S , Brussels, Belgium.
3
Institut national de la recherche scientifique, Ste-Foy, Quebec, Canada.
4
E N E A , Centro Ricerche Energia, Frascati, Italy.
5
Department of Physics, University of Wisconsin, Madison, W I , U S A .
6
Princeton Plasma Physics Laboratory, Princeton University, NJ, U S A .
7
U K A E A , Culham Laboratory, Abingdon, Oxfordshire, UK.
8
Centre de recherches en physique des plasmas, Ecole polytechnique fe'de'rale de Lausanne,
Switzerland.
9
Max-Planck-Institut fur Plasmaphysik, Garching, Fed. Rep. Germany.
10
Risa National Laboratory, Roskilde, Denmark.
" Swedish Energy Research Commission, Stockholm, Sweden.
12
Natural Sciences and Engineering Research Council, Ottawa, Ontario, Canada.
13
Imperial College of Science and Technology, University of London, U K .
14
R C C C Y F R O N E T IAE, Otwock-Swierk, Poland.
15
University of Lund, Sweden.
16
Department of Physics, University of Nagoya, Japan.
17
Department of Physics, Lappeenranta University of Technology, Lappeenranta, Finland.
' 8 F O M Instituut voor Plasmafysica, Rijnhuizen, Nieuwegein, Netherlands.
19
University of Strathclyde, Glasgow, UK.
20
Research School of Physical Sciences, Australian National University, Canberra, Australia.
21
Kernforschungsanlage Jiilich G m b H , Fed. Rep. Germany.
2

IAEA-CN-47/A-I-2

33

1. Introduction
The Joint European Torus (JET) [ 1,2] is the largest single project of the coordinated fusion research programme of the European Atomic Energy Community (EURATOM). Its
main objective is to obtain and study plasmas in conditions and with dimensions approaching those needed for a fusion reactor (ie n = 2x 1020m"3; T> lOkeV, and rE= l-2s),
and is specifically designed to study D-T reactions in the plasma.
This paper concentrates on experiments and advances during the last 2 years (the first
additional heating phase) using ion-cyclotron resonance frequency (1CRF) and neutral beam
(NB) heating, both individually and in combined situations, and indicates particulardevelopments and additions to JET that are planned to improve performance further
towards the reactor regime.

2. Technical Status of the Device
2.1. Machine Conditions
Machine conditions have been progressively improved during the operational phase andall machine systems have now met the stringent design specifications. The achieved
parameters are compared with the design values in Table I.
The toroidal magnetic field (B-f) now operates routinely at its maximum design valueof 3.45T. The plasma current (± <3%), horizontal plasma position (± < 10mm), plasma
elongation e(±5%) and shape are all controlled by feedback circuits acting on poloidal
field coils. Following considerable work on these systems, stable control has been obtained with elongations in the range 1.2-1.7. However, the plasma vertical position is naturally unstable due to both the quadrupole poloidal field necessary for the elongated plasma
and the de-stabilising effect of the iron magnetic circuit. Loss of vertical position feedback control at higher elongations can lead to large vertical forces on the vessel. Some
additional vessel support has been introduced, but the plasma current, Jp, had still to be
restricted within the operating regime given by /p2(e-1.2)<5.0(MA)2.
With a plasma elongation limited to 1.4, the plasma current has been increased to 5.IMA(for a period of 3s within a 20s pulse) exceeding the design value of 4.8MA. In addition,
plasma currents have been achieved routinely at 4MA with flat-tops of 6s duration. At
present, the full inductive flux (34Vs) is not used as the maximum premagnetisation current creates stray fields, which inhibit reliable plasma breakdown. Strengthening of thevacuum vessel, modifying the primary winding and improving the power supply will allow
an enhancement of the plasma current to 7MA, at the full elongation of t = 1.7. Stable
discharges with /P = 3.5MA, BT= 1.7T, e= 1.35 and qi = 2.2 have been studied to simulate
7MA operation.
The vacuum vessel is usually operated with wall temperatures at 250-300°C and witha base pressure of 10~7mbar H and 10~9mbar residual impurities. The vessel is conditioned
by glow discharge cleaning (GDC) in hydrogen and/or deuterium. To reduce the level ofmetallic impurities and oxygen, the torus walls are carbonised by glow discharge cleaning
in a mixture of hydrogen or deuterium and hydrogenic methane (CH4) [3].
In most recent experiments, eight carbon plasma limiters have been located symmetricallyon the outer equatorial plane of the vessel. Since disruptions mostly terminate on the innerwalls, these have been covered by carbon tiles to a height of ± lm around the mid-plane.
Similar tiles also protect the frames of the RF antennae, eight octant joints, and the outerwall from neutral beam shine-through. Additional tiles have been installed at the top and34

JET TEAMTABLE I. PRINCIPAL PARAMETERS OF JET:DESIGN AND ACHIEVED VALUES
Parameter
Plasma minor radius(horizontal), a
Plasma minor radius(vertical), b
Plasma major radius, RB
Plasma elongation ratio,€=b/aFlat-top pulse length
Toroidal magnetic field(plasma centre)
Plasma current:
circular plasma
Elongated plasma

Design Values

Achieved Values

1.25m

0.8—1.2 m

2.10m
2.96 m

0.8—2.1 m
2.5—3.4m

1.68
Up to 20 s

1.2—1.7
Up to 20s

3.45T

3.45T

3.2 MA
4.8 MA

3.0 MA
5.1 MA

34 V-s28 V sFlux Drive Capability
Additional heating power(in plasma)
RF Power
Neutral beam power

1986 Values

Full Values

8MW(3 antenna)

32MW(8 antenna)

<7MW

10 MW
(1 box)

20 MW(2 boxes)<9MW

bottom of the vessel to protect the vessel during X-Point (Separatrix) operation. The total
surface area covered is 45m 2 , corresponding to —20% of the vacuum vessel area. The inner wall tiles used as limiters and those for X-Point protection have provided powerful
pumping (with speeds up to lOOmbar.I.s"1). This has allowed operation at low density near
the plasma edge and was used to reduce the density after neutral injection to avoid disruptions. Recently, helium discharges prior to normal operation have improved the inner walltiles pumping capacity.
In early 1987, it is planned to extend the coverage in the machine and the existing limiters
will be replaced with two toroidal 'belt' limiters (surface area ~ 15m2), and the existing
RF antennae will be replaced with eight water-cooled models. In late 1987, the belt limiter
tiles and those from the eight antennae will be replaced by beryllium plates to investigate
the expected advantages with this low Z material.
2.2. RF Heating System
Since early 1985, three RF antennae have been installed at the outer equatorial wall. Poweris transferred to the plasma at a radiation frequency (25-55MHz) equal to the cyclotron
resonance of a minority ion species (H or He 3 ). Each antenna is fed by a tandem amplifier

IAEA-CN-47/A-I-2

35

delivering up to 3MW in matched conditions. The three units have been regularly operated
up to 7.2MW for 2s pulses. Experiments with 8s pulse duration have also been performeddelivering ~ 40MJ to the plasma. Recently, a fourth RF generator has been installed sothat two generators can be coupled to one antenna [4],When completed, the JET RF system will have eight generator-antenna units initiallydelivering 24MW for 20s pulses. Subsequently, each amplifier unit will be upgraded to
4MW, using a more powerful tetrode in the final amplification stage. New water-cooledantennae have been built and tested for insertion inside the toroidal belt limiter.
2.3. Neutral Beam System
A long pulse (~ 10s) neutral beam (NB) injector with eight beam sources and one integrated
beam line system has been operated on JET since early 1986. H beams have been injected
into D plasmas with particle energies (in the full energy fraction) of up to 65keV. The neutral
power fractions were 69%, 23% and 8% in the full, half and one-third energy components,
respectively, giving a total beam power of ~ 5.5MW injected into the torus. D beams havealso been injected into D plasmas, with particle energies up to 75keV (injected power fractions of 76%, 17% and 7%) giving a total power up to 10MW. Up to 40MJ have beendelivered to the plasma during a pulse. The second neutral injection box will be available
for operation on the machine in mid-1987.
2.4. Diagnostics
About 30 different diagnostics have been installed on JET, allowing cross measurementsof the main plasma parameters and detailed analysis of some of the plasma features suchas sawteeth oscillations and MHD behaviour [ 5 ] . The ECE diagnostic is the main instrument for providing electron temperature profiles and infra-red interferometry is the maintechnique for density profiles. The ion temperature is measured by three different techniques; charge-exchange neutrals, Doppler broadening of Ni 2 6 + and of resonant chargeexchange lines of light impurities. Flux measurements and spectrometry of 2.4MeV neutrons
are also used. Zef/is measured by visible bremsstrahlung and is in agreement with estimatesfrom neoclassical resistivity.

3. Experimental Results
3.1. Overall Plasma Parameters
Experiments have been performed in JET in different configurations: limiter discharges
on the inner or outer walls and X-Point discharges with single or double null. Various
heating scenarios have been used ranging from ohmic heating alone to combinations ofohmic, RF and NB. The variation of the main JET parameters is presented in Table II.
Several examples of typical pulses under these various conditions are indicated in Fig.l,
which show the density (n), electron temperature (Te) and safety factor (q) profiles, with
ion temperature (7)) measurements where available.
Each heating method has its own characteristic power deposition profile. For ICRF,
it is well localised (half width ~30cm),as shown in Ref. [ 4 ] , and can be varied acrossthe plasma by changing either the toroidal field or the frequency. The NB deposition profiles are usually less peaked on axis, and are quite flat at high densities.

36

JET TEAM

i

i

1

12 I

I P = 5MA

Bj = 3.45T

P, = 4 M W

TE = 0.70s

-

6

•5

-

-

P, = 17.9MWT,

• -4
• -3

n
T.

4

Pulse No. 10363
I P = 5MA
BT = 3.45T

(d)

10 8

•6

Pulse No. 10363

14 -(a)

^

••2

•1

2
0

•0

14

•(b)

12 -

(e)

I P = 3.5MA

BT = 1.7T

P, = 3.7MW

TE = 0.50 s

5
-

10 -

0

-+- -+- -+-Pulse No. 7772

Pulse No. 9775

7

I P = 3MA

BT = 3.4T

Pt = 8.7MW

TE = 0.32s

6

5 r

1

4

8
n
6

3 I

.

4

.-: T NT~ " "
0 T
14

I

I

I

"(c)

12

Pulse No. 10204

IP = 2MA

BT = 2.2T

P, = 8.7MW

TE = 0.35s

q*/

Te

/

BT = 3.3T(X-point)/

P, = 11MW TE = 0.60s

/
/

8 -

I P = 3MA

:.

10 -

6 _

7

Pulse No. 9603

/

-

-

s

4

^\.

n

2

T e "-\

0

i

2.8

3.0

3.2

i

i

1

1

4

3.4

3.6

3.8

4.0

4.2

Major Radius (m)

2.8

3.0

3.2

3.4

3.6

3.8

4.0

4.2

Major Radius (m)FIG. 1. Density (n), electron temperature (TJ and safety factor (q) as a Junction of major radius.
The ion temperature (Tj) is shown where available. These profiles are indicated for the following cases:(a) 5 MA pulse (No: 10363), ohmic heating only, BT = 3.45 T. (b) Low q pulse (No. 7772), ohmic
heating only, Ip = 3.5 MA, BT - 1.7 T. (c) Monster sawtooth pulse (No. 9603), P^ = 6.1 MW.
(d) 5 MA pulse (No. 10363) with 15.2 MW of RF + NB power, BT = 3.45 T. (e) High 7). pulse(No. 9775), Paid = 8MW. (f) X-point configuration pulse (No. 10204), Ip = 3 MA, P^ = 8.2 MW.

37

IAEA-CN-47/A-I-2

TABLE II. RANGE OF MAIN JET PLASMA PARAMETERS
Parameter

Value

Toroidal field (plasma centre) B,

1.7-3.4T

Plasma current, Ip
Flux safety factor at edge, q+
Volume averaged density, neCentral electron temperature, teCentral ion temperature, %Energy confinement time, TE

1-5 MA
2.2-16
0.5-5x10 l9 m- 3
2-7.5 keV
1.5-12keV
0.2-0.9 s

o NB+RFa NBcoD NBct* RF
x OH
p OH with Pellet

-eit

Otr

2
3
n e (1019/m3)

FIG. 2. Zeffas a function of average density for various plasma conditions.

38

JET TEAM
0.7

Ohmic and RF Density Limit(m-3) =1-2xio 2 0 B T (T)/qR(m)• O.H.
« RF
0.6 - x NB"

+ NB+RR

'•

^K.

\

o Pellets05• •x . x X x x

x^>'x

Vx

• •
•

0.4

•

• * * *
V^0« x0.3•• • • ,O^v^*• 0 0^. *° °iss x

X

0
x

»
0.2

t

0.1

• *<

v

*
o

c +(£-

* "x

"xx

OOO

X>tx

++++

* + x

^-—"^

+ +o+^--r^" ^ - - - ^ ^Density Limit320n c (m- ) =2x10 B T (T)/qR(m)

*
i

1

+ +

°

« , ^ "X X

0

0

x

x

2

i

3
4
n R/BT (x10'9m-aT-1)

i

i

5

6

FIG. 3. Normalized current (= (2irRIp/ABT) = 1/q) versus normalized density (— nR/BT).

The following parameters have been obtained under different conditions:(a) With ohmic heating, peak ion and electron temperatures of 3keV and 4keV, respectively, were achieved with a plasma density ~ 4 . 2 x 1019m~3 and energy confinement
times exceeding 0.8s;(b) With RF heating, the peak electron and ion temperatures have reached 5.5keV withpeak ion densities of 3.5xl0 l9 m" 3 . However, the confinement time has dropped toT£ = 0.3s;(c) Ion temperatures up to 6.5keV at mean densities of 3 x 1019m~3 have been producedwith neutral beam heating. At lower peak densities (~ 1.5 X 1019m"3), ion temperatures> 12.5keV have been observed (Fig.l(e)). Again, there was a degradation in confinement time down to 0.3-0.4s. In addition, with — 7MW of neutral beam injection, current drive of 0.4MA has been observed at a mean density of 2 x 10 l9 m" 3 ;(d) Combined RF and neutral beam heating has coupled power up to I5MW to the plasma
in a 5MA discharge: the plasma energy content was ~6MJ (Fig.l(d)).
The JET impurity behaviour is similar to that observed in other ungettered tokamaks
with graphite limiters [ 3 ] . The main impurities in ohmic and neutral beam heated plasmas
are carbon and oxygen and principally carbon at high electron density, ne. During ICRH,
moderate metal concentrations (0.1% ne) are responsible for ~20% of the total radiation
power losses. These can be prevented temporarily by carbonisation. Radiation power losses
represent 30-60% of the total heating power, Plot, during NBH and 40-70% of P,o, for
ICRH. Most of the radiated power is emitted from the edge plasma. ZeJy varies from 3.5
at low ne to 2.5 at high ne. With NB injection, Ze/f falls to values of 2-3 and with pellet
injection, it can fall to values close to 1 (see Fig.2).

39

IAEA-CN-47/A-I-2ii

1.05 sPulse No: 9603

T.(0) (keV)Power (MW)

6

7

8

9

10
11
Time (s)1213FIG. 4. Sawtooth oscillations of the central electron temperature, showing the effects ofNB and RF
additional heating.

3.2. Density Limits
The Hugill diagram for JET is shown in Fig.3, where the operating domain is limited bydisruptive instabilities. In ohmic plasmas, the density limit is nc (OH)(m~3)= 1.2 x 1020Br(T)/qR(m). This limit depends on plasma purity. In RF heated discharges, it is only slightly increased, possibly because the effect of the extra power is cancelled by an increase in
impurities. In neutral beam heated plasmas, the limit is substantially increased, as shownin Fig.3, to nf(NB)(m"3) = 2.0 x 1020B7<T)/qR(m). Switching off neutral beams at high density causes the plasma to disrupt, which indicates that the power input plays an importantrole in the disruption mechanism. Preliminary experiments with a single-shot pellet injector have also exceeded the OH density limit.
Density limit disruptions are always preceded by an increase in the impurity radiation
at the plasma edge [6]. This is seen to cause a contraction of the electron temperature
profile, which is followed by the growth of coherent MHD activity (principally m = 2, n = 1).
Alternative theoretical models predict that either the contraction of the temperature profile leads to an unstable current profile [6], or that the increased radiation losses at theq = 2 surface lead to a thermal instability [7], but it is possible that both of these effects

JET TEAM

40
1

6 -

i

1

i

1

(a)

1

1

a

•-•"•

I

• • • • •

"D

=*+fD
• ' <£> + §&ja. 3 ,

8- A _

4...•••'

+

.

D

D

m

a

a

.o%^

2

-•

iB =3MA- • • ' • • •BT =2.2T

o OH

+ NB
0

i

6 -

1

I

D+

LfTTTi • • •1

a RF+NB
1

(b)

4 -

SP+&J

2 -

-D

+

Ip =3MAo°BT =2.2T

o OH

+ NB
0

1

D RF+NB

1

1

i

1.2 "

(c)

0.8 a

<x>++i BntS-€l Ctf«> ++I2> 0.4 h-

D

^
+

D

Ip =3MAC

BT =2.2T

o OH

+ NB
r

0

1

i

1

i

I

1

D RF+NB

1

2
3
4
5
6
Total Power/Average Density (x10- 19 MW-m- 3 )7

8

FIG. 5. (a) Electron temperature at the centre, (b) electron temperatures at the inversion radius(q => 1), and (c) inversion radius as a Junction of power input per particle, for ohmic, RF, NB and combined heating cases (Ip = 3 MA; BT = 2.2 T).

may be involved [8]. Both models result in the growth of an m = 2, n= 1 island before
the disruption. Observations and theoretical considerations both show that the central
plasma density can be increased by deep fuelling.
3.3. Temperature Effects

Sawtooth oscillations occur in almost all JET discharges. As these instabilities, in general,
limit the peak ion and electron temperatures, and may affect, in certain cases, the global

41

IAEA-CN-47/A-I-2

14
12 . (a)
10

f 8
6

0
01

h-

• a

4
O

-H-

BT = 3.4T

1

O

2

OH

A

0

RF

Q RF+NB

+• NB

+
(b)

12
10

18
o

6

^

4
2
0
0

1

2
3
4
5
Total Power/Average Density (x10~19MW-nv3)

6

7

FIG. 6. (a) Electron temperature (Te) at the centre, and (b) ion temperature (J) at the centre, as afunction of power input per particle for various heating scenarios.

energy confinement times, they justify a careful analysis. With central deposition of additional power, sawteeth may develop large amplitudes (up to doubling the central electron
temperature) and long periods (up to 0.6s) [6]. In some circumstances,with combined NB
and RF, 'monster' sawteeth have been observed (see Fig.4). They are characterised by long
durations of 0.8 -1.2s and strong reduction of the low m, n numbers MHD activity. The
instability mechanism of the sawteeth is described by a model involving ideal MHD [8].
The peak electron temperature 7^(0) can be raised significantly above the value reachedin the ohmic phase as shown in Fig.5(a). The dispersion seen in 7^(0) is due to sawteething
and to variations in profile deposition of the additional power. By contrast, the electron
temperature at the inversion radius (q~ 1 surface) shows only a weak dependence withthe power input per particle (/#/) for a given plasma current and toroidal field (Fig.5(b),and no explicit density dependence. The radius of inversion is independent of P/n in therange shown (Fig. 5(c)).
The ion temperature behaviour appears quite different; the peak ion temperature 7/(0)(deduced from the X-ray crystal spectrometer) is plotted versus Pfn in Fig.6(b). Above
4 x 1019MW. m~3, the ion temperature can greatly exceed the electron temperature and canreach 12.5 keV in JET. A broadening of the ion temperature profile is also observed incases of off-axis ICRF heating [4].

42

JET TEAM

l.U

3.4T> BT>3T

\

(a)

S\
N

l p =3MAss. X-Point
0.8
A

''•
0.6

" S "9

-

a
'. D
e (S)

••••CP

• .

D

0.4

lp=5MAr ..

'•••••. B f e b

1

. "'4

-

'

•

H>.

D.flp

D

-.

•

Q DO

-£k

- rr -l p =4MA

2 ii

° '---....

•

"' \

0.2 -

*

•a..

•aa

a

=2 MA

X
•

-

%

-

x

•

xxx

l p =3MA
\

0

a

MA

1
1

1

1

1

(b)
6t

...-• -"i
s

1

it''
-

•

.

.

*

-

•

•

-

•

•

'

*»'

• o 1 "
Op.

pi

tne

••-•'•••••
*

:

•

• •

a

'

a

i

OT 2
Q.

" '"s

_

.

. . » • * • » - ' " '

.

.

.

.

.

•

•

•

•

'

;

-

-

"

n
'

^

*
a

l p =5MAl p =4MA(X-Point»x

l p =2MA
L=1MA^• • # » " "

•m
I*"

0

•

*

3

-•

#

1

i

6
8
10
Total Power P, (MW)

I

12

14

16

FIG. 7. (a) Energy confinement time TE as a Junction of additional heating power (at Ip = 1, 2, 3, 4
and 5 MA; 3.4 T & BT a 3.0 T. (b) Plasma energy content Wk versus total input power P, with NBI
and ICRF heating (Ip = 1, 2, 4 and 5 MA; 3.4 T > BT 2: 3.0 T). X-point values at lp = 3 MA are
shown for comparison.

43

IAEA-CN-47/A-I-2

0

1

2

3
Major Radius (m)4FIG. 8. Single-null X-point configuration at I. = 2.0 MA.

3.4. Confinement Degradation
The definition of the total energy confinement time used at JET is T £ = W * / [ P / - dW</dt],
where W* is the kinetic energy and P, is the total input power to the plasma without
subtracting radiation losses. Reported values of rE are quasi-stationary.
With additional heating, the confinement time, TE, degrades with increasing input power(Fig.7(a)), as seen in a number of other experiments. The degradation is independent oftype of heating, whether RF, NB or combined. The rate of increase in W* with P,(= AWVAP/)
appears to reach a limit of 0.1-0.3MJ/MW( = s) at high powers (see Fig.7(b)). This suggests
a lower limit to the global confinement time, rE, of 0.1-0.3s in JET, independent of typeof additional heating. Confinement time depends weakly on plasma density but scales
favourably with plasma current.
Since the plasma energy is a function of n,-, Te and 7], the degradation in confinement
time is consistent with the observation that the electron temperature, at the inversion radius(q=1 surface) increases little with power input and only the central temperature (7;(0)) shows
an increase dependent on the central input power (see Fig.7(a) and (b)).
3.5. X-Point Regime
A better confinement regime with additionally heated plasmas has been observed (H-mode)in some Tokamaks with magnetic limiters or divertors. Stable discharges with magnetic
separatrix (or X-Point) inside the vessel have been maintained in JET for several seconds,

44
TABLE III.

JET TEAM
MAXIMUM VALUES OF (ni(0)TETi(0))
Peakdensity

Ion
Energyconfinement temperature
time

Experimental
programme

Ti(0)

nKO)
(10

19

Fusionparameter

3

m" )

QDT

equivalent

Plasmacurrent

QDT
3

(s)

(keV)

(10"nT -s-keV)

(MA)

Ohmic
(4.6 MW)

4.2

0.8

3.0

10

0.010

5

ICRF (7 MW)

3.7

0.3

5.4

6

0.012

3

NBI (6 MW)High n.
Low n;

4.4
1.5

0.4
0.4

4.0
10

7
6

0.10 a0.20 a

3
3

5.0

0.4

3.5

7

0.10 a

3

5

0.65

6

20

0.15 a3

Combined
NBI + RF
(14 MW)
X-point(NB 10 MW)a

Beam-plasma reactions are dominant.

at plasma currents up to 3.0MA with a single null (Fig.8) and up to 2.5MA with a double
null [9]. While interaction of the discharges with the limiters was curtailed, localised power
deposition on the top and bottom target plates has limited the total input power to 10MW.
The main objectives of these experiments are (i) to study the conditions of creation of ahigh density highly radiative region at the X-point capable of screening and isolating themain bulk plasma and (ii) to study the global confinement characteristics in comparison
with limiter discharges. Evidence exists of the formation of a dense plasma near the XPoint, with average density 1 - 2 X 1020m " 3 , corresponding to an order of magnitude higher
than the average value of the main plasma density. There is also evidence that up to 40%of the total input power could be radiated by a ~ 20cm x 20cm toroidal annulus aroundthe X-Point in the high density discharges with additional power.
As shown in Fig.7, the increase in total energy content (and hence global confinementtime) in single-null configurations is observed to be about twice that observed with limiters.
In these conditions, both average electron and ion temperatures appear to be higher, andthere are indications of changes in edge plasma suggesting an H-mode.

IAEA-CN-47/A-I-2

45

3.6. Slide-away Discharges
A preliminary study of slide-away discharges has also started with a view to decouplingcurrents and electron temperature. In these conditions, a plasma with 3MA and a meandensity just below 1019m~3 has been obtained with a good confinement time for these
conditions (~ 0.6-0.7s). The voltage per turn was 0.2V. When trying to increase the density
by gas puffing, the discharge reverted to normal conditions.
3.7. Progress Towards Breakeven
Neutron yields > 2 . 8 x l 0 l 5 s ~ ' have been obtained with D injection, mainly from beamplasma D-D reactions. The best ratio of fusion power to input power obtained was
Q D D =4xl0~ 4 , which is equivalent to Qor~0.2 and would have corresponded to a fusion
power production of - 2 M W (see Table II).
Fusion products <n,- TET-,> = 102Om"3-s-keV have been achieved in JET (compared withthe value 5 x 1021m~3s-keV required in a reactor) in an ohmic discharge at 5MA and a valueof 2xl0 2O m"-s-keV has been obtained in an X-Point discharge at 3MA current. However,additional heating has not yet been optimized into 5MA discharges. Under optimum
conditions at lower currents, the values are similar for ohmic heating only, RF, NB, andcombined heating cases, the degradation in rE with additional heating offsetting gains inother parameters (see Table III).

4. Progress in Understanding Heat Transport
JET data have been used to assess the status of theoretical understanding of heat transport[10]. Local fluxes have been derived from measurements and compared with presently
available theoretical values. No acceptable agreement has been found. However, the scalingof global energy confinement time with the dimensionless plasma physics parameters suggeststhat the underlying heat transport is a consequence of resistive MHD instabilities [11].
The topology of the magnetic field including chaotic regions and small islands could accountfor the heat fluxes observed in JET [12]. Such a model is also consistent with the existenceof an H-mode.
As a-particles mostly heat the electrons, it is imperative to obtain better control of theirtemperature, especially inside the q = l surface (ie sawteeth). A possible approach to increase
Te(0) is to control the profile of current and achieve a flat current distribution ( q s 1, q' =0)
in the central region, in order to stabilize or at least delay the sawtooth crash significantly.

5. The JET Strategy and Future Programme
The JET strategy is to optimize the fusion product <M,7]T £ .>. Attempts will be made tomaintain T£ with additional heating near to present values, and to increase the central density
and temperature.
In order to obtain higher levels of performance, additional equipment is required on JET.
Fig.9 shows the plasma profiles which are envisaged. They correspond to an energy content
of ~20MJ and would produce -10MW of a-power. If T £ ~0.3-0.4S, Qcrshould reach 1 —2.

46

JET TEAM
Low shearzone

Shear zone

1.0
FIG. 9. Density (n), temperature (T) and safety factor (q) profiles aimedfor according to the latest JET
strategy. The shaded area represents the zone q = 2.

Such plasma parameters could be obtained in the following ways:

(i)

(ii)

Sawteeth oscillations should be stabilized or have periods of several confinement times.
This should be achieved by flat q profiles ( q > l , q' = 0) over a large plasma crosssection, working with a low q boundary and driving part of the current (~2-3MA)in the outer plasma by Lower Hybrid Current Drive (LHCD) (~ 10MW) and by NB
injection. The ICRF antennae with proper phase control could also be used to launcha travelling wave and provide some additional localised current;
Radiation cooling between the q = 2 surface and the plasma boundary would be
minimised to allow operation below the major disruption limit and to provide highTe on the edge. Low Z material (C or Be) would be used to take advantage of its low
ionisation potential;(iii) Since the temperature at the q ~ 1 surface is higher when the density is lowered, edge
pumping will be used. Further advantages of low densities in the outer region would
be the prevention of disruptions and higher efficiency for LHCD;(iv) The D-T density on axis would be increased up to l-2xlO2Om~3 by high velocity pellet
injection (>5km/s);(v)

High central temperature (~12-15keV) would be achieved by a combination of 'on
axis' ICRF heating, NBI at 160keV and current profile control.

IAEA-CN-47/A-I-2

47

This scheme will be tested in two configurations: (a) with a magnetic separatrix at 4MA
and (b) with limiters at current values up to 7MA. Such a low-q discharge has been simulated
in JET (see Fig.l(b)).
The main aim of JET operation during the next phase will be to reach maximum
performance with full additional heating power in deuterium plasmas, before proceeding
to the final phase requiring the introduction of tritium. The starting date for D-T operations
is linked to the performance reached in the previous phase and is expected in 1991. aparticle production must be sufficient to analyse its effects on the plasma and to answer
the question whether a-particle heating degrades energy confinement as other heating
methods. Although degradation is foreseen, this has direct consequences on the ignition
margin and on the size of future devices.

6. Implications for a Reactor
Present results suggest that a way for a future reactor is to rely upon tight control of current
and density profiles and intense heating inside the q = l surface to benefit from improved
energy confinement and high central temperature (as proposed for JET). Alternatively, asize and current increase could lead to the required ignition temperature, without additional
heating. The resulting simplification of the overall system could well compensate for thelarger size. Scaling [13] suggests that in a Tokamak with a major radius of ~ x2.5 thatof JET, plasma volume of - 2000m3 and magnetic field slightly higher than in JET, a 15keV
plasma with a confinement time of 15-25s could be reached with ohmic heating alone ata peak density - 2x 1019nT3, satisfying the ignition criterion. Then, the thermonuclear power
output of the D-T burner could be increased by injecting fuel pellets: the thermal insulation
should degrade as the additional heating provided by the a-particles increases without astrong variation of temperature. Burn control would be performed entirely through density
control.

7. Conclusions
The following conclusions can be drawn at this stage:(a)Both ICRF and neutral beam (NB) injection methods are effective in delivering power
into the JET plasma with the expected profile depositions. Large central electron andion temperature increases have been observed; te and 7j have both reached values of
7.5keV at densities of 3xl0 19 m" 3 . At lower density (~ 1.5xl0 19 m- 3 ), 7]= 12.5keV has
been measured;(b)

For both heating methods, plasma energy has increased with input power. About 6MJ
energy has been achieved in JET plasmas with ~ 18MW of total power input. However,
the rate of energy increase with power^(AWyAP,) ~0.1-0.3s,is smaller than thecorresponding ohmic confinement times (T £ ~0.6-0.8S). The fusion product <nirj]>
has reached values in excess of 102°m~3.s:keV, both with ohmic heating at 5MA andwith additional heating; the degradation in confinement offsets the gains in density
and temperature;(c)

A higher fusion product value of ~2xl02°m~3<s-keV has been obtained in X-point
configuration at lower current (~3MA). Values of r £ = 0.6-0.7s and 7} up lOkeV wereachieved with neutron production ~2xl0 15 ns~ 1 , equivalent to a thermal Qflr=0.1, thetotal QDT being -0.15;

48

JET TEAM(d)

In the central region (inside the inversion radius) between sawtooth collapses, additional
heating is effective in increasing the electron temperature of the plasma (cf monster
sawteeth);(e)Outside this region, electron heating is poor. The electrons seem to be the main energy
loss channel, which appears related to confinement properties and not to the heatingprocess;(f)

Good prospects exist on JET for production of several MW of a-particle power at
a value QDT** 1;(g)a-particle heating is expected to behave in a similar fashion to other heating methods.
Therefore, a reactor must either: (i) work at moderate currents with sophisticatedcontrol of the central region; or (ii) work at high currents without the need for complex
control and additional heating.

REFERENCES[1]HUGUET, M., et al., in Fusion Technology (SOFT) (Proc. 13th Symp. Varese, 1984), Vol. 1,
Commission of the European Communities, Brussels (1984) 91.
[2] LALLIA, P.P., et al., Plasma Phys. Controll. Fusion 28 (1986) 55.
[3] BEHRINGER, K., et al., Paper IAEA-CN-47/A-IV-l, these Proceedings, Vol. 1.
[4] JACQUINOT, J., et al., Paper IAEA-CN-47/F-I-1, these Proceedings, Vol. 1.
[5] JET TEAM, Progress Rep. EUR-JET-PR-3 (EUR-10616.EN), Commission of the European
Communities, Abingdon (1985).
[6] CAMPBELL, D.J., et al., Paper IAEA-CN-47/A-VII-5, these Proceedings, Vol. 1.
[7] REBUT, P.H., HUGON, M., in Plasma Physics and Controlled Nuclear Fusion Research 1984
(Proc. 10th Int. Conf. London, 1984), Vol. 2, IAEA, Vienna (1985) 197.
[8] WESSON, J.A., et al., Paper IAEA-CN-47/E-I-l-l, these Proceedings, Vol. 1.
[9] TANGA, A., et al., Paper IAEA-CN^7/A-M, these Proceedings, Vol. 1.
[10] DtiCHS, D.F., et al., Paper IAEA-CN^7/A-VI-1-1, these Proceedings, Vol. 1.
[11] CORDEY, J.G., et al., Paper IAEA-CN-47/A-II-3, these Proceedings, Vol. 1.
[12] REBUT, P.H., et al., Paper IAEA-CN-47/E-III-4, these Proceedings, Vol. 2.
[13] THOMSON, K., et al., in Controlled Fusion and Plasma Heating (Proc. 13th Europ. Conf.
Schliersee, 1986), Vol. 10c, Pt 1, European Physical Society (1986) 29.

DISCUSSION
R.J. GOLDSTON: In your presentation you mentioned the 'hot ion regime' at
Ip = 3 MA, which had r E = 0.6 s, for Pe ~ 9 MW. Could you say how this fits
into the broader set of data you have for X-point and limiter discharges?
P.H. REBUT: In the case of a pure hot ion mode the confinement time (pulse
No. 9775 in Fig. l(e)) is only 0.32 s. With X-point and H-mode operation, 0.6 s canbe reached with a relatively high ion temperature of about 10 keV in the centre. Thisset of results does fit into the general JET data set.
H. HORA: What was the highest number of fusion neutrons you measured andwhat was the input energy, i.e. neutral beam plus tokamak operation?

IAEA-CN-47/A-I-2

49

P.H. REBUT: The highest number of neutrons per second emitted from JET
is 2.8 X 1015. The injected power was 8 MW, into a plasma limited on the inner
wall.
R.R. WEYNANTS: With regard to the X-point operation, can you commenton, first, the OH plasma confinement, and, second, the existence of a power
threshold?
P.H. REBUT: A minimum power of 6-8 MW was needed to obtain good confinement properties in the X-point mode of operation, which is a gain of up to a factorof two. This corresponds to a different plasma regime at the edge, near the separatrix.
If this edge regime is not achieved, there is little difference between an X-point configuration and a 'standard' one.
R.J. TAYLOR: For the high ion temperature mode, the electron energy confinement time must be considered separately. What is this time in the case of JET?
P.H. REBUT: I agree that the electron energy confinement time is dominant fora fusion reactor, but without a full profile of the ion temperature and without knowing
the ion transport in detail it is difficult to consider the electron confinement time
separately. It may be estimated from Fig. 1, taking into account that half of the power
might perhaps be transported directly by the ions.
D. OVERSKEI: You indicated that rE improved with increasing plasma current and with the use of a single-null diverted magnetic configuration. Could you say
whether this increase in TE is due predominantly to an increase in Ti(0) rather thanto better confinement of ions?
P.H. REBUT: The increase in confinement is mainly due to a widening of the
profiles and to a pedestal in the electron temperature.
B. COPPI: What values of the streaming parameters did you attain in order toobserve the slide-away regime?
P.H. REBUT: We did not make any measurement of the electron temperature.
The current was 3 MA and the central density was close to 10" m~3.

IAEA-CN-47/A-I-3

TFTR PLASMA REGIMES
R.J. HAWRYLUK,
V. ARUNASALAM,
M.G. BELL,
M. BITTER,
W.R. BLANCHARD, N.L. BRETZ, R. BUDNY, C.E. BUSH1, J.D. CALLEN 2 ,S.A. COHEN, S.K. COMBS1, S.L. DAVIS, D.L. DIMOCK, H.F. DYLLA,
P.C. EFTHIMION, L.C. EMERSON1, A.C. ENGLAND 1 , H.P. EUBANK,
R.J. FONCK,E. FREDRICKSON,
H.P. FURTH,
G. GAMMEL,
R.J. GOLDSTON,
B. GREK,
L.R. GRISHAM,
G. HAMMETT,
W.W. HEIDBRINK3, H.W. HENDEL 4 , K.W. HILL, E. HINNOV, S. HIROE1,
H. HSUAN,
R.A. HULSE, K.P. JAEHNIG,
D. JASSBY,
F.C. JOBES,
D.W. JOHNSON,
L.C. JOHNSON,
R. KAITA,
R. KAMPERSCHROER,
S.M. KAYE, S.J. KILPATRICK, R.J. KNIZE, H. KUGEL, P.H. LaMARCHE,
B. LeBLANC, R. LITTLE, C.H. MA 1 , D.M. MANOS, D.K. MANSFIELD,
R.T. McCANN,
M.P. McCARTHY,
D.C. McCUNE,
K. McGUIRE,
D.H. McNEILL,
D.M. MEADE,
S.S. MEDLEY,
D.R. MIKKELSEN,
S.L. MILORA1,
W. MORRIS5,
D. MUELLER,
V. MUKHOVATOV6,7
8
E.B. NIESCHMIDT ,J. O'ROURKE ,D.K. OWENS,
H. PARK,
N. POMPHREY,
B. PRICHARD,
A.T. RAMSEY,
M.H. REDI,
A.L. ROQUEMORE, P.H. RUTHERFORD, N.R. SAUTHOFF, G. SCHILLING,
J. SCHIVELL, G.L. SCHMIDT, S.D. SCOTT, S. SESNIC, J.C. SINNIS,
F.J. STAUFFER9,
B.C. STRATTON,
G.D. TAIT,
G. TAYLOR,
J.R. TIMBERLAKE, H.H. TOWNER, M. ULRICKSON, V. VERSHKOV6,
S. VONGOELER,
F.WAGNER 10 ,R. WIELAND,
J.B. WILGEN1,
M. WILLIAMS,
K.L. WONG,
S. YOSHIKAWA,
R. YOSHINO",
K.M. YOUNG, M.C. ZARNSTORFF, V.S. ZAVERYAEV6, S.J. ZWEBEN

Princeton Plasma Physics Laboratory,
Princeton University,
Princeton, New Jersey,
United States of America
Abstract
TFTR PLASMA REGIMES.
Significant extensions in the TFTR plasma operating regimes have been achieved with additional
heating system capability, installation of a multishot pellet injector, and the development of an enhanced
confinement regime. In ohmically heated pellet-fuelled discharges characterized by highly peaked' Oak Ridge National Laboratory, Oak Ridge, TN, USA.
University of Wisconsin, Madison, WI, USA.
3
GA Technologies, Inc., San Diego, CA, USA.
4
RCA David Sarnoff Research Centre, Princeton, NJ, USA.
5
Balliol College, University of Oxford, Oxford, UK.
6
Kurchatov Institute of Atomic Energy, Moscow, USSR.
7
EG&G Idaho, Inc., Idaho Falls, ID, USA.
8
JET Joint Undertaking, Abingdon, Oxfordshire, UK.
9
University of Maryland, College Park, MD, USA.
10
Max-Planck-Institut fur Plasmaphysik, Garching, Fed. Rep. Germany.
11
JAERI, Naka-machi, Naka-gun, Ibaraki-ken, Japan.
2

51

52

HAWRYLUK et al.

density profiles, enhancements in TE have resulted in ne(0)rE(a) values of 1.5 x 1020 m" 3 -s. In neutralbeam heated discharges, an operating regime has been developed in which substantial improvements inenergy confinement time and neutron source strength are observed. Ion temperatures ofabout 20 keV and n^T^Tfi))
values of 2 x 1020 m~ 3 -S'keV have been achieved. This enhanced
confinement regime is characterized by high values of (3p and low values of collisionality. The
observed surface voltage, which is negative during beam injection, is compared with models including
beam driven and bootstrap currents.

I.

Introduction

The goals of the TFTR project are (1) to study reactor-grade
plasmas with temperatures of order 10 keV and densities of order
10
m , and (2) to achieve approximate breakeven between the
power input to and the fusion output from the plasma (Q - 1).
During the past year, high-power neutral-beam-heating experiments
and ohmic-heating experiments utilizing a deuterium pellet
injector have significantly extended the operating regimes of
TFTR, as measured by the n (0)Tg(a)Ti(0) product and the Q-value
achieved in neutral-beam-heated discharges.
This paper gives a brief description of the TFTR device
status and the operating regimes in gas-fueled and pellet-fueled
ohmic discharges.
A more extensive discussion of the pelletfueled discharges is given by Schmidt et al. [1]. Results fromneutral-beam-heating experiments in the standard regime are also
summarized briefly.
A general description is given of theoperatingcharacteristics
of discharges
in the enhanced
confinement regime; further information is contained in the papers
by Goldston et al.[2], McGuire et al.[3], and Hill et al. [4].
II.

Machine Status

An initial series of experiments with two (co-injecting)neutral beamlines was completed in April 1985, during which TFTR
reached its original machine design specifications for plasma
current and toroidal field (I = 2.5 MA and B T = 5.2 T) [5].
Subsequently, two additional beamlines were installed; threebeamlines are now aimed tangentially in the direction of theplasma current (co-injection), and one opposite (counterinjection). During the recent experiments, the maximum injected
power was 20 MW with a full energy component of ~ 10 MW. The
maximum beam voltage was 110 kV and the maximum pulse duration
1.0 s. Most of the experiments were conducted with 0.5 s pulse
lengths and an average beam voltage of ~ 95 keV. Deuterium
injection was used in all of the experiments discussed here.
During the last beamline installation, a large-area axisymmetricinner wall limiter was also installed. This limiter is composedof water-cooled Inconel plates covered with graphite tiles. The
toroidal inner wall limiter and moveable limiter define thepresent maximum plasma dimensions as R = 2.48 m and a = 0.82 m.

53

IAEA-CN-47/A-I-3
1

PELLETFUELED

4.7

1.4
1.6
2.2
1.8

*D2
T

Ip(MA)

D2

0.6

b

'l

4.8

4.8J

MovableLimiler

JX

-

-

m

•

m
1

%

04

X

?,'f,

0

•

V
T

-

-

a

FUEL!

\J.C.

o 02

IAEA
(1984)_BT(T)lMoraWe

i.e
2.2

* He
o D2

n

i

]

1

|

1

1.0

0.5
_

2

_

3.9 '

|

1.5

I

2.0

21

n e q R a (10 )F/G. 1. Gross energy confinement versus neqjl a for ohmically heated discharges with and withoutpellet injection. The 1.6 MA discharges had reduced aperture, a = 0.7 m, compared with the otherJull-size 0.8 m discharges.

III.

Ohmically Heated Discharges

Initial ohmic experiments reported in 1984 [6] at modest
toroidal fields (< 2.8 T ) , plasma currents (< 1.H MA), anddensities (n < 3 x 10'' m~3) demonstrated that the global energy
confinement cime, Tg, scales consistent with neqR a, reaching amaximum value of tg ~ 0.3 s.
Subsequent experiments utilizing
both gas- and pellet-fueled discharges have concentrated on
exploring the applicability of this scaling law over a wider
operating range in density, toroidal field, and plasma current.
The analysis of the energy confinement time has reliedprincipally upon a time-independent kinetic analysis code, SNAP,[5] and has been supplemented by magnetic measurements [7].
Figure 1 is a summary of the ohmic-heating studies for full bore
plasmas. In gas-fueled deuterium discharges, the confinement timeincreases up to 0.44 s. in reasonable agreement with n q sealing
for n g < 4.8 x 10'' m"^. At higher densities, saturation is foundto occurfor both helium-gas and deuterium-pellet fueleddischarges. In the high density regime, the confinement time isobserved to be a weak function of plasma current.
In addition to increasing the line-averaged density, pellet
injection produces highly peaked density profiles.
A lineaveraged density of 1.4 x 10 2 0 m~^ has been achieved 200 ms after
the injection of five 2.7 mm pellets in experiments conducted on

HAWRYLUK et al.

54
2.0
•

2.2 MAo

1.4 MA

1.5

Fe = 3.4-4.3xl0'V 3
P T0T (MW)1015

FIG. 2. Variation of the stored energy for 1.4 and 2.2 MA power scans, for a constrained density range
and BT = 4.8 T, R — 2.48 m, a = 0.82 m. Magnetic measurements are compared with prediction of
the Goldston L-mode model.

the inner graphite limiter.
This corresponds to a Murakami
parameter, n R/BT = 6.5 x 10'' m
T . The central electron
density was 1 8 x 10 ^ m~-* and central electron temperature was
1.4 keV, as measured by Thomson scattering.
This peaked density
profile was achieved by operating with a reduced plasma minor
radius of 0.7 m and a plasma current of 1.6 MA, in order toimprove pellet penetration.
The energy confinement time was~0.50 s according to kinetic and diamagnetic measurements
corresponding to an n e (0)r E = 1.4 x 10 2 0 m"^>s.
The energy
confinement time is ~ 20$ greater than that achieved in gas-fueleddischarges with the same values of I , R, and a, using either
deuterium or helium as the working gas.
IV.

Neutral-Beam-Heating Studies in the Standard Regime

The variation of energy confinement time with injection
power up to ~ 15 MW was studied systematically. These experiments
were conducted in the large plasma configuration on the inner wall
limiter with deuterium gas fueling, deuterium beams, I = 1.4 and
2.2 MA, and B T = 4.8 T [7]. Figure 2 shows the variation in the
total stored energy for the 1.4 and 2.2 MA power scan with aconstraint on the density at the end of injection. Kinetic andmagnetid measurements (using the techniques described in Ref. 7)
are in good agreement. The stored energy increases linearly withheating power; however, the rate of increase of stored energy,dW D /dP^ ea £, is appreciably less than the ohmic confinement time.
The gross energy confinement time t E (a), defined as W,
dW,r,QT/dt), where W T Q J = W e + W, + W.

data shown in Fig. 2, can be fitted to ais the "incremental" confinement time.

y> - P 0H * > J ** t h *orm a + S/P-pQ^, where a
In the plasma density

IAEA-CN-47/A-I-3

55

range of 3.4 < n e < 4.3 x lO 1 ^ m~^, a increases with plasma
current from 61 ms to 94 ms as the current is increased from 1.4
to 2.2 MA.
The energy confinement time is observed to depend
weakly upon density; however, a negative scaling of o withincreased density is observed in the 1.4 MA discharges. At low
density, the enhancement in stored energy at high power occurs inthe beam and thermal ions, since the electron stored energy
decreases slowly with decreasing density.
The high-power beam
heating in this current and density range results are also in fairagreement with the Goldston L-mode model [8] for hydrogen
injection into deuterium plasma. Powerjlaw fits to the data shown
for P
in Fig. 2 indicate that W T Q T « P-rOT
V
TOT > 4 m
V.

Neutral Beam Heating in the Enhanced-Confinement Regime

Previous operation of TFTR at low I (0.4-1.0 MA) andmoderate beam power (Pb < 6 MW) using only co-injection £5]
allowed access to a very low-density regime (n ~ 1 x 10^" m ~ ' ) ,characterized by high values of ion temperature (- 9 keV) and
rapid toroidal rotation velocity (up to ~ 7 x 10^ m/s). Recent
experiments in this regime conducted at higher power (< 20 MW),using both co- and counter-injection, have demonstrated enhanced
confinement relative to the predictions of L-mode scaling, alongwith central ion temperatures of ~ 20 keV at n - 3 x 10 1 ' m"^with a central n e (0) ~ 7 x 10 1 ' m~3 [9]. The achievement of thisimproved performance relative to the previous experiments is due
to extensive degassing of the limiter and near-balanced injection(P

ctr ~ PcO*-

A procedure that results in a low recycling rate, as shown bydecreases in the D a emission and edge neutral pressure, isidentified to be critical in giving rise to the enhanced
confinement regime [4,10]. Both low-density deuterium and helium
discharges have been used to degas the inner wall. These highpower discharges (0.8-1.8 MA) were not fueled following
breakdown. Helium discharges were more effective in reducing the
low-density limit and the decay time of the density following adiagnostic gas puff. The density decay time has been decreased
from > 5 s to as low as = 0.15 s after extensive degassing. The
low initial density of the target plasma and relatively high Ze~p
- 6 is a consequence of the degassing procedure which effectively
removes D + from the target plasma. The large density rise during
injection [ne(final)/n"e(prior to injection) < 3-4] is accompanied
by a decrease in ZQff to - 3 at the end of injection.
Figure 3 shows a comparison of the temporal evolution of two
discharges during neutral beam injection with different initialline-integral densities, different deuterium recycling rates due
to different degassing histories, and comparable final lineintegral densities.
In the discharge with the lower initial
density, both the neutron source strength and the stored energy

56

HAWRYLUK et al.

4.5
TIME (sec)

FIG. 3. Comparison of the evolution of line average density, neutron source strength, Ha emission(which includes both Ha and Da emission) and stored energy for two discharges. The solid curve
corresponds to a supershot. R = 2.48 m, a = 0.82 m, Pb = 12.5 MWandBT = 4.7 Tin the supershot
discharge, whereas Pb = 11.4 MW and BT = 5.0 T in the other shot.

increase throughout the 0.5 s neutral-beam pulse. The global
energy confinement time is also substantially longer. Discharges
that exhibit continually rising stored energy and neutron source
strength during the half-second beam-injection have been
colloquially termed "supershots". Experiments conducted in the
final days of the experimental run demonstrated continued increasein stored energy with 0.7 s duration beam pulses and that thestored energy approached equilibrium with 1.0 s duration pulses.
At plasma currents up to 1 MA, values of & = 2.0 from plasma
diamagnetism have been measured. Maximum values of e up to 2.2
have been achieved transiently. These high values of gl result in
substantial outward shifts of the magnetic axis (Rn,ag-Rgeo^/a ~0.1 and distortions of the plasma shape. The vertical %loTigation,
K, decreases from 1.05 to < 0.9 in the most extreme cases. Figure
4 shows a comparison of the electron temperature and density
profiles measured near the end of the pulse for the discharges
shown in Fig. 3. The discharge with the lower initial density(supershot) has a more peaked density profile, with a higher
central density and a much broader electron temperature profile.
Peak central electron temperatures of ~ 6.5 keV during injection
have been obtained in supershots.

57

IAEA-CN-47/A-I-3

at

2.0

2.5
MAJOR RADIUS ( m )

3.0

2.0

2.5
3.0
MAJOR RADIUS ( m )

3.5

E
O

1.5

FIG. 4. Comparison of Thomson scattering measurements of the electron temperature and density
profiles at 4.45 s for the discharges shown in Fig. 3. The solid points are for the supershot discharge
and the open circles are for the other shot.

The central ion temperature is measured using several
techniques.
X-ray Doppler-broadening measurements have been
carried out for Fe XXV and Ni XXVII K a lines using a horizontally
viewing crystal spectrometer and three vertically viewing
spectrometers. The horizontal channel is rendered ineffective,due
to neutron noise during supershots.
In the analysis of thevertical detectors the large shift in magnetic axis needs to betaken into account. For the data shown in Fig. 5, the emitting
region viewed by the vertical detector is - 0.2 m outside of themagnetic axis. Doppler-broadening measurements in the VUV havealso been used to obtain central ion temperatures.
The

58

HAWRYLUK et al.

no

20 -

~

15 -

i
—r~
—1
1
[p=2.2M
FeXEI.85A(Hor.).
Ip= 1.4M A oj
FeXSZI.85A(Vert.) A
Ip=0.6FeXXE 192 A
MA
4X
i BjA#A^"10 -

L

•a a

AQ°*

°

1

1—

•

k

A

4

>t * i

**A

* A *

A

5
11 G
n

-

o^ 8
I

1

1

1

1

1

,/nP(l0''Vm3)
F/G. 5. Ion temperature versus PINJ/nc. Closed symbols are for 4.8 < BT < 5.2 T, 0.8 < Ip
< 1.0 MA, R = 2.45 m and a = 0.80 m, and open symbols are for the discharges shown in Fig. 2.

spectrometer has a horizontal radial view of the plasma. Onedimensional impurity transport simulations including chargeexchange recombination indicate that the mean radius of theemission profile for the lithium-like ions used is ~ 0.2 m. The
heatingbeams
have
permitted
preliminary
spectroscopic
measurements utilizing carbon lines excited by charge-exchange
recombination. Good agreement between the various measurements isachieved.
Figure 5 shows the variation in central ion temperature fordischarges in the standard neutral-beam-heating regime (I = 1.1
and 2.2 MA) and for both super and non-super low-i-urrent
discharges (0.6 < I < 1.1 MA). In the low-current discharges,
considerable variability in the heating results is obtained,reflecting the occurrence of MHD activity, variations in limiter
history, and variations in the fraction of power in the codirection. Taking into account uncertainties in the measurements
and the calculated differences between the impurity and hydrogenic
ion temperatures which are < 2 keV, these results indicate centralbulk ion temperatures of ~ 20 keV. The maximum value of the ion
heating efficiency parameter, n e A T J / P ^ J , is ~ 3.8 x 10 1 3 keV-W"1m~3. This is in comparison with previous experiments [5] in which
figATj/Pjuj was - 1.5 x 10 1 3 keV-W 1 . m"3.
Figure 6 shows the scaling of stored energy with power. An
important characteristic of the enhanced confinement regime is thefraction of power in the co-direction, P C Q / P I N J - A S s n o w n i n F i 8 6, at a given power the highest stored energy (and hence T^) is59

IAEA-CN-47/A-I-3,(MA) P
/P
r
C0

INJ

0.9 0.5-0.6
0.9
1.0
0.9 0.63-0.7
1.0 0.7-0.8

L-Mode' 0 . 9 MA
10

15

20

PINJ(MW)

FIG. 6. Variation of stored energy with injected power for different values of fractional power in the
co-direction, for 4.7 < BT < 5.2 T, 2.40 < R < 2.48 m and 0.76 < a < 0.82 m. Magnetic
measurements are compared with the prediction of the L-mode model of Goldston.

achieved with nearly balanced injection (pr(/PINJ ~ ° * ^ indischarges with degassed limiters. In Fig. b, the data set wasconstrained to eliminate discharges with very-large-amplitudecontinuous MHD activity [3]. With nearly balanced injection thestored energy increases linearly with heating power and theincremental confinement time is 0.15 s, comparable to the lowdensity ohmic confinement time prior to injection. Furthermore,with near balanced injection the global confinement time is notobserved to degrade with power: values up to 0.17 s are achieved
with 12 MW,as shown in Fig. 7, and the confinement time is up to ~3 times the prediction of Goldston L-mode scaling [8]. With thepresent distribution of beam sources, at full power PcQ^INJ is ~0.7, which is not optimum, as indicated in Fig. 7. The highest
power experiments were conducted at higher current (1.0 MA) to
minimize the effects of MHD activity.
The energy stored in the discharge appears to be limited bya maximum 6 p value of ~ 2. Further increases in power at a given
current result in either disruptions or MHD activity degrading thestored energy [3]. At low toroidal field and modest q a (~ 4-5),
the maximum stored energy is in fair agreement with the Troyon 6 T
limit [e^ ax = (2.2-2.5] wQI /aBTJ [11]. At high toroidal fieldand q a the Troyon B T limit would permit 6 D values up to - 3.5.
At higher plasma current (I > 1.1 MA), the favorable
characteristics of supershots have not been observed, and the

HAWRYLUK et al.

60
1
_
0.15-

s

1

<
ft

••

k

A

A

i

0.10 Ip(MA) PCO / P~~~"-~INJ

0.05 ~o-7A

0.9
0.5-0.6
0.9
1.0
0.9 0.63-0.7
1.0. 0.7-0.8

1

L-Mode/~(0.9MA)
1

10
PINJ(MW)15

20

FIG. 7. Variation of global confinement time with injected power for the data set shown in Fig. 6.
Magnetic measurements are compared with the prediction of the L-mode model of Goldston.

difference in tg between balanced and co-only injection at P^ =10 MW is less %han \5%.
Supershot characteristics at higher
current (1.3 MA with 17 MW of neutral beam power) can be achievedby ramping the current from 0.9 MA during the beam pulse,
resulting in q = 5. Further extension of the operating regime
may be possible with the longer-duration (2.0 s) and higher-fullenergy-species-mix beam sources (80? power at full energy) thatare currently being installed.
Present current-ramp experiments
are encouraging in that supershot characteristics are observed
even when fairly large current-ramp rates up to 1 MA/sec areapplied.
By tailoring the beam power waveform and the current
ramp, it may be possible to achieve supershots with much higher
currents, thus avoiding the B limits.
In discharges in which the plasma current is maintained
constant by feedback control, the surface voltage is observed todecrease from the ohmic phase with increasing 6 D and co-injected
power and is typically ~ -0.2 V at the end of injection. Though
the electron
temperature increases and broadens and Z ^decreases, changes in conductivity alone cannot produce a negative
voltage. The changes in plasma geometry coupled with changes in
conductivity are also not large enough to account for the negative
voltage, as determined by a one-and-one-half dimensional magnetic
field diffusion calculation. Monte Carlo beam calculations showthat neoclassical beam-driven currents [12] contribute to thedecrease in surface voltage, but still do not account for themagnitude. The calculations show that, due to the beam orbits,
co-injection is more effective in driving current, particularly inIAEA-CN-47/A-I-3

61

Measured
OhmicOnly
Beam ft Ohmic
Beam, Bootstrap 8 Ohmic

Pinj=IO.6MW,PCo/Pinj=O.57-0.5mzmmmmzm
4.5
TIME(sec)

FIG. 8. Comparison of the measured surface voltage with the calculated surface voltage under different
assumptions for the role of non-ohmically driven currents. The discharge parameters are BT = 5.2 T,
R = 2.45 m and a = 0.8 m.

outer regions of plasma, than counter-injection; thus, even when
the power is nearly balanced, a net beam-driven current effect on
the surface voltage is calculated. By including the neoclassicalbootstrap current predicted by Galeev and Sagdeev [13] andBickerton et al. [1*1] as formulated by Hirshman and Sigmar [12],good agreement is obtained between the calculated and measured
surface voltage, as shown in Fig. 8. These calculations indicatethat for this case approximately 350 kA of bootstrap current ispresent, compared with 420 kA of co-injected current and -160 kAof counter-injected current.
In presently analyzed discharges,
the maximum calculated bootstrap current is HkO kA in a 900 kA
plasma with 370 kA of calculated beam-driven current. While thedriven currents do not exceed the total current, the calculated
voltage is negative because the radial distribution of thebootstrap current is broader than that of the ohmic current,
exceeding the total current density in the outer region of theplasma.
Operationinthe supershot
regimehas resulted
in
substantial increases in neutron source strength and equivalent
Q.
The maximum neutron source strength of 1.2 x 10^° n/s wasTheachieved with 19.6 MW of.beam power, with P(;O /P INJ = °-?'
maximum Q D D of 8.7 x 10" 4 was achieved with 12.3 MW, when the
power was more nearly balanced (.Pcc/PINJ ~ °-56), corresponding toa source strength of 8.9 x 1 0 1 5 n / s .
In the supershot regime,
SNAP and time-dependent (TRANSP) calculations of the neutron flux
are in good agreement (± 25?) with the measured flux.
These
calculations indicate that, for the discharge with the highest

62

HAWRYLUK et al.

value of Q D D , ~ W% of the flux is due to beam-beam reactions, ~53? due to beam-target, and ~ 30% from the thermalized ionsshortly before the end of the beam pulse. In order to estimate
the Q D T which would be achieved witb deuterium and tritium, it isnecessary to assume a composition for the hydrogenic component of
the target plasma. For the hypothetical condition of deuterium
beam injection into a tritium plasma (with the same value of Zeff.
= 3.3), Q D T would be 0.23, due to beam-target reactions. Because
beam fueling is clearly very important, a more realistic casewould be to consider injection using deuterium and tritium beams
into a 50J&/50? deuterium/tritium plasma.
In this case, theresulting value of Q D T would be = 0.18. A significant improvement
over these values can be expected from the new long-pulse ion
sources with better species mix, operation at 120 kV, and nearlybalanced injection at full beam power, even without assuming
higher injected power and improved plasma parameters.

During the past year, substantial progress has been made in
expanding the operating range of TFTR. Ohmically heated pelletfueled discharges have resulted in exceptional central densities
and n g (0) tE(a) T i (0) values of ~ 2 x 10^° nT^-s-keV.
Furthermore, in the saturated regime, the confinement time inpellet-fueled discharges is observed to be longer than that
achieved in gas-fueled discharges.
In these discharges, thedensity profile is very peaked, ne(0)/iie ~ 2.
In neutral-beam-heated discharges an enhanced confinement
regime has been found.
This regime is characterized by peaked
density profiles, broad electron temperature profiles, high ion
temperatures,s 20 keV, and high neutron source strengths. In thisregime, ne(0) T£(a) T ^ O ) values of ~ 2 x 1 0 2 0 m~3.s-keV have also
been achieved.
The operating regime is characterized by a low
initial density and deuterium recycling prior to injection, which
can only be achieved at low I and after extensive limiter
degassing. At high toroidal field (B T = 5.2 T ) , this corresponds
to relatively high values of q (in the range of 7-8) and 6 p (upto 2). In reduced toroidal field experiments and in current ramp
experiments, supershots with q a ~ 4-5 have been obtained.
Significantly
improved performance is achieved with nearlybalanced injection.
In the high temperature regime, the plasma collisionality is
low and 8
is high, so that both beam-driven currents andbootstrap currents are predicted to be important. Calculations
including bootstrap current are in better accord with theexperimental measurements than those including only the beamdriven currents.
If these indications of the bootstrap current
are verified by future experiments, the performance of tokamak
reactors could be fundamentally enhanced.

IAEA-CN-47/A-I-3

63

Acknowledgements
We are grateful to D.J. Grove and J.R. Thompson for theiradvice and support, and to J. Strachan for his many
contributions. This work was supported by US DOE Contract No. DEAC02-76CHO3073. The ORNL participants were also supported by US
DOE Contract No. DE-ACO5-84OR214OO with Martin Marietta Energy
Systems, Inc.

REFERENCES[1] SCHMIDT, G., et al., Paper IAEA-CN-47/A-III-4, these Proceedings, Vol. 1.
[2] GOLDSTON, R.J., et al., Paper IAEA-CN-47/A-II-l, these Proceedings, Vol. 1.
[3] McGUIRE, K., et al., Paper IAEA-CN-47/A-VII-4, these Proceedings, Vol. 1.
[4] HILL, K.W., et al., Paper IAEA-CN-47/A-IV-2, these Proceedings, Vol. 1.
[5] MURAKAMI, M., et al., Plasma Phys. Controll. Fusion 28 (1986) 17.
[6] EFTHIMION, P.C., et al., in Plasma Physics and Controlled Nuclear Fusion Research 1984
(Proc. 10th Int. Conf. London, 1984), Vol. 1, IAEA, Vienna (1985) 29.
[7] BELL, M.G., et al., Plasma Phys. Controll. Fusion 28 (1986) 1329.
[8] GOLDSTON, R.J., Plasma Phys. Controll. Fusion 26 (1984) 87.
[9] STRACHAN, J.D., et al., submitted to Phys. Rev. Lett., 1986.
[10] DYLLA, H.F., et al., submitted to Nucl. Fusion, 1986.
[11] TROYON, F., et al., Plasma Phys. Controll. Fusion 26 (1984) 209.
[12] HIRSHMAN, S.P., SIGMAR, D.J., Nucl. Fusion 21 (1981) 1079.
[13] GALEEV, A.A., SAGDEEV, R.Z., Zh. Ehksp. Teor. Fiz. 53 (1967) 348 (Sov. Phys. - JETP
26 (1968) 233).
[14] BICKERTON, R.J., CONNER, J.W., TAYLOR, J.B., Nat. Phys. Sci. 229 (1971) 110.

DISCUSSION
A. GIBSON: In the two cases for which you quote the large values
what is the ratio nD/ne?
R.J. HAWRYLUK: In high density pellet fuelled discharges, the correction
factor for depletion is small, since Zeff is about one. In the supershot discharge, thedepletion factor, nD(0)/n,,(0), is about 0.7.
M. GREENWALD: Are you quite sure of the Zeff profiles in your pellet
fuelled shots? I ask this because there seems to be evidence of impurity accumulation.
R.J. HAWRYLUK: We do not have Zeff profile measurements for the pellet
fuelled discharges. In some discharges, there is substantial radiation on axis;
however, even small concentrations of impurities would be adequate to explain the
power loss.
Further, experiments are required to analyse impurity transport in pellet fuelleddischarges and to determine whether accumulation occurs and, if so, under what
conditions.

64

HAWRYLUK et al.

S. ITOH: I would like to ask about the target plasma for the 'supershot'. Is thetarget plasma in a slide-away regime? Judging by the target parameters, it would
seem to be. Secondly, did you see any input power limit or threshold for the injection? The supershot discharges seem to be similar to the ASDEX discharges withslide-away electrons reported at the 1982 Conference.
R.J. HAWRYLUK: In some of the target plasma, electron cyclotron emission
measurements indicate the presence of slide-away electrons. However, they are notevident in all of the supershot discharges and the occurrence of a slide-away does notappear to be a critical parameter.
Regarding your second question, at a given current there is an upper power limit
present that corresponds to attainment of the j8p limit d8p ~ 2) which I discussed.
However, higher current operation enables us to inject the available 20 MW ofneutral beam power and to maintain supershot characteristics.
K. LACKNER: Troyon-type scalings, ^pol) crit ~ q0, are obtained for current
density profiles adjusted to maintain q on axis at a constant value of about one. A
reduction in /3pol cri, and independence of it from qa, as reported by you for low
currents, would follow from theory for a current density profile flatter and nondependent on qa.
R.J. HAWRYLUK: In their paper, McGuire and co-workers* show that the
profiles are not optimal, which is why the experimental beta limit is below thatpredicted by Troyon. The lack of sawteeth indicates that q on axis is greater than one.
However, since the current penetration time is long compared with the beam duration, q is not substantially greater than one, according to our analysis.
M. BRUSATI: You have reported degradation of performance correlated withenhanced MHD activity. Can you comment on the kind of MHD spectra that you get?
R. J. HAWRYLUK: The mode number of the continuous MHD activity which
degrades performance is typically m/n = 2/1 or 3/2. For further details, I would refer
you to the paper by McGuire and co-workers that I have just mentioned.*

Paper IAEA-CN-47/A-VII-4, these Proceedings, Vol. 1.

IAEA-CN-47/A-I-5

EXPERIMENTAL STUDIES IN JET
WITH MAGNETIC SEPARATRIXA. TANGA, D.V. BARTLETT, K. BEHRINGER,
R J . BICKERTON, A. CHEETHAM, J.G. CORDEY,
A. GIBSON, N.A. GOTTARDI, A. GONDHALEKAR,
P.J. HARBOUR, H. JAECKEL, M. KEILHACKER,
E. LAZZARO, V. MARTENS, P. NOLL, S. NOWAK,
F.C. SCHULLER, A.J. TAGLE, A. TARONI,
P.R. THOMAS, F. TIBONEJET Joint Undertaking,
Abingdon, Oxfordshire,
United Kingdom

Abstract
EXPERIMENTAL STUDIES IN JET WITH MAGNETIC SEPARATRIX.
The main objectives of the experiments described are (i) to compare the global confinement
characteristics of separatrix and limiter discharges and (ii) to study the conditions for the creation ofa high density, highly radiative, cool plasma region near the X-point capable of screening and isolating
the bulk plasma. Compared to similar limiter discharges, the thermal energy content of single nullX-point discharges (SN) is roughly a factor of two higher. Plasmas with even modest separation between
X-point and target plates show typical signatures of H-mode discharges. Multi-channel infrared interferometer meausurements provide evidence of the formation of a high density diverted plasma withaverage density — 1020 m" 3 , an order of magnitude higher than the main plasma. Measurements witha bolometer camera array show that, in high density discharges with additional power, up to 40% of
the input power is radiated within a —0.2 x 0.2 m2 toroidal region around the torus.

1.

INTRODUCTION

The main aim of JET is to study plasmas under conditions
approaching those needed for a fusion reactor. At present, in
limiter discharges, confinement with additional heating is low(L-mode). Better confinement (H-mode) has been observed intokamaks with magnetic limiters or divertors. In JET, amagnetic configuration with a separatrix inside the vacuum
vessel can be created [ 1 ] . Two ("double null configuration",DN) or one ("single null configuration", SN) poloidal fieldnulls are then produced near the top and/or bottom of the vessel
as shown in Figs. 1a and 1b.
In these configurations the plasma is detached from boththe limiter and the inner wall and recycling occurs in an open
divertor region near the X-point. Experiments have been carried

65

66
Z(m)
20

1.0

iBH)*

0

-1.0

\
-2D

1.0

2.0

3.0

4.0
R(m)

FIG. I. (a) Flux plot of a double null configuration. Plasma current = 2 MA, toroidal field = 2.8 T,
average electron density = 3 X 1019 m'3, poloidal beta = 0.44, plasma elongation = 1.8.
(b) Flux plot of a single null configuration. Plasma current = 2 MA, toroidal field = 2.2 T, average
electron density = i x 1019 m~3, poloidal beta = 0.6, plasma elongation = 1.65(ttlO237,
t = 13.5 s).

out with plasma currents up to 2.5MA in the DN and up to 3-OMAin the SN configuration. A maximum additional power of 9MW has
been injected (D° into D + ) .
2.

FORMATION OF A MAGNETIC SEPARATRIXA magnetic separatrix can be formed within the vacuum
vessel of JET using the multlpolar field normally used tocontrol the elongation and triangularity of the plasma
cross-section. In the DN configuration, the present poloidal
field equipment allows operation at plasma currents up to 2.5MA.
The resulting plasma has an elongation of 1.8; the vertical
position is unstable with an open loop growth rate of 150s" 1 .
The separation between the X-point and the graphite target
plates is a linear function of the currents in the shaping
coils. For the configuration in Fig.1a the separation is -0.1m.
In the SN configuration, plasma currents up to 3.0MA have been
produced with an elongation of 1.65; the growth rate of thevertical instability is substantially smaller than for the DN
configuration.
The maximum separation between the X-point and the target
plates obtained so far at 2MA with SN is -0.13m and may be
increased by displacing the magnetic axis below the midplane.

IAEA-CN-47/A-I-5

67

The radial position of the null point is held constant during
the discharge, to within a few centimetres. The distance
between the separatrix and the antennae for ICRH is finely
controlled by the radial position feedback system and this canallow good coupling of ICRH to the plasma. In both DN and SN
configurations, Langmuir probe measurements give the e-folding
length for decay of density and temperature as 0.06m and 0.10m,
respectively.
The location of the separatrix has been corroborated towithin a few centimetres by observations with a bolometer camera
array and TV cameras and by the separation between the erosion
marks on the graphite target tiles. The magnetic configurationis only slightly sensitive to changes in internal parameters,such as the poloidal beta and internal inductance.
3.

HIGH RECYCLING IN THE X-POINT REGION

The formation of a magnetic separatrix within the vacuum
vessel produces a significant change in the discharge
characteristics. For separations between the plasma and thellmiter larger than 0.07m, deuterium recycling (monitored by D asignals) shifts from the limiters to small regions near the
X-points. Both the D a monitors and TV observations in the nearinfra-red show the formation of bright regions (-0.2m across)coinciding with the X-points. In high density DN discharges,
only one bright region is observed. The application of a radial
magnetic field shifts the toroidal plasma column vertically andthe bright region moves correspondingly, indicating that theformation of only a single bright region is not connected withslight up-down asymmetries in the poloidal field.
Simultaneous with the shift in recycling characteristics,
stronger gas-fuelling is needed to sustain the average plasma
density, Z e ff is reduced and the bolometer camera array shows adecrease in bulk plasma radiation and a large increase in theradiation from the X-point region. As shown in Fig.2a for aseries of ohmic DN discharges, the total radiated power is in
most cases between 70 and 100? of the input power and the powerradiated in the X-point region increases with plasma density
more rapidly than the total radiated power. SN discharges show
similar behaviour with the power radiated from the X-point
region, following approximately a density-squared dependence.
The size of the highly radiative region in the vicinity of the
X-point can be estimated to be SO.2x0.2m 2 .
The far infra-red multichannel interferometer suggests asimilar small region with high electron density near the
X-point. When one of the vertical channels of theinterferometer passes through the X-point region, the fringe
count for that channel shows a substantial increase over theother channels. This can be interpreted as a measurement of thelocal X-point line density. In Fig.2b, the difference between

68

TANGA et al.
1.5

a)

P
(MW)

X
•—"•"*

1.0 -

XX

x

«

X

X
X

0.5-

*

k
0

b)

A(/nedl)(1Ol9m2) _

a

s-

-1

FIG. 2. (a) Radiated power versus line average electron density for Ohmic double null configuration.
Dots represent power radiated from X-point, crosses total power radiated from plasma,(b) Difference in line density between two adjacent interferometer channels one of which intercepts
the X-point.

two adjacent channels of the interferometer has been plottedagainst the line average electron density. Assuming that thesize of this high density plasma region corresponds to that of
the highly radiative region the density in the X-point region isestimated to be -10 2 O m~ 3 . No data are available for average
because the channel ofplasma densities larger than 2x101
the interferometer intercepting the X-point loses fringes.
Assuming that the thickness of the high recycling region
corresponds to the mean free path for neutrals, X°, theinterferometer data provides a measure of [2]n t \°v°
TavT

where v° is the effective speed of the emitted atoms and <ov> isthe effective rate coefficient for ionisation. Assuming a

IAEA-CN-47/A-I-5

69

single temperature, T t , for electrons, ions and neutrals, theright hand side of this equation is only a function of T^.
T t ~5-10eV is obtained for n e >1xiO 1 9 m~ 3 . The density, n t , mayalso be estimated from a zero-dimensional model which balances
the power flow from the bulk plasma with sonic flow at the highdensity region in the scrape-off layer. For n g >1x10 l 9 m" 3 ,n^.-4x1019rn~3 is obtained, in reasonable agreement with theestimates made earlier.
This trend towards high recycling is further corroboratedby Langmuir probe measurements which suggest that the Mach
number of the flow parallel to the magnetic field decreases withincreasing plasma density.
4.

IMPROVED CONFINEMENT WITH ADDITIONAL HEATING IN THE SINGLE
NULL CONFIGURATIONDN X-point discharges show an energy confinement time some
20-30$ higher than similar limiter or inner wall discharges.
However, no clear H-mode transition was identified as the power
levels available. Recent experiments have therefore
concentrated on the SN configuration. First results show asubstantial improvement of energy confinement in SN X-point
discharges compared to similar limiter or inner wall discharges.
The experiments have been carried out with neutral beam
injection heating with up to 9.OMW power. Ion cyclotron heating
has been applied in some of the shots with a power of up to 4MW(but keeping the total input power, Ptot = p fl + P NBI + P RF= 2 0 M W )•In Fig.3 the variation of the total plasma thermal energy
content with total input power P^ot *"or t w o s e r i e s °f SN X-point
discharges at plasma currents of 2MA and 3MA is compared withthe corresponding data sets for limiter discharges. While theseparation of the X-point and the target plates is -0.13m at 2MA
this is reduced to ~0.5m at 3MA and is then rather close to thewall. For either plasma current the energy content of the
X-point discharges is up to a factor of 2 larger than that of
the corresponding limiter discharge with the same additional
power. In general, the trend with increasing heating power isalso similar in all cases.
As shown in Fig.4, the electron temperatures (and also thecentral ion temperatures ifor such an X-point H-mode plasma aresubstantially higher than for a limiter plasma with the same
parameters of plasma density, plasma current, toroidal field andP t o t . It should be noted that the electron temperature profileis broader, with the edge temperature at R=4.0m being some 2.5
times higher.
As pointed out already the SN discharges at 2MA exhibit
clear indications of a transition into an H-mode [ 3 ] . As shownin Fig.5 the edge D signal (f) drops and the line integrated
electron density (d) increases more rapidly. These events occur
shortly after the crash of a sawtooth when the edge temperature

70

TANGA et al.

fl 2 MA (Limiter)
i 2 MA (X-point)
6 -•o 3MA (Limiter)• 3MA (X-point)

W(MJ)

-

1

•
SINGLE NULL

4 -

2 4^

i

LIMITER

i

10

15

P,c

F/G. 5. Tofa/ plasma thermal energy content as measured by diamagnetic loop versus total input power
for single null and limiter configurations.

(keV)

FIG. 4. Electron temperature profiles (a) X-point (#10237, t = 13.5 s) and (b) limiter discharge(§7709). Central ion temperature for SN (X) and Limiter (L) discharge.

71

IAEA-CN-47/A-I-5
Transition
Ptot

(MW)

12.0
FIG. 5.

13.0

14.0
time (s)(a) Total input power Pm, for #10237;(b) Power conducted to the target plates from, bolometric analysis, Plol — Prad — W;(c) Total energy content from the diamagnetic loop, W;{d) Line electron density, $nedl;(e) Plasma edge electron temperature,(f) Da signal from the plasma boundary.

(e) increases abruptly. The power flow to the target plates (b)is substantially reduced. This phase is subject to an abrupt
termination when the average density reaches some limiting value(as evidenced more clearly in other JET Pulses where this event
is not obscured by the end of additional heating). The averagedensity, temperature and thermal energy content (c) fall in
about 40ms, accompanied by a spike in the X-point radiation.
During the H-phase the plasma density profile broadens withsteep gradients developing at the edge.
Impurity radiation from the plasma edge (0 VI, C IV) falls
by about a factor of two at the transition to the H-regime. At
this time the ICRH power ceases to be coupled to the plasma(probably due to the changed edge conditions) and the nickel
influx from the antenna screen falls to zero. Nevertheless, thetotal nickel content as observed from the Ni XXV and Ni XXVI72

TANGA et al.
neXe(irfW-s1)

0.5
Normalized radius

P MAX

FIG. 6. nexcprofiles for SN, limiter and inner wall discharges, (a) OHphases, (b) dots for dischargeof Fig. 4att = J3.8s, (c) for reference limiter dischare, (d) for reference discharge lying on the innerwall with NBI heaitng (#8899), with larger input power but lower energy content than (b).

line intensities remains constant indicating substantially
improved confinement of metals in the interior of the plasma.
As indicated by the soft X-ray radiation profiles, however, thisdoes not lead to accumulation of impurities.
A preliminary analysis of transport properties comparing SN
X-point with limiter and inner wall discharges has been donewith the transport code Jetto [1], Using the transport models
described in [5] good simulations of all the relevant measured
quantities (n e , T e , radiation profiles, central ion temperature,Z e f f ) have been obtained in the ohmic phase of these three typesof discharges. With additional heating, however, a degradationof the coefficient for the electron energy transport must be
introduced for the inner wall and limiter case but not for the
X-point discharge. The radial profiles of n g x e (Fig.6) showthat these differences hold over the whole plasma
cross-section.

5.

CONCLUSIONS

Separatrix operation appears to induce substantial changes
in the plasma behaviour compared to limiter operation. This ispossibly a result of the high shear at the boundary of theplasma. Even with a modest separation of the X-point from thetarget plates, a high recycling regime has been observed in thisregion at high plasma densities. With additional heating thestored energy increases substantially above that forcorresponding limiter discharges. A further increase in stored

IAEA-CN-47/A-I-5

73

energy is observed for both 2MA and 3MA discharges leading to
global energy confinement times in excess of 0.6s, about afactor of two higher than for the corresponding limiter
discharges. These discharges also exhibit other signatures(e.g., reduced edge recycling) normally associated with
H-modes.
ACKNOWLEDGEMENTS
The authors would like to acknowledge the effort of thewhole JET team and, in particular,of the RF and NB Divisions.

REFERENCES[1] TANGA, A., in Controlled Fusion and Plasma Physics (Proc. 12th Europ. Conf. Budapest,
1985), Vol. 1, European Physical Society (1985) 70.
[2] NEUHAUSER, J., WUNDERLICH, R., in Plasma-Surface Interactions in Controlled Fusion
Devices (Proc. 7th Int. Conf. Princeton, NJ, 1986), to be published in J. Nucl. Mater.
[3] WAGNER, F., et al., Phys. Rev. Lett. 49 (1982) 408; WAGNER, F., et al., in Plasma Physics
and Controlled Nuclear Fusion Research 1982 (Proc. 9th Int. Conf. Baltimore, 1982), Vol. 1,
IAEA, Vienna (1983) 43.
[4] DtiCHS, D., et al., in Controlled Fusion and Plasma Physics (Proc. 12th Europ. Conf. Budapest,
1985), Vol. 1, European Phsyical Society (1985) 22.

DISCUSSION
H.P. FURTH: Your results seem to me to be extremely significant as well as
encouraging. One could say that you have shown that the quality of confinement in
JET need not deteriorate at heating powers much higher than Ohmic, but this would
be an understatement.
A critical figure of merit for tokamaks is VH = Ptola/Iplasma. For Ohmic heating,
VH is normally less than 1 V. For a practical tokamak reactor, VH must exceed
50 V, and even for an ETR, VH must exceed ~ 10 V. Your demonstration that JET
confinement is unimpaired for VH > 3 V is therefore a truly major step forward.
A. TANGA: Thank you for the comment.

IAEA-CN-47/A-n-l

ENERGY CONFINEMENT AND
PROFILE CONSISTENCY IN TFTR
R.J. GOLDSTON,
V. ARUNASALAM,
M.G. BELL,
M. BITTER,
W.R. BLANCHARD, N.L BRETZ, R. BUDNY, C.E. BUSH1, J.D. CALLEN 2 ,S.A. COHEN, S.K. COMBS1, S.L. DAVIS, D.L. DIMOCK, H.F. DYLLA,
P.C. EFTHIMION, L.C. EMERSON1, A.C. ENGLAND1, H.P. EUBANK,
R.J. FONCK, E. FREDRICKSON, H.P FURTH, G. GAMMEL, B. GREK,
L.R. GRISHAM, G. HAMMETT, R.J., HAWRYLUK, W.W. HEIDBRINK3,
D.B. HEIFETZ, H.W. HENDEL 4 , K.W. HILL, E. HINNOV, S. HIROE1,
H. HSUAN,
R.A. HULSE, K.P. JAEHNIG,
D. JASSBY,
F.C. JOBES,
D.W. JOHNSON,
L.C. JOHNSON,
R. KAITA,
R. KAMPERSCHROER,
S.M. KAYE, S.J. KILPATRICK, R.J. KNIZE, H. KUGEL, P.H. LaMARCHE,
B. LeBLANC, R. LITTLE, C.H. MA 1 , D.M. MANOS, D.K. MANSFIELD,
R.T. McCANN,
M.P. McCARTHY,
D.C. McCUNE,
K. McGUIRE,
D.H. McNEILL,
D.M. MEADE,
S.S. MEDLEY,
D.R. MIKKELSEN,
S.L. MILORA1,
W. MORRIS5,
D. MUELLER,
V. MUKHOVATOV6,7
8
E.B. NIESCHMIDT ,J. O'ROURKE ,D.K. OWENS,
H. PARK,
N. POMPHREY,
B. PRICHARD,
A.T. RAMSEY,
M.H. REDI,
A.L. ROQUEMORE, P.H. RUTHERFORD, N.R. SAUTHOFF, G. SCHILLING,
J. SCHIVELL, G.L. SCHMIDT, S.D. SCOTT, S. SESNIC, J.C. SINNIS,
F.J. STAUFFER9,
B.C. STRATTON,
G.D. TAIT,
G. TAYLOR,
J.R. TIMBERLAKE, H.H. TOWNER, M. ULRICKSON, V. VERSHKOV6,
S. VON GOELER,
F. WAGNER10,
R. WIELAND,
J.B. WILGEN1,
M. WILLIAMS,
K.L. WONG,
S. YOSHIKAWA,
R. YOSHINO 11 ,6
K.M. YOUNG, M.C. ZARNSTORFF, V.S. ZAVERYAEV , S.J. ZWEBEN

Princeton Plasma Physics Laboratory,
Princeton University,
Princeton, New Jersey,
United States of America

1

Oak Ridge National Laboratory, Oak Ridge, TN, USA.
University of Wisconsin, Madison, WI, USA.
3
GA Technologies Inc., San Diego, CA, USA.
4
RCA David Sarnoff Research Center, Princeton, NJ, USA.
5
Balliol College, University of Oxford, Oxford, UK.
6
Kurchatov Institute of Atomic Energy, Moscow, USSR.
7
EG&G Idaho, Inc., Idaho Falls, ID, USA.
8
JET Joint Undertaking, Abingdon, Oxfordshire, UK.
9
University of Maryland, College Park, MD, USA.
10
Max-Planck-Institut fur Plasmaphysik, Garching, Fed. Rep. Germany.
11
JAERI, Naka-machi, Naka-gun, Ibaraki-ken, Japan.
2

75

76

GOLDSTON et al.
Abstract

ENERGY CONFINEMENT AND PROFILE CONSISTENCY IN TFTR.
A new regime of enhanced energy confinement has been observed on TFTR with neutral beam
injection at low plasma current. It is characterized by extremely peaked electron density profiles andbroad electron temperature profiles. The electron temperature profile shapes violate the concept of profile consistency in which <Te0/Te>v is assumed to be a tightly constrained function of c^, but they arein good agreement with a form of profile consistency based on examining the temperature profile shapeoutside of the plasma core. The enhanced confinement regime is only obtained with a highly degassed
limiter; in discharges with gas filled limiters corrective losses are calculated to dominate the edge electron power balance. Consistent with the constraint of profile consistency, global confinement isdegraded in these cases. The best heating results in the enhanced confinement regime are obtained withnearly balanced co- and counter-injection. Much of the difference between balanced and co-only
injection can be explained on the basis of classically predicted effects associated with plasma rotation.

1.

Introduction

A new regime of enhanced energy confinement has been
observed on TFTR with neutral beam injection at low plasma
current. It is characterized by extremely peaked electron density
profiles and broad electron temperature profiles. The best heating
results in this regime are obtained with nearly balanced co andcounter injection. The properties of these discharges (referred
to as "supershots") raise questions about some of the recently
achieved understanding of tokamak auxiliary heating physics electron temperature profile consistency, and the lack ofimportance of toroidal rotation, among others. In this paper we
discuss the implications of our results in these areas. We begin(Sec. 2) by putting the electron stored energy results in this new
regime into the context of previous scaling laws and TFTR data,
showing that these discharges exhibit enhanced electron energy
confinement as well as the high total stored energy and ion
temperature discussed in [1]. Next (Sec. 3) we study the electron
temperature profile shapes of these enhanced confinement
discharges, in comparison with L-Mode and ohmically heatedprofiles on TFTR. In Section 4 we address issues connected withthe electron density profile shape. In Section 5 we discuss therole of rotation and balanced versus unbalanced injection.
2. Stored Electron Energy in Supershots
Figure 1 shows the electron energy content from a series ofscans of neutral beam power at I = 0.9, 1.4, and 2.2 MA, with theplasma resting on the inner-wall carbon limiter. The scans at
higher current were performed without any special limiter
degassing procedures.
They show typical L-Mode behavior: thestored energy reaches equilibrium during the beam pulse, and broad
density profiles are observed. The 0.9 MA data were taken after
extensive degassing with low density He discharges [2]; they

77

IAEA-CN-47/A-O-1
0.8
2.2 MAo 0.9 MA Super
0.6 * I.3MA I Ramp
p

o

°

B °
=

0.4

OD

a

02

10

15

Ptot(MW)

FIG. 1. Electron stored energy versus total healing power for 2.2 MA and 1.4 MA L-mode discharges
and for 0.9 MA enhanced confinement discharges. Also shown are enhanced confinement discharges
at 1.3 MA achieved by ramping the plasma current up from 0.9 MA during the injection pulse. The
curves represent 0.4 x w\ot, where w^m is the L-mode scaling predication [4] for total stored energy(see text).

generally show rising stored energy throughout the beam pulse, andat higher powers they exhibit strongly centrally peaked density
profiles.
Shots with severe MHD activity [3] which caused thestored energy and neutron flux to fall at the end of the pulse
have been excluded from this data set. The higher current data
sets have been constrained to 0.25 < p c o / p i n i < 0.75, while the0.9 MA data set has been constrained to 0.35 < p c o / p i n i < 0.65, inorder to minimize the effects of unbalanced injection, which are
greatest in the low-current region [1]. The current of 0.9 MA wasselected because at lower plasma currents (I = 0.7-0.8 MA) highpower operation
is prevented
by severe MHD activity anddisruptions, while the data set at higher currents (I = 1 - 1.1
MA) and P i n , less than 10 MW is very limited.
Up to 17 MW ofneutral beam power has been injected into discharges with an
initial plasma current of 0.9 MA, without inducing strong MHD
oscillations or disruptions, by ramping the plasma current up to
1.3 MA during the 0.5 sec beam pulse. Due to our present beam
configuration, at these high powers p c c / p i n < > 0.70, which has
been observed to result in significantly reduced heating compared
to balanced injection at lower powers.
Also plotted on thisfigure is the prediction of L-Mode scaling [4] (for H° ->• D +injection) where the prediction of the scaling law for electronenergy has been taken to be 0.4 times the prediction for total
stored energy, since this ratio is characteristic of high power
TFTR results in the standard regime.
While the higher current
results approach the L-Mode prediction at high powers, the 0.9 MA

78

GOLDSTON et al.

1.5

2.5
RADIUS (m)FIG. 2. Electron density and temperature profiles for a 1.0 MA enhanced confinement discharge (solid
points) and for a 1.4 MA L-mode discharge (open points). For the enhanced confinement discharge:
We = 0.52 MJ. For the Umode discharge: We = 0.47 MJ. PJPinj = 0.7 for both.

supershot results clearly diverge from it. The stored electronenergy of the 1.3 MA current-ramp supershots is remarkably close
to the standard regime 2.2 MA data. The magnetically-measured
total stored energies of enhanced confinement discharges show evenmore of a gain over standard L-Mode discharges. The total stored
energy of 0.9 MA supershots equals that of 2.2 MA standard regime
discharges at powers in the range of 11 MW. The fact that kinetic
stored energy calculations agree well with the diamagnetic
measurement results (± 5% in the 0.9 MA data set shown here)
indicates that the thermal and beam ions gain in stored energy as
expected due to the high electron temperatures and peaked density
profiles provided by supershots.

79

IAEA-CN-47/A-H-1
10

-i

1

F

1

T e > l keV

1

1

r

ool9MA Super
a 1.3MA I p Ramp

T e (o/2)

j

0

L

J

I

I

L

0.5 r/a1.0

FIG. 3. Electron temperature profile shape normalized tor = a/2 for the data sets of Fig. 1. The factorof three is applied in order to centre the mid-radius on a single decade scale. The error bars represent
the root mean square (RMS) variation of the data.

3. Electron Temperature Profile Consistency
Figure 2 shows an overlay of the Thomson scattering
measurements of T g (R) and n e (R) from a supershot with I s 1 HA,
B T = 5T, and from a 1.4 MA discharge at the same toroidal field.
Both discharges had a well-degassed limiter, but in the 1.4 MA
case B nonetheless saturated after 0.3 sec into the beam pulse,
and une density profile did not gain the peaked shapecharacteristic of the enhanced confinement mode. These results,
and others like them, violate the concept of electron temperature
profile consistency in which T (0)/<T > is taken as the measureof profile shape and is postulated to be a tightly-defined,
monotonically rising function of q . This measure of profile
shape is, however, already suspect as an indicator of microscopictransport mechanisms, since it can be largely enforced by sawtoothoscillations [4,5], if x e is sufficiently peaked to the outside ofthe discharge. On TFTR it has been previously observed [6] thatthe detailed electron temperature profile shape, outside of thecore of the plasma and in the region where T g > 1 keV, shows aremarkable constancy, independent of heating power - and mostsurprisingly - independent of q a .
Figure 3 shows electron
temperature profile shapes normalized at r = a/2 for the points
with P t o t > 8 MW in the power scans of Fig. 1. The electron
temperatures measured by Thomson scattering have been mapped toflux surfaces, and plotted as a function of midplane minor
radius. The two-sided electron temperature measurement ensuresthat this process unambiguously corrects for the strong Shafranov
shift observed in the low current, enhanced confinement regime.

80

GOLDSTON et al.

0.2

0.4
RADIUS (m)

0.6

0.8

FIG. 4. Time evolution of TJr) from Thomson scattering, mapped to time dependent magnetic flux
surfaces for 9.5 MW balanced injection into a 0.8 MA discharge, in enhanced confinement regime.
Neutral beam injection heating from t = 4.0 s to 4.5 s.

The error bars indicate the R.M.S. variation of the data. The 0.9
MA supershot temperature profiles line up with the L-Mode profiles
in the region outside of a/3. The few current ramp shots show anapproximately 10$ higher Te(0.8a)/T (0.4a) than the others.
Whether this is significant is difficult to say. Ohmic discharges
at these same currents overlay well with this data, although intheir case the selection for T > 1keV reduces the amount of data,
available outside 0.7a. In general it is found here, as in thedata of [6], that the temperature profiles tend to fall off more
steeply in the region where T e < 1 keV. This suggests thatprocesses such as impurity radiation, charge-exchange, andconvection, which are likely to be more important at the colderedge of the plasma, control T g in this region and help set theamplitude for the overall electron temperature profile.
While the mid-region temperature profiles are consistent
with previous results in TFTR, the central region of thetemperature profile in high-power enhanced confinement discharges
is truncated in comparison with high-q ohmic target discharges (t= 4.45 sec versus t = 4.0 sec in Fig. 4 ) , despite the absence ofregular sawtooth oscillations during the heating phase [3]. One
hypothesis to explain this effect is that the neutral beams
provide a large source of cold electrons in the core region.
Time-dependent transport analysis of balanced injection shots
indicates that core region electron thermal convection losses

81

IAEA-CN-47/A-n-li11<"2.2MA1, . . .
. . . . . [L-Modex I.4MA-I
O0.9MA Super0- o £/

X

o

D

X

x

x

x

a

& I.3MA IpRamp
*

i -5
"

-in

-

x
x

a

a>

-

-

r/a = 5/6

o° o°°

°

1

1

1

5

10

15

P

abs (MW)

FIG. 5. Edge temperature gradient versus absorbed beam power for the same data sets as in Fig. 1.

(5/2reTe) are comparable to electron conduction losses at the endof injection. The argument that an effect of this magnitude isadequate to alter T g (r) in the core is bolstered by theobservation that an even stronger central flattening of T is seen
early in the injection pulse (t = 4.15 sec in Fig. 4 ) , when the
rapid build-up of central density results in a dWg/dt term whichis also comparable to conduction, but over a larger region.
Furthermore, when the cold electron source is turned off at theend of injection, the central electron temperature rises (t = 4.6
sec in Fig. 4 ) , and the profile shape regains the characteristic
central peak of a high-q ohmic discharge. (Issues associated with
the A' stability of these discharges are discussed in [3]).
If the electron temperature profile in the bulk of the minor
radius in supershots is controlled by a mechanism of profile
consistency such as transport coefficeients which depend strongly
on vTg/T , and the core region is possibly clamped by convective
losses, chen we should look to the outer region for the enhanced
confinement properties of supershots. Figure 5 shows dTg/dr at
r = 5a/6 versus absorbed power for the data set of Fig. 1. As
expected for consistency with L-mode current scaling, the 1.4 MA
shots show a lower edge temperature gradient than the 2.2 MAshots. However at high power the 0.9 MA supershots have a steeper
outer region temperature gradient than even the 2.2 MA shots. On
the basis of L-mode scaling, and profile consistency, these
gradients would normally be expected to vary inversely withdensity, since in the L-Mode W g is found to depend only weakly on
n g . In the power range of P . , > 10 MW, the 2.2 MA discharges in
thisdata set have fL a
4.4 X 1O1'm"3, the 1,4 MA shots and theLaid udtd s m [lave u_~ H.t
ave n p %n3.>{ x 10 19 m" 3 , while the 0.9 MA
current ramp shots both have
The lower line average
supershots have n » 2.7 x

82

GOLDSTON et al.

density of the supershots is therefore not adequate to explain
their steeper edge gradients.
This suggests that the reduced
recycling coefficient of the degassed limiters, coupled with deep
beam fueling, permits a low edge density and less convective
losses in the enhanced confinement regime, giving rise to a hotter
edge.
Absolute Da measurements, coupled with neutral modelingcalculation, indicate T of the order of 30 msec for the gasfilled limiter case shown in Figure 4 of [1], and dTg/dr at 5a/6
in that case is only 3.5 keV/m.
Transport analysis of thisdischarge indicates that at the edge of the plasma convective
losses dominate conduction in the electron power balance (as inhigh recycling conditions in D-III [7]), explaining the low edge
electron temperature and, through the constraint of profile
consistency, perhaps the overall reduction in confinement.
4.

The Electron Density Profile

The peaked density profiles observed in the enhanced
confinement regime appear to be the result of a beam build-up
instability [8]. In the discharge illustrated in figure 4, at 0.1
sec into the beam pulse transport analysis calculations indicatethat nb(0)/ne(0) ~ 0.4, and the beam density is strongly peaked on
axis, contributing most of the core density gradient. As the beamions slow down the background density builds up, and at the end of
the beam pulse nb(0)/ne(0) has dropped to ~ 0.2.
The central
electron density rises at about 7055 of the beam fueling rate forthe first 0.3 sec of injection, and later rises at 20-30? of thesource rate. The importance of central beam deposition to thisbuild-up process may explain the sensitivity of the supershot mode
to plasma current. The minimum attainable initial target density
in cases with a well-degassed limiter is observed to be
proportional to I D
in the range of 0.8-1.2 MA, and theShafranov shift drops rapidly as a function of current, so that/ndl between the edge and the plasma core increases strongly with
I . The shift may be especially important, because the tightly
compressed outer flux surfaces present a thin target to neutralbeam injection, and particles deposited in that region are
distributed around the volume of the flux surface, which islargely hidden from the beam by the plasma core. Spectroscopic
measurements also indicate higher carbon influx rates at higher
currents.
This further helps to explain the absence of peaked
density profiles and supershot behavior at high current. L-Modeand ohmically heated plasmas generally have more peaked density
profiles at q >5 as well, suggesting an increased susceptibility
to this process at high qa.
There are a number of ways in which a peaked density profile
can mediate improved confinement. One natural hypothesis is thatthe peaked profiles reduce r\- and/or TI6 which could be importantfactorsintheinstabilities
which
drive
transport(rii e s d m T j /dinn, ). n< has not been measured on TFTR, but
n. 'can be calculated' from the Thomson scattering data.
In theIAEA-CN-47/A-II-l

83

region inside 0.4a n e is significantly lower in supershots than instandard discharges, dropping at high powers from values in therange of 1.7 at 0.4a for standard discharges to 1.2 in theenhanced confinement mode. An alternative hypothesis discussed in
section 3 above, is that high edge electron temperatures, in thepresence of reduced recycling and low edge densities, may result
in relatively low values for edge transport coefficients as in the
H-mode.
In the presence of even constant edge transportcoefficients, electron temperature profile consistency arguments
which take into consideration the presence of a centrally peaked
density profile, but assume that the edge electron thermal
diffusivity responds only to the local n g , predict a significant
improvement in W [9].
5.

Balanced vs. Unbalanced Injection and Plasma Rotation

As indicated in [1], the stored energy in supershot plasmas
shows a considerable improvement as the beam powers in the co andcounter directions are brought into balance.
Because of thepresent configuration of TFTR (3 co-beamlines and 1 counterbeamline for the usual direction of I ) we have not obtained highpower data with pure counter-injection and a well-degassed
limiter.
At low currents (600-800 kA), however, enhanced
confinement discharges with favorable time evolution and highly
peaked density profiles can be obtained with only 5 MW of balanced
injection power, and the full spectrum of pCr/pin-j c a n °e
explored. We find that the density increase at fixed injection
power drops with rising P c o /P i n ^, but the peakedness of thedensity profile falls off as one moves away from balanced
injection in either direction.
Stored energy and especiallyneutron Droduction fall off as well. Plasma rotation speeds of »8 x 10 5 m/sec have been measured in co-injected discharges,
corresponding to a Mach number of * 0.7. Rotation speeds withcounter-injection are lower than with co-injection at the same
power.
A number of classically expected factors may lead to reduced
density peaking and poorer heating efficiency at high rotation
speeds with unidirectional injection [10]. In the rotating plasma
frame the beam ions, and especially the fractional energy
component beam ions (which constitute about one half of the totalbeam), have considerably reduced energy. This results in larger
atomic cross-sections for beam deposition and a larger fraction of
the deposition on hydrogenic species in the form of chargeexchange rather than impact ionization. Because of the lower beam
ion velocity in the plasma frame, the beam ion slowing down time
and therefore the beam ion density is substantially reduced. In
an extreme example (13 MW of co-injection at I = 0.9 MA withv ^ O ) = 7.5 x 10^ m/sec) taking these effects into account insomewhat simplified calculations results in a 30$ reduction ofcentral electron source rate. The central beam ion density at theend of injection is reduced by a factor of 3. If the presence of84

GOLDSTON et al.

a central peak of circulating beam ions is important forestablishing a sharp electron density profile, it is reasonable toexpect that this sharpness will be reduced in a rapidly rotatingplasma.
At the high Mach numbers we have observed, one expects
significant centrifuging of plasma density, and especially of
impurity density, to the large major radius side of the plasma.
Thomson scattering measurements show a definite trend in thisdirection, correlated with unbalanced injection. Detailed scalingstudies have not been made, but effects of the magnitude expected
are observed. This centrifuge effect should tend to reduce beam
penetration to the center of the plasma.
To determine whether
these classically expected effects of rotation are enough to
explain the differences in the density profiles we observe,
without additional changes in central particle diffusivity, will
require more detailed calculations and examination of the data.
The increased density rise in counter-injected discharges almost
certainly also plays a role in the broad density profiles observedin the more slowly rotating counter-injected plasmas.
Lower beam velocity in the plasma frame results in a strong
reduction in the beam-target and beam-beam fusion neutron yield;
unidirectional injection reduces beam-beam neutron production as
well. In the example above the effects of plasma rotation reducethe calculated neutron flux by a factor of 2, and good agreement
is obtained with the measurement.
The importance of plasma
rotation for neutron production is most clearly illustrated by the
rapid rise in neutron production observed when a low-current
plasma is first heated with 5 MW of co-only injection, and thenthe co beams are switched off and 5 MW of counter beams turned
on. The neutron emission rises by a factor of 2 in 50 msec.
With unbalanced injection a substantial fraction of the beam
power is invested in applying torque to the rotating plasma. Muchof this power -is expected to be returned to the bulk ions via
viscous damping mechanisms in the region where the gradients inrotation speed are high, probably close to the plasma surface. In
this region ion thermal losses through convection and chargeexchange are large however.
Rotational energy may also be
partially lost from the plasma via the same mechanisms. In theexample discussed above, this process, plus the reduction in beam
penetration, results in a 50$ reduction of central heating power,
and an overall HQ% investment of beam heating power in driving
rotation.
If the power is not coupled back to the plasma, thelower total stored energy observed with co-only injection [1] isreasonably well explained. Central ion temperature measurements
in co-only cases are also reduced compared to balanced injection
with similar parameters, consistent with the calculated reduced
central input power to the ions.
Rotational energy is probably
coupled back into the bulk thermal plasma with some efficiency,
and it is thus a mix of this mechanism and the broader density
profiles observed with co-only injection which together result in
reduced stored energy.

IAEA-CN-47/A-II-l
6.

85

Conclusions

A new mode of enhanced energy confinement has been observed
on TFTR. It exhibits improved electron stored energy, as well as
improved thermal and beam ion stored energy. The broad electron
temperature profile shapes observed in this regime at high q a do
not support a concept of profile consistency based on theassumption that T e (0)/<T g > v = f(q a ), but do fit with a model basedon examining normalized profile snapes outside of the plasma core,
independent of q a [6, 11]. The observation that increased
recycling results in a convection-dominated edge electron power
balance, which then correlates with reduced overall confinement,fits well with an understanding of this regime based on profile
consistency arguments [9]. On the other hand it is also a
reasonable hypothesis that the reduction of n e and/or r^ in thecore region plays an important role.
The peaked density profiles observed in this regime appear
to be due to a beam build-up process which results in highly
centralized beam deposition. Density profiles are most peaked withnear-balanced injection, and some classically expected mechanisms
have been put forward to explain this, but it is not clear whether
other effects are required. Neutron production and beam heating
efficiency are also highest with near-balanced injection.
Classical calculations including plasma rotation are adequate to
explain the difference in neutron production rate with co-only andbalanced injection, given the differences in plasma profiles. The
fractional power delivered to the plasma in the form of torque on
a rotating body is comparable to the fractional difference instored energy between co-only and balanced injection. Some ofthis power, however, is probably re-invested in ion heating near
the outside in minor radius. The combination of the resulting
reduction in heating efficiency and the broader density profiles
observed with unbalanced injection together explain the importanceof balanced injection for obtaining the best results in the TFTR
"supershot" regime.
ACKNOWLEDGMENTS
We are grateful to D.J. Grove and J.R. Thompson for theiradvice and support, and to J. Strachan for his many contributions.
This work was supported by US DOE Contract No. DE-ACO2-76CHO3O73.
The ORNL participants were also supported by U.S.
Department of Energy Contract No. DE-AC05-8H0R21400 with Martin
Marietta Energy Systems, Inc.

REFERENCES[1] HAWRYLUK, R.J., et al., Paper IAEA-CN-47/A-I-3, these Proceedings, Vol. 1.
[2] HILL, K.W., et al., Paper IAEA-CN-47/A-IV-2, these Proceedings, Vol. 1.
[3] McGUIRE, K., et al., Paper IAEA-CN-47/A-VII-4, these Proceedings, Vol. 1.

86
[4]
[5]
[6]
[7]
[8]
[9]
[10]
[11]

GOLDSTON et al.
GOLDSTON, R.J., Plasma Phys. Control. Fusion 26 (1984) 87.
WALTZ, R.E., Comments on Profile Consistency, Rep. GAT-A 18365, Goodyear Atomic
Corp., Portsmouth, OH.
FREDRICKSON, E., et al., in Controlled Fusion and Plasma Heating (Proc. 13th Europ. Conf.
Schliersee, 1986), Vol. 1, European Physical Society (1986) 148.
NAGAMI, M., et al., Nucl. Fusion 24 (1984) 183.
POST, D.E., private communication, 1978.
FURTH, H.P., Plasma Phys. Controll. Fusion 28 (1986) 1305.
GOLDSTON, R.J., in Basic Physical Processes of Toroidal Fusion Plasmas (Proc. Course andWorkshop Varenna, 1985), Vol. 1, EUR-10418-EN, CEC, Brussels (1986) 165*.
MURMANN, H.D., et al., in Controlled Fusion and Plasma Heating (Proc. 13th Europ. Conf.
Schliersee, 1986), Vol. 1, European Physical Society (1986) 216.

DISCUSSION
A. BHATTACHARJEE: I have a comment and a question. In the presence of
toroidal rotation, the density on any given flux surface becomes poloidally asymmetric. This, as you know, was observed in PDX. However, the magnitude of toroidal
rotation necessary to explain the asymmetry experimentally observed was found to
be greater than the rotation actually observed, for example, by S. Semenzato et al.
at the Centre de recherches en physique de plasmas, Lausanne. It appears that thesame could be said of TFTR.
My question is: why is it that the rotation velocity observed is different for coinjection and counter-injection when the beam power is held fixed?
R.J. GOLDSTON: On TFTR, as on PDX, we do see some in-out asymmetry
in the Thomson scattering density profiles, even in Ohmic plasmas. We have found,however, that there is a systematic further asymmetry which correlates with
unbalanced injection. The degree of this further asymmetry is roughly in agreementwith what would be theoretically expected, but we have not yet carried out detailed
scaling studies on the asymmetry.
Counter-injection does indeed drive less central rotation speed than coinjection, for fixed power, but the density is higher. We have not yet measured theradial profile of the rotation, so I am hesitant to suggest an explanation.
B. COPPI: I should just like to comment that the lack of dependence ofTe(0)/T<Te>av on q^ when q(r = 0) > 1 was in fact pointed out in the original formulation for the principle of profile consistency. Therefore, the observations you:

The last term in Eq. (26) of this reference should read:2
Em.<jj

+
,,

.

_

„

,t

v

Vp
;

2
_ <R->> ^ _ [<(5P)' > («_)„. (-£)]}

IAEA-CN-47/A-H-1

87

reported do not require a reformulation of this principle. In addition, the temperature
profile may deviate considerably from a Gaussian, depending on a variety of factors,
for example, the presence of trapped electrons, under the same formulation.
R.J. GOLDSTON: Thank you for the comment.
A. GIBSON: Now that you have the concept of profile consistency with somany 'ifs' and 'buts' and restricted it to such a special region of the profile, do you
still believe it has any physical significance, and if so what is it?
R.J. GOLDSTON: Researchers have frequently divided the tokamak into threezones — a central sawtoothing zone, a middle 'confinement' zone, and an edge zone
dominated by convection and atomic physics. The refinement we proposed at the 13th
European Conference in 1986 and which we are discussing here is that the resiliencyof the profile applies mostly in the middle 'confinement' zone. Lower hybrid heating
experiments certainly show that the presence or absence of sawteeth can strongly
attack the core zone, and the H-mode is an example of how the edge can be modified.
The fact that modifications of the edge region can have profound effects on coretransport strongly suggests a transport mechanism which resists profile shapechanges. One example would be a model which includes vT e or d lnT e /dr in theelectron thermal diffusivity itself.

IAEA-CN-47/A-II-2

ENERGY AND PARTICLE CONFINEMENTOF JT-60 PLASMASWITH HIGH POWER HEATING
JT-60 TEAM
(Presented by M. Nagami)
T. Abe, H. Aikawa, H. Akaoka, H. Akasaka, M. Akiba, N. Akino, T. Akiyama,
T. Ando, K. Annoh, N. Aoyagi, K. Arakawa, M. Araki, K. Arimoto, M. Azumi,
S. Chiba, M. Dairaku, N. Ebiswa, T. Fujii, T. Fukuda, H. Furukawa,
K. Hamamatsu, K. Hayashi, M. Hara, K. Haraguchi, H. Hiratsuka, T. Hirayama,
S. Hiroki, K. Hiruta, M. Honda, H. Horiike, R. Hosoda, N. Hosogane, Y. Iida,
T. Iijima, K. Ikeda, Y. Ikeda, T. Imai, T. Inoue, N. Isaji, M. Isaka, S. Ishida,
N. Itige, T. Ito, Y. Ito, A. Kaminaga, M. Kawai, Y. Kawamata, K. Kawasaki,
K. Kikuchi, M. Kikuchi, H. Kimura, T. Kimura, H. Kishimoto, K. Kitahara,
S. Kitamura, A. Kitsunezaki, K. Kiyono, N. Kobayashi, K. Kodama, Y. Koide,
T. Koike, M. Komata, I. Kondo, S. Konoshima, H. Kubo, S. Kunieda,
S. Kurakata, K. Kurihara, M. Kuriyama, T. Kuroda, M. Kusaka, Y. Kusama,
S. Maehara, K. Maeno, S. Mase, S. Matsuda, M. Matsukawa, T. Matsukawa,
M. Matsuoka, N. Miya, K. Miyati, Y. Miyo, K. Mizuhashi, M. Mizuno, R. Murai,
Y. Murakami,
M. Muto,M. Nagami,
A. Nagashima,
K. Nagashima,
T. Nagashima, S. Nagaya, H. Nakamura, Y. Nakamura, M. Nemoto, Y. Neyatani,
S. Niikura, H. Ninomiya, T. Nishitani, H. Nomata, K. Obara, N. Ogiwara,
Y. Ohara, K. Ohasa, T. Ohga, H. Ohhara, M. Ohkubo, T. Ohshima, K. Ohta,
M. Ohta, M. Ohtaka, Y. Ohuchi, A. Oikawa, H. Okumura, Y. Okumura,
K. Omori, S. Omori, Y. Omori, T. Ozeki, M. Saigusa, K. Sakamoto, A. Sakasai,
S. Sakata, M. Satou, M. Sawahata, M. Seimiya, M. Seki, S. Seki, K. Shibanuma,
R. Shimada, K. Shimizu, M. Shimizu, Y. Shimomura, S. Shinozaki, H. Shirai,
H. Shirakata, M. Shitomi, K. Suganuma, T. Sugie, T. Sugiyama, H. Sunaoshi,
K. Suzuki,
M. Suzuki,
M. Suzuki,
N. Suzuki,
S. Suzuki,
Y. Suzuki,
H. Takahashi1, M. Takahashi, S. Takahashi, T. Takahashi, M. Takasaki,
M. Takatsu, A. Takeshita, H. Takeuchi, S. Tamura, S. Tanaka, K. Tani,
M. Terakado, T. Terakado, K. Tobita, T. Tokutake, T. Totsuka, N. Toyoshima,
H. Tsuda, T. Tsugita, S. Tsuji, Y. Tsukahara, M. Tsuneoka, K. Uehara,
M. Umehara, Y. Uramoto, H. Usami, K. Ushigusa, K. Usui, J. Yagyu,
K. Yamada, M. Yamamoto, O. Yamashita, Y. Yamashita, K. Yano, T. Yasukawa,
K. Yokokura,
H. Yokomizo,
H. Yoshida,
M. Yoshikawa,Y. Yoshinari,
R. Yoshino, I. Yonekawa, K. Watanabe

Naka Fusion Energy Research Establishment,
Japan Atomic Energy Research Institute,Naka-machi, Naka-gun, Ibaraki-ken,
Japan

1

Princeton Plasma Physics Laboratory, Princeton University, NJ, USA.

89

90

JT-60 TEAM
Abstract

ENERGY AND PARTICLE CONFINEMENT OF JT-60 PLASMAS WITH HIGH POWERHEATING.
Results are reported from initial confinement experiments of divertor and limiter plasmas, carrying a current of up to 2 MA and heated by neutral beam power up to 20 MW, over a wide range ofdensities (0.3-8) x 1019 m~3, and toroidal fields, 1.9-4.5 T. Preliminary experiments on the concurrent application of NB heating and ICRF or LHCD are also described. The plasma stored energy has
an offset linear relation with the absorbed power, and the energy confinement time (in seconds) scalesas (0.051 + 0.35/P abs [MW]) X I [MA] 0 4 . The energy and particle confinement times have a similar
dependence on the absorbed power. Improved particle and energy confinement is discussed for NB,
NB + ICRF, and NB + LHCD discharges.

1.

INTRODUCTION

This paper discusses confinement results from the initial auxiliary heating
experiments. A companion paper [1] provides a review of the JT-60 programme andits experiments. Another companion paper [2] describes impurity and MHD effects.
The additional heating and current drive programme started in early August, using
either NB heating, ICRF second harmonic (120 MHz) heating or LHCD (2 GHz).
The injected NB power (Pinj) was raised to 20 MW, the ICRF power to 1.4 MW andthe LHCD power to 1 MW, within two to three months from the start of theexperiments.
The particle confinement and the energy confinement of NB heated normal discharges are discussed in Sections 2 and 3, respectively. The regimes of improved
confinement are described in Section 4. Conclusions are presented in Section 5.
2.

PARTICLE CONFINEMENT

The line integrated Ua emission intensity was measured along a chord in thepoloidal plane at four different locations each in the main and divertor chambers.
Figure 1 shows the time evolution of ne and four channel Ho signals for an 8 MW
NB heating divertor discharge. With NB injection the main plasma density started
to decrease, and the H a intensity in the divertor chamber increased by a factor of
5 to 20 from its Ohmic level while the intensity only rose by a factor of two to three
in the main chamber. This suggests an enhanced scrape-off plasma flow from themain chamber into the divertor chamber during NB injection.
On the assumption of toroidal symmetry, the global particle confinement time
T can be determined from an absolute intensity measurement of the H a emission at
one toroidal location. This assumption is believed to be valid for divertor discharges.
For Ohmic discharges, rp was found to be proportional to n^ 7 and was 70 ms at
ne = 3 x 1019 m-3 (Fig. 2(a)). The rp degraded with increasing Pinj, and, at 20 MW,
was lower than its Ohmic values by a factor of three.

91

IAEA-CN-47/A-II-2o5.0
10.0 Q_
0.0

0.0
0.5
0.0
C4

•E
_c
a.

0.5
0.0

at

"O
5.0
o

0.0

20.0
0.0
3.0

5.0

7.0

Time (s)FIG. 1. Evolution ofne and four channel Ha signals for NB heated divertor discharge.

The hydrogen neutral pressures measured in the main and divertor chambers areplotted against ne_in Fig. 2(b) for Ohmic and NB heated discharges. The pressure
ratio, Pdiv/Pmain, which is sometimes regarded as an indicator of efficient divertor
function, was 30 to 40 for Ohmic discharges, but increased up to 40 to 60 for discharges heated with 10 to 20 MW of NB power. The measured Pdiv was proportional to n^. This dependence, together with the experimentally determined nedependence of rp, leads to a conductance of the neutral particle backflow from thedivertor to the main chamber which is nearly independent of ne. The conductance,
consistent with the measured Pdiv and rp, was about 900 m3-s"' and was one-half to
one-third of the calculated mechanical conductance. This difference appears reasonable when the presence of the scrape-off plasma that obstructs the backflow is taken
into account. For a constant backflow conductance, the normalized divertor pressure,
Pdiv = Pdiv/n£, is inversely proportional to rp. The dependence on the absorbed
power (Pabs) (sum of Ohmic and injected beam power minus shinethrough loss) of
J/P d i v was similar to that of the energy confinement time (Fig. 2(c)). This resemblance suggests a correlation between the increased scrape-off plasma flow and thedegradation of energy confinement.

JT-60TEAM

92

I

8" 3"
go

1

<J

-a.

I 8"
O

JH

d

s

a

• 4

z
X
oa <

o
i

a

q in O
C«J

Q.

•4

4

a
a

"i

D

m

•
a

a •

m

o

U

D
O

o

o

o

o

o

o

o

o

^

N

O

ffl

(0

^

N

80

IAEA-CN-47/A-n-2

93

(MJ)

W

Ip(MA)
2.0
1.5
1.0

D

L

n

•

A

0

•

1
2
(0.051 Pabs(MW) +0.35) Ip(MA) 04
FIG. 3. Plot of W against (0.05 IP^

3.

[MW] + 0.35) I [MA]0-4.

ENERGY CONFINEMENT

Hydrogen neutral beam power of up to 20 MW at energies of 50 to 75 keV,
was injected into hydrogen or helium plasmas. The fraction of hydrogen in the (nominally) helium plasmas was estimated to be up to one-half, from H a emission intensity, beam shinethrough loss and neutral hydrogen flux measurements.
In divertor discharges roughly 80% of the absorbed power can be accountedfor. Energy balance in the divertor chamber is accomplished through a bolometer
array used to measure the radiation loss and a thermocouple array utilized to measure
the divertor plate heat flux. The measurement showed that the divertor chamber radiation accounted for about 20%, the divertor plate heat deposition for about 50% andthe main chamber radiation for less than 10% of the absorbed power.
3.1. Energy confinement scaling
The plasma stored energy was evaluated by both kinetic and magnetic means.
The kinetic determination of the stored energy was based on a six point Thomson
scattering profile of the electron temperature and the spectroscopically determined
central ion temperature. When the calculated fast ion energy was added to the kinetically determined stored energy, the sum agreed with the magnetically determined
stored energy (W*) within 20%.
The value of T*E = W7P abs was the same for hydrogen and (nominally) helium
discharges and did not depend on the density. A toroidal field (Bt) scan from 1.9 to
4.4T at Ip = IMA did not yield any obvious dependence of T*E on Bt, or on qeff,

94

JT-60 TEAM

X e (-§-a) (rr Vs)

7«-,Ip=l.5MA

•
•

•

4 -

•
3

OH

•

•

2
1

•

•

(b)

1
...

5

10

-

i

15

20

Pob, (MW)

^eff

1.3

DIVERTOR. Hi
n,= 4 . 5 - 5 x 1 0

o

-

o

0

0°

o

O

o °

Hero)

1.0
1.0

D
(rrfc)
o
oo

0.5

o
o

o
o

o

o

(c)

i

10

15

20

Pabs (MW)FIG. 4. (a) Plot of Te(0)/(Te) and rs/a against llqegfor ohmically heated divertor and limiter
discharges, and NB heated divertor and limiter discharges; (b) plot of electron thermal diffusivity, Xe,at r = 2a/3 against P a i j ; (c) plot of ne(0)/ne and particle diffusivity (D) against Pabs.

IAEA-CN-47/A-II-2

95

which varied from 2.2 to 6. Under similar conditions, the W* of the limiter discharges was about the same as that of the divertor discharges. The discharges canbe represented by an offset linear relation, r* [s] = (0.051 + 0.35/Pabs [MW])/Ip[MA]0-4, which results in an increased confinement time of 50 to 70 ms for
Ip = 1 to 2MA (Fig. 3). This relation has a weaker Ip dependence than the
Kaye-Goldston scaling.
3.2. Energy transport analysis
The ratio of the central to the volume averaged electron temperature,
Te(0)/<Te>, and the ratio of the sawtooth inversion radius to the minor radius, rs/a,are plotted against l/qeff in Fig. 4(a). These ratios varied little between Ohmic and
NB heated discharges, or between divertor and limiter discharges. This fact may be
an indication that the electron temperature profile and the current profile in the coreregion did not vary substantially.
The energy transfer from the fast ions to the plasma electrons and ions was estimated as a function of the magnetic surface co-ordinate using Monte Carlo beam
deposition and fast ion orbit/slow-down calculations. The main ion energy loss channel was the energy transfer to the electrons, and the ion heat conduction played aminor role in the ion energy balance.
The electron heat conduction dominated the energy confinement in the plasmacore. The electron thermal diffusivity (Xe) at r/a = 2/3, calculated by the transport
analysis assuming that X: is three times the Chang-Hinton value, was 1-2 irf-S"1 for
Ohmic discharges and increased by a factor of three to five for discharges heated with
20 MW of NB power (Fig. 4(b)). The ratio of the central to the line averaged density,ne(0)/ne, is plotted against Pabs in Fig. 4(c). This ratio, which is around 1.2, indicates a very flat density profile, nearly independent of the NB power. The particle
diffusivity (D) at r = 2a/3, deduced from NB particle fuelling and density profile
evolution, increased with the increase of absorbed power. The convective energy
flow made a contribution of approximately 20% in the plasma core, and was estimated to be comparable to the conduction loss in the peripheral regions.

4.

IMPROVED PARTICLE AND ENERGY CONFINEMENT DISCHARGES

There were three different sets of conditions in which an improvement in either
particle or energy confinement was observed.
The first set of conditions was found during an effort made to produce anenvironment which was considered conducive to the H-mode, based upon theexperience from medium sized devices: increasing Pinj to 20 MW, lowering Bt to
1.9 T, and qeff to 2.2, maintaining a large separatrix-limiter distance up to 5 cm,
maximizing the volume averaged power density by reducing the plasma volumeJT-60TEAM

96(VW) "l

MW) d

c

« S

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111

|j|

00

7
3

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§

IAEA-CN-47/A-II-2

91(14 MW/31 m3), the use of helium instead of hydrogen, fuelling through the divertor chamber, etc. These efforts produced no clear, reproducible H-mode discharges.
The example shown in Fig. 5(a) possessed, however, most of the characteristics commonly associated with the H-mode. The discharge was a small minor radius(0.71 m) plasma heated with Pinj = 18 MW (Pabs = 14 MW) and showed an abrupt
reduction of the Htt emission in the divertor chamber (at 0.37 s into the beam
heating phase) and a simultaneous increase in ne from 1.6 to 2 X 1019 nr 3 , in A/3p
by 20 %, and in the radiative power from 1.4 to 4.8 MW. Furthermore, this discharge
had the best T'E among discharges with a similar minor radius. The increase in j3p
was, however, insignificant and only a few similar discharges have been obtained todate. Many discharges with similar parameters that can be set by the operator resulted
in bursts of radiation, originating from the molybdenum discharge contamination
beginning at about the time when the H a reduction would have happened.
Reduction of particle recycling in the divertor region was observed during ICRF
injection into Ohmic discharges and NB injected discharges with the injection power
not greatly exceeding the ICRF power. The second example shown in Fig. 5(b) is
a 1.5 MA divertor discharge with NB and ICRF heating. The increased H a emission
in the divertor with the NB injection turned out to diminish with the start of the .ICRF
injection, and at the switchoff of the ICRF injection the emission started to increaseagain, suggesting a reduction of plasma flow into the divertor chamber during ICRF
injection. The behaviour of H a and ne signals during ICRF injection is totally different from that in NB heated discharges. The increased energy confinement time of
70 ms during the initial NB injection was extended by 40% at the end of the ICRF
injection.
As the third example, two discharges heated with NB are compared in Fig. 5(c):one (shown by broken lines) was a normal, inductively driven discharge and the otherone (solid line) was a discharge driven entirely by LHCD. Both discharges wereoperated at Bt = 4T, L, = 700 kA, with no gas feed (prefill only); they were heatedby 40 keV, 8.1 MW NB power. A dramatically increased nc level in the LHCD discharge over the inductively driven discharge indicates improved particle confinement. The three chord density profile measurements showed substantially a peaked
ne profile (ne(0)/ne = 2.1), suggesting Tp improvement in the plasma core. Also, theincreased energy confinement time was raised by 30%. The improvement may berelated to the suppression of sawtooth activities and/or a change in current profile.
The internal inductance fell by 0.2 from the Ohmic heating level, indicating a flattening of the current profile by LHCD.

5.

CONCLUSIONS

Energy and particle confinement was studied, by taking full advantage of efficient divertor functions of the JT-60 device that allowed operations over a wide density range by controlling the particle exhaust and impurity contamination, and by

98

JT-60 TEAM

utilizing three different means of additional heating. The divertor was particularly
useful in allowing operations at extremely low densities at which LHCD had tooperate.
The availability of three different means of additional heating, used individuallyand in various combinations, allowed investigations of the confinement properties
that lead to operating modes with improved confinement.

ACKNOWLEDGEMENTS
The authors wish to thank the members of JAERI who have contributed to the
JT-60 project throughout its progress. They also wish to express their gratitude to
Drs S. Mori, Y. Iso, and K. Tomabechi for continued encouragement and support.

REFERENCES[1][2]

YOSHIKAWA, M., Paper IAEA-CN-47/A-I-1, these Proceedings, Vol. 1.
JT-60 TEAM (presented by H. Takeuchi), Paper IAEA-CN-47/A-IV-3, these Proceedings,
Vol. 1.

DISCUSSION
F. WAGNER: The H a radiation in the divertor chamber reflects the local ionization equilibrium fuelled by the power flow into it. H a can transiently drop when
the plasma confinement changes. On the other hand, H a also drops when the power
flow into the divertor chamber is reduced by enhanced impurity radiation in the main
plasma chamber. To what extent is the H a intensity reduction which you observed
during ICRH caused by enhanced impurity radiation?
M. NAGAMI: Although the radiation loss in the main plasma, P^ i n , increased
during ICRF, the power flux into the divertor, Pt0,-P£din> also increased. Furthermore, it was experimentally demonstrated that an impurity (Ne) injection into an
Ohmic discharge or a neutral beam heated discharge does not produce the same H a
reduction as that observed during ICRH.
D.E. POST: Did the divertor operate in a 'high recycling' regime with highneutral compression at the low densities used for lower hybrid current drive?
M. NAGAMI: Neutral particle pressure and recycling in the divertor chamber
are low for the low density discharges with lower hybrid current drive.

IAEA-CN-47/A-D-3

ENERGY CONFINEMENT IN JET
WITH OHMIC HEATING AND
STRONG AUXILIARY HEATING
J.G. CORDEY, D.V. BARTLETT, V. BHATNAGAR1, R.J. BICKERTON,
M. BURES,. J.D. CALLEN 2 ,D.J. CAMPBELL,
C D . CHALLIS,
J.P. CHRISTIANSEN,
S. CORTI,
A.E. COSTLEY,
G.A. COTTRELL,
G. DUESING,
J. FESSEY,
M. GADEBERG3,
A. GIBSON,
A. GONDHALEKAR,
N.A. GOTTARDI,
C.W. GOWERS,
M. VON HELLERMANN,
F. HENDRIKS,
L. HORTON,
H. JACKEL 4 ,J. JACQUINOT, O.N. JARVIS, T.T.C. JONES, E. KALLNE, J. KALLNE,
M. KEILHACKER, S. KISSEL, L. DE KOCK, P. LALUA, E. LAZZARO,
P.J. LOMAS,
N. LOPES-CARDOZO5,
P.D. MORGAN,
P. NIELSEN,
J. O'ROURKE, R. PRENTICE, R.T. ROSS, G. SADLER, F.C. SCHULLER,
A. STABLER4, P. SMEULDERS, M.F. STAMP, D. STORK, P.E. STOTT,
D.R. SUMMERS,
A. TANGA,
P.R. THOMAS,
E. THOMPSON,
K. THOMSEN3, G. TONETTI 6 , B.J. TUBBING5, M.L. WATKINSJET Joint Undertaking,
Abingdon, Oxfordshire,
United Kingdom

Abstract
ENERGY CONFINEMENT IN JET WITH OHMIC HEATING AND STRONG AUXILIARYHEATING.
The energy confinement properties with strong auxiliary heating are found to be well describedby an offset linear law relating the stored energy W to the input power P (W = a P + 0). The scalingof the energy confinement time with the non-dimensional plasma physics parameters suggests that theunderlying mechanism is resistive MHD. The heating profile and impurity radiation are identified as
making a substantial contribution to the degradation in the global energy confinement time with auxiliary
heating.

1

2
3

4
5
6

Plasma Physics Laboratory, Association "Euratom-Belgian State" ERM/KMS,
Brussels, Belgium.
Department of Physics, University of Wisconsin, Madison, WI, USA.
Risa National Laboratory, Roskilde, Denmark.
Max-Planck-Institut fur Plasmaphysik, Garching, Fed. Rep. Germany.
FOM Instituut voor Plasmafysica, Nieuwegein, Netherlands.
Centre de recherches en physique des plasmas, Ecole polytechnique fid6ralede Lausanne, Lausanne, Switzerland.

99

100
1.

CORDEYetal.
INTRODUCTION

In this paper the energy confinement properties of the JET
plasma with intense auxiliary heating are discussed. In
previous papers [1],[2], confinement in ohmically heated plasmas
has been reported. The additional heating was provided byneutral beams (60-75keV D and H) at powers up to 9MW and ion
cyclotron resonance heating [3] of helium and hydrogen
minorities at powers up to 7MW. The two heating methods werealso combined to give a total input power of 16MW which was
between 4 and 20 times the ohmic input power. The toroidal
field B, plasma current I and average density <n> were allvaried over a wide range (B = 1.7 - 3.4 T, I = 1-5MA, <n> = 0.5- 5 x 10l9rn~3), with the geometry kept essentially fixed (major
radius R - 3.0m, horizontal minor radius a = 1.2m, elongation K= 1.45) so that the dependence of the global energy confinement
T (= W/P-ft) upon I, B, <n> and the heating scheme could beassessed.
The plasma edge conditions have, however, been varied bylimiting the plasma on the outer mid-plane limiters, the innerwall or separatrices (X-points) near the top and bottom of thechamber. Generally, the confinement time T £ degrades with bothtypes of heating from its ohmic value - 0.5 - 0.8s down to~ 0.2-0.4s. The smallest degradation occurs for plasmas locatedon the inner wall or with an X-point.
The remainder of the paper is divided as follows: insection 2.1 a simple empirical scaling law describing the
behaviour of the T with the plasma parameters is derived andthen in section 2.2 the scaling of T ? with the Connor-Taylor [4]dimensionless constraint parameters is determined. In Section 3
the degradation in confinement is discussed. First the questionof whether the heat flux is self-regulating to maintain a fixed
electron temperature profile or whether it is determined by the
local temperature gradient is examined. The latter model isfound to be more consistent with the data. Then a procedure isgiven for correcting the input power for impurity radiation
losses and heating profile effects, and it is shown that thesecan account for a substantial part of the degradation in
confinement.
2.1. Scaling of t £ with Plasma Parameters
The plasma confinement properties with additional heating
can be most clearly seen by exhibiting the data in a total(diamagnetic) energy W versus total input power P diagram suchas that shown in Fig. 1 where examples of power scans withcombined heating at four different currents and toroidal fields
are shown. The data in Fig. 1 can be represented by an offset
linear scaling of W with P in the form:

101

IAEA-CN-47/A-II-3... * 5 M A

*
*

4MA

r4

,-2 MA
X.... 1 MA* • • '

s"'

jf'''
10

20

P(MW)
F/G. 7. Total stored energy versus total input power for four different power scans:
I = 5MA.B = 3.4 T; 1 = 4 MA, B = 3.4 T; I = 2 MA, B = 2 T; I = 1 MA, B = 2 T.

W = W(O)(1)

where the incremental confinement time T-^nc and VJ(O) are functions of the I, B, <n>, etc. To make the representation fully
general a dependence on power P has to be included in T:... The
reason for this is that, although the data in Fig. 1 appear
linear with P, the density <n> is increasing with P in all thepower scans and this could be masking the true power dependence.
However, by combining data from scans at different starting
densities, we have been able to show that the density dependence
in i i n c is weak. A good fit to the NBI data set has been foundby use of standard regression techniques; it is:W(0) = 0.225 n 0 ' 6 I 0 - 5 B 0 '" and T l n c = 0.047*1 (Units: MJ,
m" 3 x 1 0 ~ 1 9 , MA, T ) . The complete data set (NBI, ICRH andcombined heating) is shown against this fit in Fig. 2.
The scatter in the data in Fig. 2 is thought to be mainly
due to the method of heating (NBI or ICRH) and the type ofconfinement, limiter or inner wall. It is found that thefraction of the input power radiated by the impurities issignificantly larger for ICRH and limiter confinement than for
inner wall with NBI heating.
In Fig. 3 the radiation profiles are shown for threepulses: (A) a typical ohmic discharge. (B) an NBI discharge
located on the inner wall, (C) a combined heating discharge on
the limiter. Typically 30? of the input power is radiated in
the central confinement zone (r/a < 0.7) in case (C) andonly &% in case (B); the source and concentration of the102

CORDEY et al.

FIG. 2. Total stored energy versus the fit given in Eq. (]). The symbols are: • — neutral beam injection heating alone, A — including ICRH with H minority, and o — including ICRH with He minority.

0
FIG. 3. Radiation profiles for: (A) an ohmic discharge located on the inner wall, (B) an NBI heated(9 MW) discharge located on the inner wall, and (C) a discharge with combined heating (4 WM of ICRH
and 5 MW of NBI).

impurities is discussed in detail in ref. [5]. The radialheating profiles of the two schemes are also very different:
the ICRH profiles are always peaked, ie. similar to the ohmic
heating profile, whilst the NBI profile is usually less peaked,
and at high densities the NBI profile is quite flat.

IAEA-CN-47/A-II-3

103

(0

P,o, <MW)
FIG. 4. Energy confinement time versus total power. The symbols are: x — X-point with double null,
A — X-point with a single null, * — inner wall discharges, and D — limiter discharges. The solid lineis the Kaye-Goldston fit.

The effect on confinement of changing the plasma
configuration is shown in Fig. 4, where the global T (diamagnetlc) is given as a function of power for the complete 2MA data
set; also shown on the same figure is the empirical L-mode
scaling of Kaye and Goldston [6] derived from data from previous
experiments. Although for all configurations and heating
schemes the T £ degrades with power, the inner wall and double
null X point data with NBI heating are up to a factor 2 above
the Kaye-Goldston law. The single null X point data which had
clear H mode characteristics was up to a factor 3 larger thanthe Kaye-Goldston law.
The majority of the NBI experiments have been made with
deuterium injection in the co-direction. A few experimentswith counter-injection using deuterium and also hydrogen
injection in the co-direction have also been made. As far as
energy confinement was concerned, there was no apparent
difference between co-and counter-injection, and the hydrogen
injection results which were reported earlier [2] were similar
to the present deuterium injection data.
2.2. Scaling of T F with Dimensionless Parameters
The additional heating data have been compared with theConnor-Taylor models [4]. It is found that the low beta models
cannot represent the confinement time scaling, especially when

104

CORDEY et a).
0.6 ITn

0.4 .*!. , 1 1

n

n

•

D

+
b<
it*

1

Jtpp "'"

0.2 -

00 -

1 1 1 1 1 1-ri-rr
1 IT T

0.0

l I I I •i m

T

T

+

+

X
*
D
0

X
*
0
0

i i i

h Comb DC H )< C o m b D ( HE )
NBD
] NBH
RFH

-I-

0.4

0.2

'
0.6

-exp
FIG. 5. Energy confinement time versus the fit from the resistive MHD model.

comparison is made with data from smaller devices. The scaling
invariants originally given by Connor and Taylor were chosen
arbitrarily to represent density, temperature and magnetic fielddependences. For the high beta models, it is more appropriate
to use the magnetic Reynolds number (S = T R / T ^ ) , the cyclotron
frequency times the poloidal Alfve'n time (w c T Ap V, and the ratioof total to ohmic powers (ptot'/pohm^* T h e l a " e r variable isequivalent to the poloidal beta in power law scaling forms andis used in order to prevent the energy content from appearing inboth the dependent and independent variables. This set of
scaling variables make the relationship between models moretransparent and ensures that geometrical quantities such as
safety factor and elongation are included in a natural fashion.
The energy confinement time for the collisional high beta modelis,P

tot / P ohiJ(2)

The collisionless models are obtained by combining the lastarguments as (Ptot /p ohm^ S ~ 1 a n d t h e f l u i d models by droppingo)cTAp. It is found that the JET data for (P t o t /P o h m )>2 are best
represented by the resistive MHD model asT £ = 0.91 T A p-0.5 Q 0.75
(Ptot/Pohm)(3)

The fit is shown in Fig. 5. Data from other machines can be
reproduced when the aspect ratio ( E = a/R) is included and theresulting scaling law is equivalent tor

R T Ap

0-5

IAEA-CN-47/A-II-3

105

In view of the surprising simplicity of this form and theabsence of any significant normalising constant, it might beconcluded that Eq.(4) is indicative of the underlying transportmechanism. However, this should be qualified by noting thedifficulty in varying S and 3 p in such a way as to clearly
confirm the separate dependences in a given device. At firstsight, Eq.(3) appears to be inconsistent with the offset linear
scaling law Eq.(1). In fact, the implicit dependence of P o h m andS on power make the two indistinguishable.

3-

DISCUSSION OF CONFINEMENT DEGRADATION

There are many possible explanations for the degradation in
confinement and some of these are examined briefly in thissection. A current popular explanation is that the heat transport is self-adjusting such that the same electron temperature
profile is maintained ('profile consistency 1 ). The evidence for
profile consistency in JET data was reviewed by Bartlett et al.
[7] recently. The electron temperature profiles in highcurrent (I S 3MA) discharges certainly exhibit the conventional
profile consistency features, that is T e (p)/T e (0) is only afunction of q and the profile is unchanged by additional
heating, at least at the presently available power levels(P t o t /P o h n ] ~ **) • I n l o w current (I = 1MA) discharges, whereP t o t / P Q h m - 20, the electron temperature profile appears to
respond to the changing heating profile. In Fig. 6(a), thetemperature profiles are shown for the two cases of low and highcurrent before and during heating,and the corresponding heating
profiles are shown in Fig. 6(b).
In the low current case, theinitial ohmically heated profile is substantially broadened by
the application of NBI heating, whilst for the higher current
pulse the ohmic and NBI heating profiles are rather similar.
Although the change in both temperature profiles is small withadditional heating, it is consistent with what is expected forthe change in heating profile shown in Fig. 6(b).
Indeed,asWaltz et al. [8] have shown, very large changes in the heating
profile only produce small changes in the temperature profile
shape for discharges in which the central q is limited byinternal relaxations (sawteeth).
Thus, in summary, there is no really clear evidence that'profile consistency1 plays a deterministic role in heattransport. There is, however, evidence that whilst the confinement time degrades with power input, the underlying heat transport is independent of the auxiliary heating power. First, thefact that the total energy W increases linearly with power P, as
shown in Fig. 1, means that the incremental confinement timeT
inc = AW/AP does not vary significantly with power level. For
example, at I = 3MA we have x i n c = ab/i(Tinc-2.9 +, 0.6m2/s.

106

CORDEY et al.
0.0

-1D-

\
\

-2.0-

\
\
\

-3D-_

QO

\
\

\
\

_

05

1

I.U

'

1

(b)

p
(x10 5 Wm 3 )

\
\
\
\
\

n

\

B

-

\ X^
< ^>
1.0

FIG. 6. (a) Loge (Te(p)/Tt(0)) versus p, for a 3 MA discharge before heating (curve A) and after
9 MW ofNBl heating (curve B), and fora 1 MA discharge before heating (C) and during 9 MW ofNBl
heating (D). (b) Power input to the electrons versus p for the four cases shown in (a).

Second, as shown in Fig. 7, the electron heat flux per unit
density, Pe(r)/nA = / d3xIQ/n/ dA, is proportional to the local
temperature gradient. This implies a Xfiux =A(Pe(r)/nA)/A(dTe/dr) - 1.8 - 3-3m2/s for the 3MA case. (Thereason for the large range in xe is due to the uncertainty inestimating the loss through the ion channel.) Finally, and most
locally, the XHP determined using the heat pulse propagation
method [9] is found to be approximately constant (for I = 3MA»
a = 1.2, K = 1.45) at a value of 2.4 +_ 0.7mVs, independent of
the heating power [10]. The fact that these three different
methods of determining the electron heat diffusivity x e yieldnearly the same power-independent value (xinc ~ Xfiux ~ Xfjp)
implies that the transition from ohmically to auxiliary heated107

IAEA-CN-47/A-D-3

Xe = 1-8 - 3.9 m2/s

201

« , • XHP = 2.4 ± 0.7

A 101
V
c

0

25

50

Heat flux (kW/m 2 )FIG. 7. n< V r,> at p = 0.75 versus heat flux for a 3 MA power scan (o) and a 1 MA power
scan (M). The dotted lines are the best straight line fits to the data.

discharges may be interpretable in terms of a transport modelwith electron heat conduction characterised by a constant valueof x e •An example of such a model will now be developed and used
to clarify the heating profile and radiation loss effects.
First, assume that the heat flux is given generally by heat
conduction plus convection in the form <} = -nxvT + 33cco n v Utilising this in the equilibrium heat balance,
7-q = EQ = Q 0 H + Q a u x - Q rad ,and assuming that x and q c o n v are
not strongly dependent upon T or VT, one can integrate once todetermine the temperature gradient and once more to calculate
the temperature profile. A further integration over the plasma
volume gives the total stored energy W, which in cylindricalgeometry has the formaW = 3/d3x n(r)[/ - ^ /rdrZQ(r)r rnx o{[Pohmaux/dr'r- Pconv(5)Here, T = ab/4x,and the effective power inputs take account of
the effects of heating, radiation, density and x profiles on the
total stored energy W. For constant n and x> o n e obtainspeff = c o / rdl "(1~ r2 /a 2 )Q(r), where C Q is a normalising constant.

108

CORDEY et at.

W (MJ)

~t

*o

3

°»

1

"T
10

15

Pen
F/G. 8. Total energy density versus effective power.
Symbols: it = ohmic heating, • = NBIonly, A = ICRHonly, o = combined ICRH and NBI. Opensymbols: before correction for radiation losses and heating; solid symbols: after correction for radiation; dotted symbols: after correction for radiation and heating profile effects.

The main reason for deriving Eq.(5) is that it has the same
form as the linear offset model Eq(1) used to represent thedata, the offset being identified as the convected power.
However, it would be negative in this model, indicating a heat
pinch [11]. An alternative heat transport model without a heat
pinch (q0) would require a x which, while constant overthe usualconv
power input (or VT) range,varies strongly at low
values of VT [12],[13]. From Fig. 7, either of these models
could fit the present data.
An example of the use of Eq.(5) is shown in Fig. 8,where theabove profile analysis techniques are applied to an I = 3MA,
B = 2T, power scan with combined heating. The data is shown
before correction for the power lost by radiation (openpoints) and then after correction for the power lost byradiation (solid points). Correcting for radiation losses
reduces the scatter in the data and increases the incremental
confinement time by 30%. Correcting for the differing heat
deposition profile further increases the incremental confinement
time (arrowed points) and the degradation in confinement ingoing from ohmic to strong auxiliary heating is then muchsmaller. Now the results in Fig. 8 will depend on theassumptions that are made for the radial profile of x- However,this example does demonstrate qualitatively that the degradationin the confinement time T £ could be due to the broader

IAEA-CN-47/A-II-3

109

deposition profile of NBI and/or enhanced radiation, and achange in the basic heat transport mechanism may not be
required.

4.

CONCLUSIONS

A simple empirical scaling law, Eq.(1 ), has been found toadequately describe the energy confinement properties of JET
with intense additional heating. The scaling of the energy
confinement time with the dimensionless plasma physics
parameters suggests that the underlying heat transport is aconsequence of resistive MHD instabilities. Electron heattransport with auxiliary heating has been found by threedifferent methods to be describable in terms of a heat flux witha constant coefficient (x e ~ 2.5 + 1 .OmVs for 3MA discharges).
A substantial contribution to the degradation in global
confinement with additional heating is shown to be due to the
profiles of radiation and heating.

REFERENCES
[1]
[2]
[3]
[4]
[5]
[6]
[7]
[8]
[9]
[10]
[11]
[12]
[13]

CORDEY, J.G., et al., in Controlled Fusion and Plasma Physics (Proc. 12th Europ. Conf.
Budapest, 1985), Pt. I, European Physical Society (1985) 26.
THOMSEN, K., et al. in Controlled Fusion and Plasma Heating (Proc. 13th Europ. Conf.
Schliersee, 1986), Vol. 1, European Physical Society (1986) 29.
JACQUINOT, J., et al., Paper IAEA-CN-47/F-I-1, these Proceedings, Vol. 1.
CONNOR, J.W., TAYLOR, J.B., Nucl. Fusion 17 (1977) 1047; Phys. Fluids 27 (1984) 2676.
BEHRINGER, K., et al., Paper IAEA-CN-47/A-IV-l, these Proceedings, Vol. 1.
KAYE, S.M., GOLDSTON, R.J., Nucl. Fusion 25 (1985) 65.
BARTLETT, D.V., et al., in Controlled Fusion and Plasma Heating (Proc. 13th Europ. Conf.
Schliersee, 1986), Vol. 1, European Physical Society (1986) 236.
WALTZ, R.E., et al., Rep. GA-A-18365, GA Technologies Inc., San Diego, CA; to be published in Nucl. Fusion.
CALLEN, J.D., JAHNS, G.L., Phys. Rev. Lett. 38 (1977) 971.
TUBBING, B.J.D., et al., Heat Pulse Propagation Studies in JET, Report JET Joint Undertaking,
Abingdon (to be published).
FREDRICKSON, E.D., et al., Nucl. Fusion 26 (1986) 849.
REBUT, P.H., BRUSATI, M., paper presented at 12th Europ. Conf. Budapest on Controlled
Fusion and Plasma Physics, 1985.
THOMAS, P., paper presented at 12th Europ. Conf. Budapest on Controlled Fusion and Plasma
Physics, 1985.

DISCUSSION
R.J. GOLDSTON: In your model, heat flow is a strictly defined offset linear
function of r^ v Te. How does this model deal with the heating profile experiments

110

CORDEYetal.

from T-10, ASDEX, TFTR and JET, and how does it cope with the apparent sensitivity of the confinement to edge parameters, for instance in the H-mode?
J.G. CORDEY: It has been shown by Waltz et al. (paper IAEA-CN-47/
A-VI-1-3, these Proceedings) that nx V T can produce 'profile consistency' for Te
profiles. In the H-mode the magnitude of x in the edge region is greatly reduced andthis accounts for the improved confinement.
H. HORA: With regard to your discussion of the inhibition of thermal conduction, have you included in this the effects of electric (inhomogeneity) fields anddouble layer1, since there they provide an explanation for pellet fusion2 and since
the generally predicted fields3 have now been measured in toroidal plasmas?
J.G. CORDEY: No; this model is only in its infancy and these effects have notbeen taken into account.
B. COPPI: I do not understand how a theory of purely resistive modes can beapplicable to low collisionality regimes (multi-keV), as you have in JET. Could youcomment on that?
J.G. CORDEY: Yes, I would expect this theory to apply equally to the low collisionality regimes; however, the data certainly do not appear to fit them. There are
two possible explanations: one is that the true dependence of r t on the plasma
parameters is being masked by power deposition and radiation profile effects; theother is that the non-linear development of the general kinetic theory leaves resistive
MHD for determining the transport.
R.J. TAYLOR: What does your nTET; function look like for all the JET data
versus power input? Does it support Rebut's position regarding the lack of significance of auxiliary heating?
J.G. CORDEY: Until recently, the best nTETj value was obtained with Ohmic
heating (1 = 5 MA); however, using X-point operation at 3 MA, this value has
recently been doubled to 2 x 1020 m~ 3 -s-keV. We do not yet have its scaling withpower.

1
2
3

HORA, H., et al., Phys. Rev. Lett. 53 (1984) 1650.
CICCHITELL, L., et al., Laser Part. Beams 2 4 (1984).
LALOUSIS, P., Phys. Lett., A 99 (1983) 89.

IAEA-CN-47/A-II-4

EFFECT OF p AND j PROFILESON ENERGY CONFINEMENT
IN ECH EXPERIMENTS ON T-10
V.V. ALIKAEV, A.A. BAGDASAROV, E.L. BEREZOVSKIJ,
A.B. BERLIZOV, G.A. BOBROVSKU,
A.A. BORSHCHEGOVSKIJ, N.L. VASIN,
A.N.I. VERTIPOROKH, V.V. VOLKOV, E.P. GORBUNOV,
Yu.A. GORELOV, Yu.V. ESIPCHUK, S.L. EFREMOV,
V.S. ZAVERYAEV, V.I. IL'IN, A.Ya. KISLOV, P.E. KOVROV,
S.Yu. LUK'YANOV, Yu.S. MAKSIMOV, A.A. MEDVEDEV,
G.E. NOTKIN, A.B. PIMENOV, V.I. POZNYAK, LA. POPOV,
S.V. POPOVICHEV, K.A. RAZUMOVA, P.V. SAVRUKHIN,
E.V. SITAR, A.M. SOLNTSEV, M.M. STEPANENKO,
V.S. STRELKOV, A.V. TARAKANOV, K.N. TARASYAN,
K.B. TITISHOV, V.M. TRUKHIN, G.S. YARAMYSHEVI.V. Kurchatov Institute of Atomic Energy,
Moscow
S.N. VLASOV, A.L. GOL'DENBERG, V.A. MARKELOV,
V.A. FLYAGIN, B.V. SHISHKIN
Institute of Applied Physics of the USSR Academy of Sciences,
GorkijUnion of Soviet Socialist Republics
Presented by B.B. Kadomtsev
Abstract
EFFECT OF p AND j PROFILES ON ENERGY CONFINEMENT IN ECH EXPERIMENTS
ON T-10.
ECH experiments with doubled HF power, P HF , were carried out on T-10, and the dependence
of energy confinement on P HF was shown to remain the same as in previous T-10 experiments. The
dependence of the transport coefficients on the pressure gradient was shown to be strongly non-linear.
The results of off-axis heating agree with the idea of pressure (current) profile consistency.

1.

INTRODUCTION

An ECH experiment with increased HF power has recently been carried out on
T-10. The HF set-up consisted of six gyrotrons with f = 81 GHz and provided a
PHF value of up to 2 MW, which was twice as high as previously [1, 2]. Diamagnetic measurements showed the absorbed power, Pab, to be 70 to 75% of PHF111

112

ALIKAEV et al.

The main goal of our recent investigations was to study the influence of the plasma
parameter profiles on the energy confinement. The profiles were varied by shifting
the ECR zone. The conservation of the p and j profiles (i.e. the dependence of thetransport coefficient on the Pab profile [3, 4]), its dependence on vp e and the connection of this dependence with the limit profile idea [3] were studied.
The plasma current Ip was varied from 180 to 230 kA, which corresponded toqL = 6.4 to 5.0 at aL = 34 cm. The ratio Ptot/PoH was increased up to 5.5, P,ot
being defined as Pab + PQ£H2.

PECULIARITIES OF RECENT EXPERIMENTAL CONDITIONSThe experimental conditions differed from those in the previous experiments[1,2], which resulted in a difference in the Ohmic stages. First, the vacuum chamber
was replaced by a new one. Second, two circular carbon limiters were removed
because the large amount of graphite has led to hydrocarbon film production on thefirst wall. This resulted in breakdowns between the wall and the waveguide and in
a cut-off of the power input. To prevent this, we needed a great number of training
discharges, along with the HF pulses.
In the recent experiment, it was only the movable rail limiter that was made ofgraphite. The reduction of the quantity of graphite, together with waveguide baking,eliminated arcing and, on the other hand, changed the near wall conditions significantly. The metal impurity influx increased so that the iron density in the centre rose
up to 1-2 X 1010 cm" 3 , i.e. by a factor of five. This seems to be the reason for theincrease in radiation power; Prad = 0.7 POH at fle = 3 X 1013 cm" 3 in comparison
with 0.25 POH in [1. 2]. The recycling and the additional particle influx during ECH
went down markedly. Figure 1 shows the energy content in the Ohmic plasma,
W 0H , as a function of n,. at Ip = 180 kA, Bt = 2.9 T. At the beginning of theexperiments (solid line), a saturation in W0H(n,,), accompanied by a rise in them = 2 mode activity (Bp/Bp = 4 X 10 " 4 at fig = 3.2 x 1013 cm" 3 ) was observed
near to the limit value of fl,., fleiim = 3.5 x 1013 cm" 3 , above which disruption
occurred. The energy confinement time TE did not exceed 50 ms and was lower thanthat in [2]. In the course of the experiments, the heavy impurity level, the radiation
losses and the loop voltage were going down, apparently because of changes in thewall conditions. The value of ne|im then rose up to 4.2 x 1013 cm" 3 , MHD activity
stopped up to fij = 3.6 x 1013 cm" 3 , and no evident saturation on the W0H(n<.)
graph was observed (Fig. 1, dashed line). TE increased up to 70 ms.
3.

ENERGY CONFINEMENT DURING ON-AXIS ECH

In contradiction to the previous experiments [2], the decreased particle influx
no longer compensated for the outflux rise caused by the drop of TP at ECH. As aresult, ne decreased during ECH without additional gas puffing (Fig. 2(a)). The

IAEA-CN-47/A-II-4

113

n e .10 1 3 cm
FIG. 1. W0H(ne) dependence on Ohmic stage of discharges with Ip = 180 kA, B, = 2.9 T. Solid linerefers to results obtained at beginning of experiments, dashed line to the final stage.

decrease was stronger at higher r^ and PHF. As before [2], the most significant
decrease in ii,. was recorded at the plasma centre. It caused ne profile flattening(Fig. 3(a)).
The energy content reached its maximum Wm at At™ = 25 ms of the ECH
pulse (Fig. 2(a)) and went then down while Te continued growing during the wholepulse all over the cross-section (Fig. 4(a)). The decrease in energy content (fromWm = 37.5 kJ to 26 kJ at At = 65 ms) can be explained, within the experimental
accuracy, by the dependence TE OC %, which is valid for ECH as has been shown in[1] and was confirmed in the recent experiments. If fie is maintained constant by anadditional puffing during the ECH pulse, there is virtually no change in W after
reaching the stationary value (Fig. 2(b)). The dependence of Wm on P,ot is displayedin Fig. 5, together with the data of [2]. The 7E(Pt0t) dependence does not, within theexperimental accuracy, deviate from that estimated earlier [2]:- CD I'D\ -0.55

if the drop in r^ and the r^fi,.) dependence are taken into account. This result does
not contradict the conclusion reached in [2] that the electron energy confinement time
for the column as a whole, TEe(aL), is proportional to <Te> ~'A, which is equivalent to=(P,O,/P O H)"' A

because (/VWOH = 0.7, and the ions were not heated significantly by ECH in thecase considered.

114

ALIKAEV et al.

0.56

FIG. 2. Time behaviour of energy content increase, AW, and ft, during ECH: (a) on-axis heating,B, = 2.9 T, 6gyrotrons, low level gas puffing during ECH; (b) same with high level puffing;(c) off-axis heating, B, = 3.25 T, ECR zone shift, AR = 18 cm, 4 gyrotrons.

In summary, the P HF doubling did not change the scaling for the global
confinement time obtained earlier.

4.

Te PROFILE IN ON-AXIS ECH

The electron energy confinement in the core varies more dramatically with thePtot variations than the global time rE. The profiles of the relative Te increase at
the end of the ECH pulse were measured for different numbers of gyrotrons in

115

IAEA-CN-47/A-II-4

(a)

10

15

20

25

30

r, cm
FIG. 3. ne profiles at different moments of ECH pulse starting at At = 0: (a) same conditions as
Fig. 2(a); (b) same conditions as Fig. 2(c).

plasmas with i^ = (3-3.5) x 1013 cm 3 and are displayed in Fig. 6(a), (b) and (c).
EC radiometer data are used. For Pab < 1 MW, Te(0) is proportional to Ptot
(Fig. 7) and TJTtOH is practically constant along the radius, providing, thus,
evidence of a Te profile change in a similar way. For higher Ptot, Te(0) almost ceases
rising, which indicates a drastic increase of transport in the plasma centre. The Te
profile tends to flatten (Figs 6(a) and 8), which agrees with the conclusion of [2] thatthe heat conductivity coefficient K<. increases less strongly with increasing radius.

116

ALIKAEV et al.

1

i

I

i

-1
0.55

Q45

/

1

ECRH

0.45

0.5

A,A
0.55

t.S

I , S

FIG. 4. Te time behaviour during ECH at different radii: (a) same conditions as Fig. 2(a);(b) same conditions as Fig. 2(c). ISxR is soft X-ray intensity emitted by plasma core.

A*
30

-

20

/>
- ,'OOH
/ /

cij,

OH» /
/ /

10 " / /
1/
1/
'/
>

1

0.5

A.

1

V5

Ptot
• IUI »

F/G. 5. Dependence of maximum energy content, Wm, on total heating power, Pm,
together with the data ofRef. [2], (1): lp = 180 kA, B, = 2.9 T, ht = 2.8 X 1013 cm~3.

(2),

117

IAEA-CN-47/A-O-4

h? 2
O

o

(b)

O

o

o

o

O-

8
8o

o

o
o

(c)
5

-5

10

15

20

R - R o . cm
FIG. 6. Relative profiles of electron temperature, TJTe 0H, as measured by EC radiometer at end of
ECH pulse. B, = 2.9 T (on-axis ECH), ne = (3-3.5) x 1013 cm'3, (a), (b) and (c) are for 6, 4 and2 gyrolrons, respectively.

In Ref. [3], the dependence of K,. on the pressure gradient, v p e , was shown.
The new results confirm this dependence and point out its strongly non-linear character, i.e. the existence of a limit gradient near which the transport coefficients rise up
significantly. This is illustrated by Fig. 9, where the data obtained for the end of the
ECH pulse at different Ptot and n,. values is displayed. These data relate to theplasma centre (r < 10 cm). The S value,

s=

1

118

ALIKAEV et al.

FIG. 7. Dependence of central Te on P,o,for end of on-axis ECH (B, = 2.9 T,», = 3.2 x 1013 cm'3).

FIG. 8. Tt profiles for OH and end of on-axis ECH with 6 (I) and 4 (2) gyrotrons. B,
ht = 3.2 x 1013 cm~3. PHA data are used for Tt0H.

2.9 T,(POH is the Ohmic power at the Ohmic stage, Pei the electron-ion heat transfer) is,in fact, the ratio IQ/Ke OH if one assumes that the Te profile changes only slightly inthe plasma core. In Fig. 9, v pe is replaced by the central pressure, in accordancewith the idea of p profile consistency [3]. The results of the ECH experiment on T-7
are also shown in Fig. 9 [5]. The rise in Te(0) from 1.6 to 4.2 keV at low HF power

119

IAEA-CN-47/A-H-4

CO
2-

10
l3

n.(0)T.|0),1O keVcm-

15
3

FIG. 9. Increase in Ke for plasma core (r < 10 cm) as a Junction of pressure gradient (see the text).
+ - calculations for Ke oc fes; © , © T-7 data [5].

(Ptot = 0.3 MW) obtained in this experiment can be explained by the fact that thepressure gradient was far from the limit gradient in T-7, because of low
fi,. (6 x 1012 cm- 3 ).
Could the observed K^ v pe) dependence be interpreted as a relation between
their local values? Or is it the result of a distortion of the p and j profiles by ECH,
i.e. is it determined by the P^ profile?
We have not yet enough data to decide which of these explanations is thecorrect one.

5.

ON PROFILE CONSISTENCY IN PLASMA (OFF-AXIS ECH)

Consistency of p and j profiles in plasma (if it exists) can only be realized
through variations of the transport coefficients which would depend on the profile ofthe heating power, Pab(r). This means that the outward shift of the ECR zone would
result in a decrease of the transport coefficients in the plasma core after switchingon the ECH pulse, because we apply no additional power there.
To test the hypothesis of profile consistency, the experiment was done with an
outward shifted ECR zone (AR = 1 8 cm). The results are displayed in Figs 2(c),4(b) and 10. As previously [1, 2], the maximum rise in energy content, AWm, ispractically the same compared with on-axis ECH at the same Pab value, and theconfinement degradation observed at the end of the ECH pulse is stronger (Fig. 2(c)).
The data presented in Fig. 10(a) show that the rise in Te at At™ = 20 ms is stronger
behind the ECR zone, in this case, than with on-axis heating and, hence, AWm does
not change with the AR variations, but for At > At,,,, Te falls in the plasma

120

ALIKAEV et al.

exterior, in contrast to the central ECH. In the plasma core, Te rises during theentire ECH pulse (Fig. 4(b)). These results qualitatively indicate the tendency
towards pressure profile conservation, especially if n,.(r) peaking is taken into
account (Fig. 3(b)), which is the other difference between on-axis and off-axis ECH.
This tendency is confirmed by the energy balance at At = 50 ms in the plasma core(r < 9 cm), which is not supplied with additional power (Fig. 10). K,, decreases in
the core during ECH so that~Z—

- Pei - Prad " (3/2) d/dt

x

(PoH ~ Pei ~~ Prad)0H

f9< n e T e > = 2TT n e T e r d r ; K , . O H S 7 X

1

e0H

~
<

d
dr

~~;— T e ECH^

1016 c m - ' - s " 1JoThis estimation is the upper limit of K,, at ECH because the current profile rebuild
which decreases the P 0H value for off-axis ECH is not taken into account.
The soft X-ray signal traces (Fig. 4(b)) demonstrate rise in I SxR at the centre
20 ms after switching off the ECH pulse. This displays a slow decrease in Te(0) anda possible continuous accumulation of impurities ('B-regime' on T-10).
The slow rebuild of the Te profile in the plasma core during the ECH pulse isconnected with the fact that POH is nearly balanced there by the electron energy
losses Pe. From a rough estimate, the characteristic time of profile rebuild is of theorder of seconds. So, to confirm the p and j profile conservation, an experiment isneeded with longer ECH pulse at lower n,., i.e. with a higher value of (P 0 H -P e )It was assumed in [6] that the low K, in the plasma core ( < 1017 cm" 1 -s~') athigh tig/Ip values was the result of a balance between diffusivity and convective
influx (heat pinch). If balance prevails, then at the beginning of the HF pulse(At = 20 ms) one could expect a strong rise in the heat flux because both the diffusive flux (Fig. 10(a)) at r = 6 cm and the heat pinch would be directed inward.
Hence, the K,, value measured as the ratio of the total electron heat flux to v Te
would also rise strongly. However, an energy balance analysis showed that in theplasma core K,. was at least not higher than in the Ohmic stage. We may conclude
that the heat pinch does not play an essential role under these conditions.

6.

INTERNAL MHD ACTIVITY

A specific MHD activity was detected in recent experiments by pickup coilswith good reproducibility (Fig. 11 (a)). This activity had also been observed

IAEA-CN-47/A-II-4

1.5

121

4SO
60

40

05

20

(b)
10

15

20

25

R-R0.cm
FIG. 10. (a) Teprofiles at two instants of ECHpulse together with profile ofTe '(dTJdt) at the startof non-central ECH. B, = 3.25 T, ne = 3.6 x 1013 cm~3, 4 gyrotrons; (b) relative increase ofTe at
end of ECH pulse (At = 50 ms) for same conditions. A refers to Te profile at At = 50 ms of central
heating.

previously. It manifests itself in sinusoidal oscillations on the traces of the EC radiation and of the SXR intensity (Fig. ll(c)) at intermediate radii. The amplitude of Bp
oscillations (see Fig. 11 (a) for Bp/Bp) is so high that these oscillations would resultin a major disruption or in strong confinement degradation if m were equal to 2 or
3 (Fig. ll(b)). However, this 'quiet' mode does not significantly enhance theplasma-wall interaction (Fig. 11 (a)). It results in a drop in Te(0) and suppresses thesawteeth, i.e. it exists mainly in the plasma interior. The mode is observed forn; = (1.5-2.5) X 1013 cm" 3 at L = 200-230 kA (Bt = 3 T).

ALIKAEV et al.

122

(c)

FIG. 11. Time behaviour ofMHD activity characterized by Bp/Bp amplitude (Bp is the poloidal field)
and traces of He, Prml and loop voltage U: (a) for internal MHD mode; (b) for m = 2 mode;(c) time behaviour of Te at different radii of auxiliary-heated discharge (on-axis ECH) with internal
MHD mode.

7.

SUMMARY

1. Doubling the HF power during the ECH experiments on T-10 practicallydid not change the character of the energy confinement dependence on Ptot,
compared to previous experimental results. This dependence does not contradict the'Ohmic' scaling, T&. OC <Te> ~0-5.
2. The dependence of Ke on v pe for the plasma core obtained previously is
confirmed; it is shown to be essentially non-linear. A limit value of v pe exists inthe vicinity of which the transport coefficients rise significantly.
3. The results of the off-axis heating experiments do not contradict the idea ofp and j profile consistency.
4. The low Kg values in the plasma core which are characteristic for highfle/Ip, ohmically heated plasmas are shown to be not the result of the existence of aheat pinch.

ACKNOWLEDGEMENTS
The authors would like to thank B.B. Kadomtsev and V.S. Mukhovatov for
fruitful discussions and the T-10 staff for providing the experiments.

IAEA-CN-47/A-II-4

123

REFERENCES
[1]
[2]
[3]
[4]
[5]
[6]

ALIKAEV, V.V., et al., in Plasma Physics and Controlled Nuclear Fusion Research 1984 (Proc.
10th Int. Conf. London, 1984), Vol. 1, IAEA, Vienna (1985) 419.
T-10 GROUP, in Controlled Fusion and Plasma Physics (Proc. 12th Europ. Conf. Budapest,
1985), Vol. 1, European Physical Society (1985) 42.
ESIPCHUK, Yu.V., RAZUMOVA, K.A., in Controlled Fusion and Plasma Heating (Proc. 13th
Europ. Conf. Schliersee, 1986), invited paper.
ESIPCHUK, Yu.V., KADOMTSEV, B.B., Pis'ma Zh. Ehksp. Teor. Fiz. 43 (1986) 573.
ALIKAEV, V.V., et al., Paper IAEA-CN-47/F-II-4, these Proceedings.
T-10 GROUP, in Controlled Fusion and Plasma Physics (Proc. 12th Europ. Conf. Budapest,
1985), Vol. 1, European Physical Society (1985) 38.

DISCUSSION
R.J. GOLDSTON: This is more a comment than a question. It seems that weare learning how strong the profile conserving mechanism is in the plasma core. It
appears to be able to reduce x e by something like a factor of two, but it does notturn Xe negative, which would correspond to a heat pinch, nor does it even make x e
zero.
B.B. KADOMTSEV: Yes, I quite agree.
M.A. DUBOIS: The absence of a central electron temperature spike during
central heating is in sharp contradiction with the TFTR results. Is it certain that theq = 1 surface is present inside those plasmas?
B.B. KADOMTSEV: No, it is not absolutely certain. We do not yet have any
experimental data on this point.

IAEA-CN-47/A-n-S

ACHIEVEMENT OF REGIMES WITH
HIGH DENSITY, LOW qa AND
GOOD CONFINEMENT ON ASDEX
H. NIEDERMEYER, F. WAGNER, G. BECKER, K. BUCHL, A. EBERHAGEN,
G. FUSSMANN, O. GEHRE, G. VON GIERKE, O. GRUBER, G. JANESCH1TZ,
M. KAUFMANN, K. LACKNER, R. LANG, V. MERTENS, H. MURMANN,
W. SANDMANN, E. SPETH, K. STEINMETZ, G. VLASES1, U.S. BOSCH,
H. BROCKEN, J. GERHARDT, E. GLOCK, G. HAAS, J. HOFMANN,
A. IZVOZCHIKOV2,
F. KARGER,
M. KEILHACKER3,
O. KLUBER,
M. KORNHERR, M. LENOCI, G. LISITANO, F. MAST, H.M. MAYER,
K. McCORMICK,
D. MEISEL,
E.R. MULLER 3 ,Z.A. PIETRZYK1,W. POSCHENRIEDER,
H. RAPP, H. RIEDLER, H. ROHR, J. ROTH,
F. RYTER4,
F. SCHNEIDER,
C. SETZENSACK,
G. SILLER,
P. SMEULDERS3,
D.ZASCHE

F.X. SOLDNER,

K.-H. STEUER,

O. VOLLMER,

Euratom-IPP Association,
Max-Planck-Institut fur Plasmaphysik,
Garching, Federal Republic of Germany

Abstract
ACHIEVEMENT OF REGIMES WITH HIGH DENSITY, LOW qa AND GOOD CONFINEMENT
ON ASDEX.
The density limit in ASDEX decreases for % < 2.7. However, in gettered discharges at highcurrent (500 kA), the density limit shows again a steep increase to a higher second maximum(M = neR/B, = 7.2 x l O ^ m ^ - T " 1 ) at c^ = 2.05. The lowest c^-value achieved so far in Ohmic
and beam heated plasmas is 1.9 (M 0H = 6.6 x 1019). With continuous pellet injection, peaked density
profiles with very high central values (2.2 x 10 20 m" 3 ) were obtained in Ohmic and weakly beamheated discharges. In pellet refuelled discharges, rE-values of 160 ms are achieved with nerE =3.5 x 1019sm~3. The confinement in the L-regime seem to be governed by the same principles as in
the OH-saturation regime. T E can be increased by deuterium operation or by ICRF instead of NI heating. A recipe for achieving the H-mode has been developed which was successfully applied to the casesof Q^ < 2, with ICRH or at high density with pellet refuelling.

1

University of Washington, Seattle, WA, USA (DOE contract).

2

Academy of Sciences, Leningrad, USSR.
Present address: JET Joint Undertaking, Abingdon, Oxfordshire, UK.
CEA, Centre d'&udes nucleaires de Grenoble, Grenoble, France.

3
4

125

126

NIEDERMEYER et al.

1. Density limit of gas refuelled discharges

The density limit of ASDEX divertor discharges has been studied
at q -values (cylindrical approximation neglecting the separatrix) above 2.2 by ramping up the density to the disruptive
limit. Feedback controlled gas puffing has been applied inohmic and beam heated divertor discharges without gettering/1/. An increase of the density limit with additional heating
was achieved both in B - or I D (1/q ) -scans(see Fig. 1 ) . TheIradensity limit, however, increases much less than proportional
to the heating power. The Hugill-Murakami scaling was found tobe valid in ohmic discharges but not with strong additional
heating, where at given q -values M = n R/B at the density
limit scales like 1//B • 3©
Xi
In most cases edge cooling is believed to be responsible forthe density limit. In low-q

discharges (below about 2.7) the

q=2 surface is obviously destabilized directly while at higher
q -values thermal effects cause the current channel to shrink
and as a consequence trigger the disruptive m=2-mode. In thesecases the onset of marfes is always observed before the discharge shrinks. This kind of a two dimensional thermal
instability increases edge radiation substantially and is thus
strongly involved in the shrinking process.

2. Pellet discharges

Pellet injection as a tool for more efficient refuelling was
expected to reduce edge cooling and permit higher densities. In
fact an appreciable increase of the maximum density has beenachieved by this method as well as an unexpected improvement ofparticle and energy confinement. Injection of a series of D«pellets with a centrifuge / 2 / , /3/ (size corresponding to an
1q

averaged density of 1 • 10m_Q

, velocity of 650 m/s, repeti-tion rate less than 40 per second, penetration depth about a/2)

127

IAEA-CN-47/A-n-S

OH gettered^ V

1/qB t =19T

*"»

0.4OH

y

?*

1
T

Pellets
OH caibonized
non-getteredB, =2.2 1

non-gettered0.2-, / N |

aS

11.7 MWI
non-gettered

010
nR/B T

FIG. 1. Maximum achievable densities for different types of divertor discharges plotted in the
Hugill diagram.

may cause a peaking of the density profile as already observedin experiments with smaller numbers of pellets /*»/, / 5 / , /6/,
/7/. Penetration to the center is not necessary to get triangular density profiles. Sufficient recycling, however, is anecessary condition for the onset of this regime. We attributethe change of the profile to an increased ratio of the inward
drift velocity to the particle diffusion constant evidently
caused by pellet injection. Unfortunately an accumulation of
metal impurities in the plasma center results from the improved
particle confinement and now central radiation determines thedensity limit. With wall carbonisation a Murakami parameter of
19
20 — ^
8.5 • 10and a central density above 2 • 10m

can beachieved in ohmic discharges. Before excessive central radiation arises, the T -profiles are nearly unchanged compared togas refuelling. With this strongly increased density high confinement times of up to 160 ms in ohmic discharges are obtained. We surmise that the transport properties are changed asshown by the fact that triangular density profiles with nearly
constant peak density are sustained for 300 ms after having
stopped pellet injection. There are indications that evensmaller pellet penetration depths are sufficient to cause

128

NffiDERMEYER et al.

density peaking. High power c o - i n j e c t i o n prevents peaking of
the p r o f i l e s .

The few discharges with c o u n t e r - i n j e c t i o nper-formed so far show improved p a r t i c l e and energy confinement.
This r e s u l t and a remarkable curvature of the p e l l e t

trajec-t o r i e s as observed on photographs suggest t h a t a l s o during theohmic phase t o r o i d a l plasma r o t a t i o n plays an important r o l e .

3. Low-q

discharges at high densities

With only gas-puff refuelling a Murakami parameter of 7.2 •1Q
10
was reached in gettered divertor discharges at very low
q -values of about 2 (cylindrical approximation) performed withaa plasma current near the design limit of 500 kA (Fig. 1 ) . In
earlier experiments on ASDEX a steep drop of the density limit
with decreasing q has been found when approaching q = 2 inaaagreement with all known results of other tokamaks. At small
q -values a strong saturated m=2-mode develops already at
3.
moderate densities and increases slowly with increasing density(Fig. 2 ) . Finally it leads to a disruption. Now it has been
discovered that at even smaller q -values the m=2-activity may
suddenly drop to a very small level during the slow density
rise (Fig. 2 ) . It reappears again but then at much higher
densities and determines the density limit, q -values of 1.9
3
were obtained, however, the optimum q is about 2.0. The onlyadifference in the external parameter settings of the two cases
compared in Fig.2 is the plasma current. The attempt to raise
the plasma density evidently results in an increase of MHD
fluctuations for the lower plasma current, but in theirdecreas.e for the higher one. Due to the singularity induced by
the stagnation point, the true q= 2 -surface for the twocases will always be inside and close to the separatrix, but atdifferent locations: the qualitatively different response of
the MHD signals might therefore result from the predicted
reaction of tearing modes, which are stabilized by flattening

129

IAEA-CN-47/A-II-5# 19514
•ICQ

0

500-

Us)
FIG. 2. Time evolution of plasma current, mean density and MHD activity during discharges at
qa = 2.15 {if = 420 kA) (top) and 2.05 (IF = 440 kA).

of the current density profile (and hence by cooling) inside
the resonant surface, but destabilized by it on the outside.
Stable discharges with long density plateaus 5 % below the density limit were run at q = 1.9. T profiles are trapezoidalaewith the flat central part extending to a/2 in ohmic discharges
and high power beam heated discharges. Density profiles areparabolic in ohmic discharges, in beam heated discharges they
are rather flat with a very high edge density. Confinement ofohmic discharges is good, -x is comparable to the ip-values ofthe saturation region away from the density limit (80 ms
typically), the toroidal beta values rise with increasing
density and with decreasing q . At q = 1.9, 8. = 0.61 % was
3
3
L

130

NIEDERMEYER et al.

•

0

2

4

6

8

10

0

2
4 6
n e (10 1 3 c m " 3 )

8

0

2

4

Ip=0.38MA
P,c = 1.2 MW| c 1 . 7 MW
2n C K (D*/H*l
2fl C H OIH)

6

8

10

FIG. 3. Survey of confinement time values versus density in (a) ohmically, (b) NI and (c) ICRH heatedplasmas, (a) BT = 2.2 T; gas puff: IP = 420 kA, pellet: IP = 380 kA; (b) BT = 2.2 T;(c)BT = 2.4T.

reached compared to 0.43 % at q

= 2.5. With beam injection atQ.

a power level of 1.7 MW, L-mode confinement is being interrupted by short phases of the H-mode. At 2.6 MW the signatureof the H-mode is more clearly pronounced. Measured confinement
times of about 50 ms are only slightly higher than those
expected from L-mode scaling. This might be attributed tostrong impurity radiation present with high power beam heating
which was not observed in ohmic discharges.
4. Confinement

On Asdex the most important confinement regimes which seem tobe characterized by intrinsically different properties can berealized. Figure 3 summarizes density scans of the global
confinement time !„ for ohmic and auxiliary heated discharges.

IAEA-CN-47/A-II-5

131

The ohmic data represent pellet refuelled plasmas and scans inpure hydrogen or deuterium discharges with gas-puff refuelling.
The auxiliary heated data comprise density scans of discharges
in the L-regime heated either by NI or ICRH, and plasmas in the
H-mode for the injection scenarios H

-»• D

and D

•* D , andfinally, the result achieved in the quiescent H-mode when ELMs
are suppressed /8/.
The degradation of confinement with ICRH is less severe thanthat with NI. The reason is not clear but could be due to thebetter particle confinement with ICRH or the lack of toroidal
rotation. Deuterium as working gas is superior to hydrogen bothwith ohmic and auxiliary heating. This isotope dependence improves the confinement of 2Q- ICRH applied to basically

deuterium plasmas (with the minimum necessary hydrogen concentration of -20 %) or D(H) - minority heating at the fundamental
frequency. Also the H-mode confinement is subject to the isotope effect.
Pellet refuelling yields ir-values up to 160 ms /9/ and n T.,19,#evalues up to 3.5 • 10sm"->. In the quiescent H-mode (H ) ,*T E = 120 ms is found. (As the H -mode strongly suffers from
impurity accumulation the core-radiation overlapping with thepower deposition profile has been corrected for the -r^-analysis.
M. 1. L-Regime

The ohmic saturation and the L-regime are both regimes withdegraded confinement. Common features are the lack of density
dependence of T_ and a similar improvement with isotope mass.
Another aspect common to OH-saturation and L-regime is "profile
consistency" /10/: An increase of I (decrease of q ) broadens
P
3
the T

profile in the same way. Figure *) shows the volume aver-aged electron temperature <T > versus plasma current both for
NI-heating at different heating power levels and for OH-heating
for two different densities. Irrespective of the heating method

132

NIEDERMEYER et al.

(keV)

• a

1.6 MW• A

0.4 MW

> O

OH

1.0-

0.5-

0.2

0.4

0.3

rp (MA)
FIG. 4. Volume averaged electron temperature Te versus plasma current for different NI power levels
and plasma densities. Solid symbols: fte = 2 x 1013 cm'3; open symbols: he = 3 x 10l3cm~3.

(OH or N I ) , <T > increases with I

in a similar way. These

results indicate the same profile broadening and improvement inenergy content for OH- and Nl-heated plasmas. The broadening of
the Te-profile with increasing current leads ultimately to the
proportinality of iv and current under auxiliary heating
conditions. In the ohmic case, however, the i- of the saturation regime is independent of current despite the fact that theenergy content also increases with current. The difference
between ohmic and auxiliary heating is that in the ohmic casethe heating power increases linearly with current while it isindependent of current with auxiliary heating. Already underohmic conditions, the heating power degrades confinement as it
is known from auxiliary heated plasmas. Both plasma current andheating power are two important parameters which determine theconfinement both in the OH-saturation regime and the L-mode
with the L-mode being the continuation of the OH-saturation
regime towards higher power.
Figure 5 shows the plasma energy content versus power for anohmic current scan and for two Nl-power scans at two current

133

IAEA-CN-47/A-H-5

1.6

FIG. 5. Variation of energy content W with heating power P in Ohmic current scan and with NI at two
plasma current values (the first data point of the NI scans represents the OH starting value). Curves
represent non-linear relation between W and P at constant current; straight line is a fit through Ohmic
data points.

values. The solid lines tentatively indicate an inherent nonlinear relation between W and P. Under OH conditions, thisdependence is hidden, because the heating power and the plasma
current rise simultaneously /11/. From this analysis it isobvious that there has to be a discontinuity of the slopes in a
W versus P diagram at the transition from ohmic to auxiliary
heating as shown in Fig. 5. The reason for the discontinuity isthat an ohmic power scan occurs in a plane where I /P isconstant; with auxiliary heating, however, a power scan iscontinued in an I =const. plane. The ohmic W vs. P-trace is nota scalable relation. The assumption of a special quality of 0Hheating does not seem to be justified. The reason for thecurrent scaling which seems to be unique to all auxiliary
heated plasmas at higher densities is not the introduction of a
new confinement element. Rather it is the disappearance of acommon element which is the dominance of the ohmic heating
power and its proportionality to plasma current /10/. Therefore
also in the H-mode T _ is proportional to I p despite distinct

134

NIEDERMEYER et al.

differences in the development of L- and H-phases and in theabsolute values of T_. The operational space with good
confinement below the ohmic power is only accessible underhighly efficient current drive conditions like with LowerHybrid heating or in non-thermal discharges. Indeed, suchcurrent drive scenarios indicate improved confinement.

4.2. H-Regime
Inspite of the large differences in the profile shapes there
are common characteristics of the good pellet discharges andthe H-mode, for example the reduction of the divertor
recycling, an improvement in runaway electron confinement, thetendency for impurity accumulation, and strong indications of
the importance of the plasma edge conditions for confinement/12/. The H-mode is caused by the sudden development of an edgetransport barrier /12/ which seems to require an edge electron
temperature above 200-300 eV. The deep pellet refuelling obviously prevents the detrimental effect of gas-puff refuelling on
plasma edge conditions which leads to the T,, roll-over and thedevelopment of the saturation regime.
Apart from the insight in the role of the plasma periphery for
the H-transition a recipe for reaching the H-mode emerged fromour studies. The crucial element seems to be the increase of
T (edge) beyond a threshold requiring a minimal heating power(1.8 MW with NI in the double-null configuration) and giving
rise to the preceding L-phase (until T (edge) is sufficiently
heated up) /12/. The edge temperature threshold also causes thesensitivity of the H-transition on recycling and impurity
radiation cooling. The intrinsically better confinement of deuterium and helium facilitates the H-mode development: In purehydrogen discharges (H-> H ) the power threshold is ~3 MW.

Owing to the ion-electron coupling, the H-transition is eased
when the ion temperature is also high at the edge. Because of

IAEA-CN-47/A-II-5

135

these requirements, recycling has to be kept low and gas
refuelling into the divertor instead of the main plasma chamber
can reduce the power threshold. With low recycling, neoclassical ion transport may be dominant even at the plasma edge. In
this case, a poloidal asymmetry of the energy sink distribution(as realized by the X-point) can reduce the ion heat transport
and with it the power requirements for the H-transition /12/.
In the poloidally asymmetric situation of the single null (SN)
configuration, the power threshold is reduced to 1.2 MW if theion grad-B-drift is toward the X-point. In the opposite case,the power threshold rises to ~3 MW /12/. With the X-point in
the mid-plane, no improvement is expected. The above describedrecipe has been applied to successfully increase the operational space of the H-mode. (1) In the low q -plasmas, described in Sec. 3. when additionally.edge radiation by oxygenis reduced owing to Ti-gettering, the H-mode can be realized in
the most attractive corner of the Hugill-diagram (high current
and high density); (2) reduced recycling in pellet refuelleddischarges allowed an increase of the upper density limit of
19-3
the H-mode from n = 7-8 - 1 0 m
with gas puff refuelling to
20 -3
n > 1.1 • 10 m . (3) The application of ICRH in the
deuterium dominated D(H)-minority heating scheme (instead of
2flCH~heating in hydrogen) yielded the H-mode with a different
heating method. In the SN-configuration, 1.7 MW ICRH-power isrequired. Another improvement in operational space is counterNI which decreased the power threshold to 0.8 MW. The H-phase,
however, is heavily affected by impurity accumulation andtherefore transient.

REFERENCES[1][2]

NIEDERMEYER, H., et al., in Controlled Fusion and Plasma Physics (Proc. 12th Europ. Conf.
Budapest, 1985), Part 1, European Physical Society (1985) 159.
AMENDA, W., LANG, R.S., in Fusion Technology (Proc. 13th Symp. Varese, 1984), CEC,
Brussels (1984) 243.

13 6

NIEDERMEYER et al.

[3] AMENDA, W., LANG, R.S., to be published in J. Phys., E (London). Sci. Instrum. 19 (1986).
[4] GREENWALD, M., et al., in Controlled Fusion and Plasma Physics (Proc. 11th Europ. Conf.
Aachen, 1983), Part 1, European Physical Society (1983) 7.
[5] SCHMIDT, G.L., et al., in Controlled Fusion and Plasma Physics (Proc. 12th Europ. Conf.
Budapest, 1985), Part 2, European Physical Society (1985) 674.
[6] SENGOKU, S., et al., Nucl. Fusion 25 (1985) 1475.
[7] FONCK, R.J., et al., J. Nucl. Mater. 128 & 129 (1984) 330.
[8] GRUBER, O., JILGE, W., BERNHARDI, H., EBERHAGEN, A., et al., in Controlled Fusion
and Plasma Phys. (Proc. 12th Europ. Conf. Budapest, 1985), Vol. 1, European Physical Society(1985) 18.
[9] KAUFMANN, M., Plasma Phys. Controll. Fusion 28 (1986) 1341.
[10] WAGNER, F., GRUBER, O., LACKNER, K., MURMANN, H.D., SPETH, E., et al., Phys.
Rev. Lett. 56 (1986) 2187.
[11] WAGNER, F., et al., to be published in Nucl. Fusion.
[12] WAGNER, F., et al., in Basic Physical Processes of Toroidal Fusion Plasmas (Proc. Workshop
Varenna, 1985).
[13] HINTON, F., Nucl. Fusion 25 (1985) 1457; WAGNER, F., BARTIROMO, R., BECKER, C ,BOSCH, H.S., et al., Nucl. Fusion 25 (1985) 1490.

DISCUSSION
R.J. HAWRYLUK: How was qa defined?
F. WAGNER: We used the cylindrical definition, so that 1/e represents theplasma current.
S. YAMAMOTO: I should like to mention that in JAERI's JFT-2M, H-mode
behaviour was observed in the limiter discharge with simultaneous co-NBI and 2wci
ICRF. Recently, the limiter H-mode discharge was obtained with only co-NBI. In thelimiter H-mode discharge, the plasma is restricted by the inside limiter only, andthere should be some clearance between the plasma surface and the vacuum chamberon the low field side.
R.J. GOLDSTON: You showed that the power threshold for the H-mode transition was 1.2 MW for the grad-B drift in the direction of a single null divertor and
3 MW for the reverse configuration. Could you comment on the plasma parameters— radiated power, impurities, edge electron temperature — at, say, about 1 MW of
input power in each case.
F. WAGNER: The majority of the experiments were made at higher power(1.6 MW). No striking difference in these parameters has been observed in
SN-discharges with different ion grad-B drift directions. But there remains the observation, even in these configurational studies, that the edge electron temperature tends
to increase in the correct configuration (in grad-B drift to the X-point), causing
H-transition when it exceeds a threshold of 200-300 eV.
J. JACQUINOT: Can you comment on the stationary aspect of the H-mode?
I am thinking of a comparison between the duration of good confinement and theduration of the heating pulse.

IAEA-CN-47/A-H-5

137

F. WAGNER: At reduced power, when 0/0^ < 0.7-0.8, the H-mode
remains in a quasi-steady state (parameters only modulated edge located discharges).
Stationary H-mode phases have been maintained by stacking the beams for ~ 10T E .
Approaching the beta limit, the confinement in the H-mode degrades after a maximum. This might be due to a reduction of the beta limit when the current profile
broadens and q(0) rises on account of the high edge electron temperature.
M.C. ZARNSTORFF: Could you comment on any dependence of the H-mode
onset upon the angle of beam injection, i.e. co-injection versus counter-injection?
F. WAGNER: The lowest power threshold for the H-transition occurs withcounter-injection ( ~ 2 x P 0H ). Because of the tremendous impurity accumulation
under these circumstances, the H-phases are short ( - 5 0 ms). The high-Z impurity
problems can be avoided by wall carbonization. The required power for H-mode
onset is, however, higher in this case.

IAEA-CN-47/A-ni-l

STUDIES OF THE REGIME OF
IMPROVED PARTICLE AND
ENERGY CONFINEMENT FOLLOWINGPELLET INJECTION INTO ALCATOR C
M. GREENWALD, M. BESEN, F. CAMACHO, C. FIORE,
M. FOORD, R. GANDY1, C. GOMEZ2, R. GRANETZ,
D. GWINN, S. KNOWLTON, B. LaBOMBARD3,
B. LIPSCHULTZ, H. MANNING4, E. MARMAR,
S.C. McCOOL5, J. PARKER6, R. PARKER, R. PETRASSO,
P. PRIBYL3, J. RICE, D. SIGMAR, Y. TAKASE, J. TERRY,
R. WATTERSON4, S. WOLFE
Plasma Fusion Center,
Massachusetts Institute of Technology,
Cambridge, Massachusetts,
United States of America
Abstract
STUDIES OF THE REGIME OF IMPROVED PARTICLE AND ENERGY CONFINEMENTFOLLOWING PELLET INJECTION INTO ALCATOR C.
Pellet fueling in Alcator C can result in a plasma regime in which both energy and particle confinement are dramatically improved. In this regime, ion thermal losses (highly anomalous in gas fueled
plasmas) are reduced to approximately neoclassical values, resulting in increased global energy confinement for high density discharges where ion losses are important. There is little or no change in electron
thermal conductivity. The pellet fueled discharges exhibit peaked density profiles, which tends to confirm the importance of the ij( parameter in determining ion thermal diffusivity. At the same time,
impurity particle transport is reduced to neoclassical levels, leading to accumulation of impurities andstrong peaking of impurity profiles. Confinement of the hydrogenic species in the discharge core is also
increased. The modifications of ion thermal transport, central particle transport, and impurity transport
are well correlated; there are no discharges where only one or two of these parameters change. In thefavourable transport regime, sawteeth are partially or completely suppressed. This is probably due tothe strong peaking of low Z impurities which flatten the central current profile by increasing the plasma
resistivity. Radiation does not play an important role in the power balance of these discharges. For
plasmas in which the sawteeth are suppressed, the improved confinement is sustained for the remainderof the discharge.

1

Auburn University, AL, USA.
Los Alamos National Laboratory, NM, USA.
3
University of California, Los Angeles, CA, USA.
4
Science Research Laboratory, Somerville, MA, USA.
5
Texas University, Austin, TX, USA.
6
Lincoln Laboratory, Lexington, MA, USA.
2

139

140

GREENWALD et al.

1. Introduction
Previous work on Alcator C demonstrated that pellet fueling could
increase global energy confinement in the "saturated" regime of ohmic discharges [1]. Detailed analysis showed that pellet fueling reduces ion thermal transport to levels near the predictions of neo-classical theory [2,3](FIG. 1). It is important to note that for gas fueling in Alcator C, theion transport is always anomalous. The confinement saturation becomesapparent at high densities where significant power is lost through the ionchannel. The effect of pellets is to reduce ion losses at all densities whichcauses the observed increase in global confinement at high density. These
results are consistent with theories of ion temperature gradient driven
turbulence [4], the so-called T)I or ion-mixing modes [5,6]. Our latest experiments have concentrated on the behavior of particle transport withpellet fueling. A coherent picture has emerged, in which particle and energy confinement are inextricably linked and which may shed light on themechanisms of anomalous transport for both electrons and ions. Recent
results from ASDEX and TFTR have shown some of these same featuresand comparisons may yield additional insights [7,8].
2. Particle transport experiments
There were a number of reasons to expect little change in particle
transport with pellet fueling. Conventional wisdom has linked particle

Ion Transport Anomaly

0.25

0.3

0.35

0.4

0.45

Time (sec)
FIG. 1. Anomaly factor for neo-classical ion transport versus time. A single pellet is injected at 0.3 sec.

141

IAEA-CN-47/A-III-l+v

Molybdenum (20 - 2 9 ) Brightness
16
14

gas fueled
pellet fueled'"c 10

0.1

0.2

0.3

0.4

0.6

Time (sec)
FIG. 2. Brightness of lines from highly ionized molybdenum versus time. Pellet injection at 0.3 sec.

Limiter Radius
FIG. 3. Profiles of soft X-ray emission before and after injection.

diffusivity to anomalous electron thermal conductivity in impurity freeplasmas, (D p oc Xe)> which is unchanged in our experiments. Empiricalstudies of impurity confinement in impure plasmas have shown scalings
quite different from those for electron energy transport [9]:ocq Z bg

142

GREENWALD et al.

Normalized Density Profiles

0.4
0.2

2

3

4

minor radius (cm)FIG. 4. Carbon and molybdenum profiles inferred from X-ray data and a comparison with
neo-classical theory.

Decay of Injected V +20.3'cpellet fueled
gas fueled -

0.8

3

g 0.6

I

01 0.4
W
i

0.36

0.38

0.4

0.42

0.44

0.46

0.48

0.5

0.52

Time (sec)
FIG. 5.

Comparison of decay limes for injected vanadium before and after pellet fueling.

where mb g and Zbg are, respectively, the atomic mass and charge of the
background plasma. None of the parameters in this expression is significantly altered by pellet fueling. In contrast, the experiments demonstrate that particle confinement is substantially increased by pellet fueling.
Transport coefficients for hydrogen cannot be unambiguously obtained forthe gas fueled discharges, but analysis of density profiles after injection

IAEA-CN-47/A-m-l

143

FIG. 6. Enhancement of impurity confinement with pellet fueled density increase.

shows an increase in the ratio of inward convection to diffusion and indicates an increase in central particle confinement of about a factor of 2.
Densities of intrinsic impurities, mainly carbon and molybdenum,begin to
increase substantially in the plasma core after pellet injection. This canbe seen in FIG. 2, which shows the time history of lines of highly ionizedmolybdenum. The source, as indicated by the intensity of lines from low
ionization states of these impurities, shows little change.
Further evidence of the improved impurity confinement is seen in datafrom arrays of soft x-ray diodes. Radiation in this spectral range is sensitive to plasma density, temperature, and impurity concentration. Afteran initial drop due to plasma cooling by the pellet, the signals increase
and peak strongly as the impurities build up on axis (FIG. 3). The impurity peaking is even greater than would be apparent at first glance, since
the x-ray profile is dominated by the electron temperature profile before
the pellet and by the impurities afterwards. Careful analysis of data fromarrays with different spectral filtering has determined the time histories of
profiles of carbon and molybdenum [10] and shown them to be in reasonably good agreement with the theory of neo-classical impurity transport[11] FIG. 4. Experimental values for the diffusivity D and convective
velocity V of the carbon were 300 cm 2 /s and 1000 cm/s, respectively.
Comparisons of plasmas before and after injection indicate that the principal change is a drop in D by about an order of magnitude.
These observations are confirmed by measuring the confinement of
injected impurities [12]. In these experiments, trace quantities of nonintrinsic, non-recycling impurities are injected by the laser blow-off method

144

GREENWALD et al.
1000

10
0.01

At (msec)
FIG. 7. Correlation ofTt, the impurity confinement time, and rTe, the relaxation time for the central
electron temperature.

and their time histories followed by UV and X-ray spectroscopy [9]. Figure
5 shows results of such an experiment which compares the confinement of
injected vanadium in pellet and gas fueled plasmas. The vanadium confinement time before injection is 16 msec,which is in agreement with theempirical scaling law. After injection, the time increases to 125 msec.
(Actually, the empirical law would predict a slight decrease since in thiscase the deuterium plasma was diluted with hydrogen from the pellet.)
Typically, the increases are factors of 5 - 20. When the increase in injected impurity confinement is plotted against the normalized density increase (An/n) a threshold is seen at about An/n = 1 (FIG. 6). This same
threshold is also observed for the accumulation of intrinsic impurities, thereduction of ion thermal transport and for peaking of the electron density
profile. An/n is not a unique parameter for characterizing the threshold.
Because of the relationship between pellet size and ablation rate [13], thethreshold could also be expressed in terms of the pellet penetration depth.
This same threshold is seen in an apparently unrelated measurement,
the rate of fall of the central electron temperature in the first few hundred^seconds after injection. The pellets in these experiments were injected
at about 8.0 x 104cm/sec and were completely evaporated after travelling less than 8 - 1 2 cm, thus they never reached the magnetic axis (oreven the sawtooth mixing radius). The expected result would have beena cold annulus of plasma surrounding an unperturbed hot core. With ordinary (albeit anomalous) heat transport, Xe ~ 1 ~ 2 x 10 3 cm 2 /s, theIAEA-CN-47/A-III-l145

relaxation time would be a few msecs. Nonetheless, the central temperature, as measured by electron cyclotron emission and Thomson scattering,typically fell much more rapidly, maintaining its shape and reaching equilibrium in as little as 100 ^tsec [14,15]. The surprising result is that therelaxation time shows the same threshold and is strictly correlated withthe confinement improvements described above (FIG. 7). This observation suggests that the bifurcation in plasma regimes occurs immediately
after injection. The transport analysis is consistent with models wheretransport coefficients change promptly at injection.
3. Sawtooth effects
A common feature of pellet fueled discharges is the modification of
the sawtooth oscillations [16]. Typically the sawteeth continue through theinjection, initially with little change in period or amplitude. Gradually theperiod increases, culminating in either one or more very large sawteeth orin their complete suppression. The variation in behavior is correlated withthe peaking of the electron and impurity densities and thus with particle
confinement; plasmas with the most peaked profiles have no sawteeth.
Sorting out cause and effect is difficult. It appears that the sawteeth do
not have a direct influence on the regime of particle transport but rather
the reverse. As impurities peak on axis the central resistivity increases.
The effect on the plasma current profile can be seen by considering theequation for the evolution of the poloidal field

4TT [ dr \r<r dr

where a is the Spitzer conductivity ~ T 3 / 2 /Z e fr; substituting, one obtains
drBp _ j?_ F d_ (ldrBA
dt~ 4TC<T I &r\r

dr

_ / 3 1_&T _ 1 dZ\ drBp
)

\2T

dr

Z dr)

dr

The first term is diffusive and tends to flatten the current profile.
The second and third terms are convective and lead to growth or damping depending on their relative sizes. Ordinarily, the term proportional
to temperature gradient will dominate and give rise to the thermal instability which drives the sawteeth. However, after injection,with impurity
densities becoming strongly peaked, the third term can dominate, flatten the current profile, raise q(0), and turn off the sawteeth. A series ofsimulations were run to solve for the simultaneous evolution of an impurity species (carbon) and current density. In these simulations, Spitzer

146

GREENWALD et al.
Safety Factor — q

Carbon Density

0.05

0.1

Time (msec)

0.15

0.05

0.1

Time (msec)

FIG. 8. Contours of carbon density and plasma qfrom a simulation of impurity and current diffusion.

resistivity was used for the current diffusion and the particle transportcoefficients were adjusted to match experimental data. The results can beseen in FIG. 8 where contours of carbon density and plasma safety factor
q are plotted against space and time. After a single pellet, injected at
30 msec, the central value of q goes from an initial value of .95 to above
1.0 in 25 msec. This is in agreement with our observations of sawtooth
suppression. It should be noted that in the plasmas considered here, radiation has negligible effect on central power balance. We also note that
sawteeth can have an indirect effect on the transport regime. If the sawteeth are not completely eliminated, they flatten the density profile andthe plasma reverts to the pre-injection confinement mode.

4. Density limit studies
Pellet injection has also provided a tool to study the physics of thedensity limit. For gas fueling, a limit n« ocl p has been seen on most tokamaks [17]. The standard explanation for this limit is that high plasma
densities lead to a radiative collapse of the plasma column to an unstable
MHD equilibrium [18]. This explanation does not seem consistent withthe relative insensitivity of the limit to auxiliary power or Zeff. (Thelimit considered here must be distinguished from the Murakami limit withn,. ~ B/R which is sensitive to both these factors [19].) Analysis of thedensity decay after pellet fueling showed that the central particle confinement degraded markedly as the density limit was approached. With
pellets, the limit could be surpassed (briefly) without disruption,but theIAEA-CN-47/A-HI-l

147

FIG. 9. Density decay time after pellet injection plotted versus Jin, the average current density(MA/m2) over average density (1014 cm~3).

plasma shed the excess particles extremely rapidly. Figure 9 shows the
decay time vs J/n e , where J is the plasma current divided by the plasma
cross-sectional area. This suggests that this density limit is due to thedegradation of particle confinement, and the disruptive limit that is commonly observed with gas puffing occurs when the plasma is "choked" withgas that it cannot absorb [20]. The same scaling may explain the onsetof Marfes at densities somewhat below the limit. As particle confinement
degrades with respect to energy confinement, the average energy of anedge particle drops and triggers a thermal/radiative collapse.

5. Discussion and Summary
Several features of these experiments have implications for theoriesof plasma transport. The observation that ion thermal conductivity decreases as the density gradient is increased (with constant temperature
profiles) is consistent with theories of the T)\ mode. However, electron
thermal transport, if driven by drift wave turbulence [21,22], should increase under these conditions and this is not observed. Further, ion energy
and particle transport appear to be tightly coupled and both uncoupledfrom electron energy confinement. Electron, ion and impurity particle
fluxes are always tied through the ambipolarity constraint, with net particle flux limited by the less diffusive species. It seems reasonable thatthe electrons, which remain highly anomalous, are "held back" by theions and that particle confinement is governed by the same mechanisms
which determine ion energy transport. (The r)-s modes are predicted to148

GREENWALD et al.

drive particle diffusion in collisionless plasmas [23] and an inward pinch inthe collisional regime [5].) The decoupling of particle and electron thermaltransport might suggest that electron transport is not driven by purely
electrostatic modes which would give rise to intrinsically ambipolar particle transport through the E X B drift. With this interpretation, theexperimental observation can be taken as support for electron transport
driven by electro-magnetic mechanisms [24].
In summary, with pellet fueling we have attained a regime in which theion species show nearly neo-classical transport. The improvement seems
to be linked to steep density profiles and may be related to suppressionof 7j; modes. This could have important implications for future machines
as these modes are predicted to seriously degrade ion energy confinement
as the reactor regime is approached. On the other hand, suppression ofthis turbulence and the ensuing neo-classical accumulation of impurities
presents problems for long pulse ignited plasmas.

ACKNOWLEDGEMENT
This work was supported by the United States Department of Energy, underContract No. DE-AC02-78ET51013.

REFERENCES[1] GREENWALD, M., et al., Phys. Rev. Lett. S3 4 (1984) 352.
[2] WOLFE, S., et al., Nucl. Fusion 26 (1986) 329.
[3] CHANG, C.S., HINTON, F.L., Phys. Fluids 25 9 (1982) 1493.
[4] LEE, G.S., DIAMOND, P.H., Rep. IFSR-209, Institute for Fusion Studies, Texas Univ., Austin(1986); Phys. Fluids 29 10 (1986) 3291.
[5] COPPI, B., SPIGHT, C , Phys. Rev. Lett. 41 8 (1978) 551.
[6] ANTONSEN, T. ( COPPI, B., ENGLADE, R., Nucl. Fusion 19 (1979) 641.
[7] KAUFMANN, M., in Controlled Fusion and Plasma Heating (Proc. 13th Europ. Conf.
Schliersee, 1986), Vol. 28, No. 9A, European Physical Society (1986) 1341.
[8] SCHMIDT, G.L., MILORA, S.L., et al., in Controlled Fusion and Plasma Physics (Proc. 12th
Europ. Conf. Budapest, 1985), Vol. 2, European Physical Society (1985) 674.
[9] MARMAR, E., et al., Nucl. Fusion 22 (1982) 1567.
[10] PETRASSO, R., et al., Phys. Rev. Lett. 57 6 (1986) 707.
[11] HIRSHMAN, S.P., SIGMAR, D.J., Nucl. Fusion 21 (1981) 1079.
[12] TERRY, J., et al., Particle Transport Following Pellet Injection into the Alcator C Tokamak,
Rep. PFC/JA-86-23, Plasma Fusion Center, Massachusetts Inst. of Tech., Cambridge (1986).
[13] MILORA, S.L., FOSTER, C.A., et al., Nucl. Fusion 20 (1980) 1491.
[14] GREENWALD, M., et al., in Controlled Fusion and Plasma Physics (Proc. 11th Europ. Conf.
Aachen, 1983), Vol. 1, European Physical Society (1983) 7.

IAEA-CN-47/A-IH-l
[15]

[16]
[17]
[18]
[19]
[20]
[21]
[22]
[23]
[24]

149

GOMEZ, C., Study of Electron Temperature Evolution during Sawtoothing and Pellet Injection
Using Thermal Electron Cyclotron Emission in the Alcator C Tokamak, Rep. PFC/RR-86-8,
Plasma Fusion Center, Massachusetts Inst. of Tech., Cambridge (1986).
GREENWALD, M., et al., in Plasma Physics and Controlled Nuclear Fusion Research 1984
(Proc. 10th Int. Conf. London, 1984), Vol. 1, IAEA, Vienna (1985) 45.
AXON, K.B., et al., in Plasma Physics and Controlled Nuclear Fusion Research 1980 (Proc. 8thInt. Conf. Brussels, 1980), Vol. 1, IAEA, Vienna (1981) 413.
WESSON, J., et al., in Controlled Fusion and Plasma Physics (Proc. 12th Europ. Conf.
Budapest, 1985), Vol. 1, European Physical Society (1985) 147.
MURAKAMI, M., CALLEN, J.D., BERRY, L.A., Nucl. Fusion 16 (1976) 347.
GREENWALD, M., et al., A New Look at Density Limits in Tokamaks, Rep. PFC/JA-86-22,
Plasma Fusion Center, Massachusetts Inst. of Tech., Cambridge (1986).
PERKINS, F., in Heating in Toroidal Plasmas (Proc. 4th Int. Symp. Rome, 1984), Vol. 2, ISPP
and ENEA (1984) 977.
TANG, W., et al., in Plasma Physics and Controlled Nuclear Fusion Research 1984 (Proc. 10th
Int. Conf. London, 1984), Vol. 2, IAEA, Vienna (1985) 213.
DIAMOND, P.H., presented at Int. Workshop on Small Scale Turbulence in Magnetized
Plasmas, Cargese, July 1986.
CALLEN, J.D., Phys. Rev. Lett. 39 24 (1977) 1540.

DISCUSSION
F. WAGNER: rE in the Ohmic saturation regime depends on the isotopic massof the target plasma. It had been shown in the ALCATOR and ASDEX devices that
TE is smaller for hydrogen than for deuterium. This scaling is different from thescaling of neoclassical ion transport and I wonder whether the ion transport in thesaturation regime should be described by an enhancement factor \^cM. GREENWALD: I agree. I don't think ion transport at levels that are three
to six times those of neoclassical transport is really neoclassical. It is simply a way
to benchmark the discrepancy between experiment and theory. In our experiments,
the anomalous ion confinement does not have neoclassical scaling either concerning
the plasma current or the isotope chosen.
M. KAUFMANN: I would like to comment on your statement that the
improved energy confinement in pellet fuelled discharges might be due to -qc. On
ASDEX we also observed that confinement was enhanced while the temperature profile remained unchanged. In cases with relatively small pellets and moderate penetration we found a remarkable improvement in the particle confinement as well as in
the energy confinement before ^ changed.

IAEA-CN-47/A-m-2

CONFINEMENT STUDIES OFADDITIONALLY HEATED PLASMAIN THE JFT-2M TOKAMAK
K. ODAJIMA, A. FUNAHASHI, K. HOSHINO, S. KASAI,
T. KAWAKAMI, H. KAWASHIMA, T. MATOBA,
T. MATSUDA, H. MATSUMOTO, Y. MIURA, M. MORI,
H. OGAWA, T. OGAWA, H. OHTSUKA, S. SENGOKU,
T. SHOJI, N. SUZUKI, H. TAMAI, Y. UESUGI,
T. YAMAMOTO, T. YAMAUCHI, M. HASEGAWA1,
S. TAKADA2, K. HASEGAWA, A. HONDA, I. ISHIBORI,
T. KASHIMURA, Y. KASHIWA, M. KAZAWA,
K. KIKUCHI, Y. MATSUZAKI, K. OHUCHI, H. OKANO,
T. SHIBATA, T. SHIBUYA, T. SHIINA, K. SUZUKI,
T. TANI, K. YOKOYAMAJapan Atomic Energy Research Institute,
Tokai-mura, Naka-gun, Ibaraki-ken, Japan
Abstract
CONFINEMENT STUDIES OF ADDITIONALLY HEATED PLASMA IN THE JFT-2M
TOKAMAK.
The H-mode can be achieved in NBI and/or ICRF heated open divertor discharges, and an
H-transition with improved energy confinement is also observed in limiter discharges with both heating
methods. The gross energy confinement time T EC in the H-mode recovers that of the Ohmic plasma.
A large improvement in gross energy confinement is also observed after injection of a single pellet into
an NBI heated plasma, where during the first 40 ms after pellet injection 70% of the NBI power is stored
in the plasma. T EG reaches 70 ms, which is about twice the Ohmic value. In an additionally heatedplasma, the incremental stored energy behaves as if it had its own energy confinement time Tad, whichis independent of the heating power. In some cases of H-modes, rad is equal to the doubled L-mode
value.

1.

INTRODUCTION

Recent additional heating experiments in JFT-2M have been devoted to theimprovement of energy confinement and to the understanding of its properties during
additional heating. This paper reports the achievement of the H-mode in both diverted
and limiter bounded discharges, with NBI and/or ICRF heating, and of improved
confinement with pellet injection. The confinement properties of the H- and L-modes
are discussed on the basis of the concept of stored energy.
12Mitsubishi Electric Corp., Tokyo, Japan.
Mitsubishi Electric Computer System Corp., Tokyo, Japan

151

152
2.

ODAJIMA et al.
EXPERIMENTAL ARRANGEMENT

JFT-2M is a tokamak with a D-shaped vacuum vessel of major radius 1.31 m,
inside which both D-shaped limiter and open single or double null divertor configurations have been realized [1]. The minor radius a is 0.35 m with full size D-shaped
limiter and 0.26 m for the smallest diverted discharges. The divertor plates and boththe fixed poloidal and the movable rail limiters are made of graphite. The working
gas is deuterium in the H° NBI heating experiment or mixed gas (H/D = 0.4) in the
ICRF heating experiments. NBI heating experiments using hydrogen gas as aworking gas exhibit the same phenomena as are described in this paper, except fora slightly poorer confinement and a slightly higher threshold power for the transition.
The ICRF heating power PICH = PG(Rp—Rv)/Rp by definition, where RP and
Rv are the loading resistance of the antennas, with and without plasma,
respectively, and the PG is the generator output power. The NBI heating power PNB]
is defined by PNB] = PtOruS-Pioss> where Ptonis is the measured input power to thetorus and P,oss represents the shinethrough, ion orbit and charge exchange losses.
P,oss is calculated to be 0.1P,orus. The total stored energy WT is determined from
/3p [=A - ( ^ - l)/2] by a poloidal magnetic field fitting code using 24 magnetic
probes and eight saddle loops [1] and the value determined in this way is crosschecked by a diamagnetic measurement.
3.

ACHIEVEMENT OF H-MODE

JFT-2M has achieved the H-mode in the open divertor configuration with NBI
heating [2]. Later, H-like transitions were observed not only in ICRF heated, open
divertor discharges [3], but also in D-shaped, limiter bounded plasmas [4].
Figure 1-A shows an example for H-transition (thick line) with ICRF heating. Afterreaching the quasi-steady state, WT increases again simultaneously with an abrupt
decrease in the H a /D a line emission. Following the transition, the electron density
ne increases, which is in contrast to the case without transition. The radiation losses
PR after an initial small dip also increase strongly. As is shown in Fig. 1-A-e, anincreased emission from highly ionized metals near the centre correlates withincreased radiation losses in spite of a decrease in the emission from metals in the
low ionization states (Fig. 1-A-f). This indicates a decrease in the impurity influx andan impurity accumulation by the improved confinement. The dip in PR at the transition is due to the sudden decrease of the impurity influx, at first overcompensating
the increase of the radiation loss due to central accumulation.
In The case of D-shaped limiter discharges, an example for H-like transition is shownin Fig. 1-B (thick lines). For the H-transition in a limiter discharge, an inner shiftof the plasma is necessary. Following the abrupt decrease of the H a /D a line
emission, we also observe an increase in WT, n,. and PR. At the transition, PR also
dips briefly. These observations clearly show that the transition has the same properties as are observed in diverted discharges. In some cases, ELM-like spikes [5] in153

IAEA-CN-47/A-I1I-2(C)

FIG. 1. Temporal behaviour of plasma parameters of: (A) 0.74 MW, ICRF heated H-mode in single
null diverted discharge. IP = 220 kA, BT - 1.27 T, a = 0.30 m, K = 1.4 (note that line emission of
Fe X is obtained in different shots, where the H-transition occurs at different times); (B) H-like transitionin limiter D-shaped discharge with 0.65 MWNBI. 1P = 220 kA, BT = 1.18 T, a = 0.32 m, K = 1.4,
Tlb is measured at R = 1.03 m (limiter at R = 0.96 m); (C) Improvement after injection of single pellet
into 0.7 MW, NBI heated divertor discharge. IP = 220 kA, a = 0.26 m, K = 1.4. Traces of ne are at
central chord for the larger values and at r = 21 cm chord for the smaller ones.

the H^/DQ, emission are observed, and the appearance of the spikes suppresses thefurther increase in stored energy and radiation losses.
In Fig. 1-B-e, the electron temperature near the edge of the plasma Teb isshown. At the transition, Teb increases abruptly, which indicates the formation of atemperature pedestal. The line average density traces in Figs 1-A-c and 1-B-c suggest
the formation of a density pedestal: at the transition, the line density n<. at the central
chord increases gradually, while n,. at the r = 21 cm chord first increases strongly.
Later on, the increase becomes slower than that of the central chord.
The difference between the H- and L-modes in Fig. 1-A is the heating power,
and in Fig. 1-B is the strength of the magnetic field, BT. By increasing the ICRF
power from 0.70 MW to 0.74 MW in the former case, and by decreasing BT from
1.21 T to 1.18 T at the same plasma current in the latter case, H-transition occurs.
The minimum power required for the H-transition is 0.2 MW for optimized single

154

ODAJIMA et al.
0.6
Single-nulln e (iol9m~3)
FIG. 2. (A) Threshold power for H-transition Plh versus cylindrical q value, q^; (B) Gross energy
confinement time TEG as a function ofne in H- and L-modes. The improvement with pellet injection is
also plotted. IP = 220 kA, BT = 1.18 T, a = 0.26 m, K = 1.4, PNBI = 0.7 MW.

null diverted and 0.5 MW for limiter discharges and is roughly the same for NBI andICRF heating. A low qa value is more favourable to the transition and the threshold
power decreases as qa decreases as is shown in Fig. 2-A.

4.

IMPROVED CONFINEMENT WITH PELLET INJECTION

Confinement improvement is also observed transiently with pellet injection into
an NBI heated divertor discharge [6], as is shown in Fig. 1-C. If a single pellet isinjected before the H-transition occurs, the stored energy starts increasing linearly.
During the first 40 ms after pellet injection, about 70% of NBI power is stored in
the plasma. WT reaches 35 kJ, and then the plasma collapses, because of the growthof an MHD instability. The gross energy confinement time, rEG, defined as
TEQ = WT/(PT - WT), increases up to 70 ms (Fig. 1-C-e), which is about threetimes as high as Kaye-Goldston scaling [7] and is twice the Ohmic value at the samedensity. For a smaller pellet, WT also increases linearly with the same slope, butthen saturates abruptly, following the growth of an MHD fluctuation. This saturationlevel of WT is almost the same as the value achieved in the gas fuelled H-mode.

5.

CONFINEMENT PROPERTIES IN H- AND L-MODES

The gross energy confinement times, T EG , of the H- and L-modes are plotted
as a function of n^ in Fig. 2-B. The improvement by pellet injection is also shownin this figure. The NBI heating power is 0.7 MW, which is four to five times theJoule heating power. To obtain the L-mode at the same heating power for the sakeof comparison, a lower rail limiter is inserted near the separatrix. In the optimized
H-mode, TEG is almost the same as in an Ohmic plasma over the whole density
range, which is in contrast to the deterioration in the L-mode plasma.

155

IAEA-CN-47/A-IH-2(8)

PT(MW)
FIG. 3. (A) Total stored energy WT as a Junction of nc, for the scone conditions as in Fig. 2-B;
(B) Total stored energy WT as a function of total input power PT in H- and L-modes with NBI heated,single null divertor discharge, witha = 0.28 m and K = 1.45 (note that the L-and H-data are at different
plasma currents, but for lr > 200 kA, in the L-mode, the dependence of r^ on current
is weak).

If we study the confinement properties from the point of view of the total stored
energy, we see some features of the energy confinement. Figure 3-A shows WT as
a function of n,.. The dependence of WT on density with additional heating parallels
that in the Ohmic phase. This indicates that the following model is pausible:
WT = W 0H + Wad, where W O H is the stored energy of the Ohmic base plasma and
Wad is the incremental stored energy from the additional heating with its own
confinement time, rad (= AWT/APT|St=consl). Earlier experiments in JFT-2M haveshown that rad is almost indepent of n,. and I p (>200 kA) as well as of PT in the
L-mode [8]. Taking into account the above model, we infer, from Fig. 3-A, thefollowing features of energy confinement: (1) If h^ is affected by the heating, thenthe increase in WT becomes the sum of Wad and AWOH (because of its dependence
on density); therefore, the apparent increase in WT is affected by the change in n^;(2) in the high n,, region, where the increase in W 0H with % saturates and/or in thehigh Pad region, where Wad > W 0H , WT and rEG are independent of n^; (3) in the
low density region, where TOH « rad, no deterioration of rEG with additional heating
is seen, as compared to TOH> this can also be seen from the L-mode of Fig. 2-B,
where T EG , with additional heating, approaches TOH with decreasing %.
In the H-mode, the incremental stored energy, Wad, of the plasma can amountto the double value of that of the L-mode plasma, as is shown in Fig. 3-A. Figure 3-B
shows WT versus PT with fixed n,. om the H- and L-modes in NBI heated, diverted
discharges. The power dependence of WT is almost linear, and the slope gives
Tad = 27 ms and Tad = 12.5 ms, for the H- and L-modes, respectively. In the caseof Fig. 3-B, the H-transition, thus, improves rad by a factor of two. In the othercases, the improvement factor takes different values and is closely correlated with thedecrease in H a /D a emission. A large decrease in the H a /D o emission yields good
improvement, and a small decrease, or a spiked emission, results in poor
improvement.

156
6.

ODAJIMAetal.
SUMMARY AND DISCUSSIONS(1)

An H-mode plasma is demonstrated in NBI and/or ICRF heated open divertor
discharges, and an H-like transition with improved energy confinement is also
observed in limiter discharges with both heating methods, separately andcombined. Therefore, we may conclude that the H-transition does not dependon the microscopic nature of the heating method, and that a diverted magnetic
configuration is not a necessary condition for an H-transition.
(2) A large improvement in gross energy confinement is observed after injection ofa single pellet into an NBI heated plasma, where during the first 40 ms after
pellet injection 70% of the NBI power is stored in the plasma. TEG reaches
70 ms, which is twice the Ohmic value.
(3) At the H-transition, suddenly temperature and density pedestals are formed
before the total stored energy increases. This fact suggests that the energy andparticle transport near the plasma edge are abruptly improved at the transition.
If this is really the case, the H-transition provides evidence that the energy
confinement of the tokamak is determined by transport effects in the edge regionof the plasma [9, 10].
(4) In an additionally heated plasma, the incremental stored energy behaves as if
it had its own energy confinement time Tad, which is independent of the heating
power. In some cases to the H-mode, rad is doubled with respect to the
L-mode.

ACKNOWLEDGEMENTS
The authors are grateful to Drs M. Tanaka, K. Tomabechi, and S. Mori for continuous encouragement. They also thank Dr. J.-M. Noterdaeme for useful
discussions.

REFERENCES
[1]
[2]

[3]
[4]
[5]
[6]

SHOJI, T., et al., in Controlled Fusion and Plasma Physics (Proc. 11th Europ. Conf. Aachen,1983), Vol. 1, European Physical Society (1983) 55.
SENGOKU, S., et al., Confinement and fuelling studies during additional heating phase in
JFT-2M tokamak, in Plasma-Surface Interaction (Proc. 7th Conf. Princeton, 1986), to be
published in J. Nucl. Mater.
MATSUMOTO, H., et al., Improved Confinement during ICRF Heating on JFT-2M,
Rep. JAERI-M-86-147 (submitted to Nucl. Fusion).
SENGOKU, S., et al., Regime of improved confinement in neutral-beam-heated limiter dischargeof the JFT-2M tokamak, submitted to Phys. Rev. Lett.
WAGNER, F., et al., Phys. Rev. Lett. 42 (1982) 1408.
MIURA Y., et al., Characteristics of Improved Confinement and High Beta in Pellet and NeutralBeam Injected Single Null Divertor Discharges of the JFT-2M Tokamak, Rep. JAERI-M-86-148,
Japan Atomic Energy Research Institute, Tokyo (1986).

IAEA-CN-47/A-III-2
[7]
[8]
[9]
[10]

15 7

KAYE, S.M., GOLDSTON, R.J., Nucl. Fusion 25 (1985) 65.
ODAJIMA, K., et al., Scaling of incremental energy confinement time in the JFT-2M, to be
published in Phys. Rev. Lett.
OHYABU, N., et al., Nucl. Fusion 23 (1986) 593.
FURTH, H.P., Plasma Phys. Controll. Fusion 28 (1986)1305.

DISCUSSION
K. McGUIRE: In your H-mode limiter discharges, does the plasma contact the
inner limiter during the H-mode phase and is the inner shift of the plasma importantin obtaining the H-mode in limiter discharges?
K. ODAJIMA: Yes, in our H-mode limiter discharges the plasma is bounded
only by the inner limiter and is always in contact with the limiter during the H-phase.
But we cannot really say whether a point contact with the limiter or a separation fromthe outer limiter is important or not to obtain the H-mode.
J. JACQUINOT: Have your observed the H-mode with circular plasmas lying
on the inner wall?
K. ODAJIMA: Yes, we have observed H-transition in circular plasmas withICRF heating.
A. GONDHALEKAR: In your pellet injection experiments, there is an apparent
increase in the energy confinement time. The duration of the apparent increased
confinement is shorter than the confinement time itself. What significance do you
attach to this increase in energy confinement time?
K. ODAJIMA: The saturation value of WT before the collapse is almost thesame as that in gas fuelled H-mode discharges or slightly higher. Hence, no improvement has been brought about in the steady state phase with pellet injection so far.

IAEA-CN-47/A-HI-3

INITIAL RESULTS
FROM THE Dffl-D TOKAMAK*
J. LUXON,
P. ANDERSON,
F. BATTY1,
C. BAXI,
G. BRAMSON,
N. BROOKS, B. BROWN, B. BURLEY, K.H. BURRELL, R. CALLIS,
G. CAMPBELL,
T. CARLSTROM,
A. COLLERAINE,
J. CUMMINGS,
L. DAVIS, J. DeBOO, S. EJIMA, R. EVANKO, H. FUKUMOTO2, R. GALLIX,
J. GILLELAND, T. GLAD, P. GOHIL, A. GOOTGELD, R.J. GROEBNER,
S. HANAI3, J. HASKOVEC, E. HECKMAN, M. HEIBERGER, F.J. HELTON,
P. HENLINE, D. HILL 4 , D. HOFFMAN1, E. HOFFMANN, R. HONG,
N. HOSOGANE5, C. HSIEH, G.L. JACKSON, G. JAHNS, G. JANESCHITZ6,
E. JOHNSON, A. KELLMAN, J.S. KIM, J. KOHLI, A. LANGHORN, L. LAO,
P. LEE,
S. LIGHTNER,
J. LOHR,M. MAHDAVI,
M. MAYBERRY,
B. McHARG, T. McKELVEY, R. MILLER, C.P. MOELLER, D. MOORE,
A. NEREM,
P. NOLL7,
T. OHKAWA,
N. OHYABU,
T. OSBORNE,
D. OVERSKEI, P. PETERSEN, T. PETRIE, J. PHILLIPS, R. PRATER,
J. RAWLS, E. REIS, D. REMSEN, P. RIEDY, P. ROCK, K. SCHAUBEL,
D. SCHISSEL, J. SCOVILLE, R. SERAYDARIAN, M. SHIMADA5, T. SHOJI5,
B. SLEAFORD,
J. SMITH, Jr., P. SMITH, T. SMITH, R.T. SNIDER,
R.D. STAMBAUGH, R. STAV, H. St. JOHN, R. STOCKDALE, E.J. STRAIT,
R. STREET,
T.S. TAYLOR,
J. TOOKER,
M. TUPPER,
S.K. WONG,
S. YAMAGUCHI8GA Technologies Inc.,San Diego, California,
United States of America
Abstract
INITIAL RESULTS FROM THE DIII-D TOKAMAK.
The DIII-D tokamak was designed to exploit both (a) the strong dependence of limiting beta
values and plasma energy confinement on high current, and (b) the effectiveness of the divertor configuration in obtaining good confinement. Initial operation of the device has resulted in Ohmic confinement(r = 150 ms) comparable to that of other large tokamaks and in low levels of plasma impurities(Zefr = 1.5 at ne = 6 x 10" m~ 3 ). Both single null divertor and elongated dee shaped limiter
plasmas have been produced. The paper reports both attaining H-mode confinement during beam heatingof plasmas in open divertor configurations and values of the Troyon scaling quantity (I/aB « 2) large
enough to allow future studies of beta limits in regimes heretofore unattainable.
* Work sponsored by the United States Department of Energy, under Contract No.
DE-AC03-84ER51044, and by the Japan Atomic Energy Research Institute, under aco-operative agreement.
1

Oak Ridge National Laboratory, Oak Ridge, TN, USA.
On leave from Hitachi Ltd., Japan.
3
On leave from Toshiba Corp., Japan.
4
Lawrence Livermore National Laboratory, Livermore, CA, USA.
5
Japan Atomic Energy Research Institute, Japan.
6
Euratom-IPP Association, Max-Planck-Institut fur Plasmaphysik, Garching,Fed. Rep. Germany.
7
JET Joint Undertaking, Abingdon, Oxfordshire, United Kingdom.
8
On leave from Mitsubishi Electric Corp., Tokyo, Japan.
2

159

160
1.

LUXON.et al.

INTRODUCTION

The DIII-D device has been in operation since February 1986 [l],and is now operating at parameters comparable to other large tokamaks.
Doublet III [2,3,4] and other devices have demonstrated the role of plasma
current in attaining high beta and good confinement, and the importanceof the divertor in attaining regimes of good confinement with neutral beam
heating. The DIII-D was designed to exploit this understanding.
DIII-D is a non-circular cross section, low aspect ratio tokamak
designed to achieve high performance in a modest sized device (Fig. 1)[l]. The parameters of the device are presented in Table I. The indicateddesign capability of the device is attainable with additional shaping power
supplies, auxiliary plasma heating, and internal vessel protection armor.

DHMIC HEATING COIL

TOROIDALFIELD COILSPORTACCESS
LIMITER^PUMPED)TOROIDALFIELDCOIL BUS

COIL&.VESSELSUPPORT
VACUUM
VESSELL

2 m
POLOIDAL HELD COILS

FIG. 1. Cross-section of the DIII-D device. Areas not shown in gray are new. Superimposed is the
MHD equilibrium of a high beta divertor discharge (Fig. 3 at 2280 ms).

IAEA-CN-47/A-III-3

TABLE I.

16]

Dlll-D PARAMETERS

Major radius (m)Minor radius (m)Vessel elongation
Aspect ratio
Maximum toroidal field on axis (T)Maximum plasma current (MA)
Limiter
Divertor
Neutral beam power (MW)
Neutral beam pulse length (s)ICRH power at 30-60 MHz (MW)ECH power at 60 GHz (MW)

Oct.
1986

1987

DesignCapability

1.67
0.67
2.0
2.5
2.2

1.67
0.67
2.0
2.5
2.2

1.67
0.67
2.0
2.5
2.2

2.5
2.0
6
0.4
—
—

3.5
2.5
12
0.7
2
1.8

5.0
3.5
14
5
20
10

Neutral beam injection is the principal source of auxiliary heating
power. The beamlines have been reconstructed to improve beam transmission and to accommodate the new 5 sec sources which are now being
delivered. The beamlines inject at an average angle of 19.5° from radialin the co-direction. The results reported here were obtained using two of
the beamlines fitted with 0.5 sec sources producing up to 6.0 MW of powerat energies up to 70 keV for 0.45 sec pulses into the plasma. Conversion
to long pulse sources with a capability of up to 14 MW of power should
be complete in the spring of 1987.
2.

NON-CIRCULAR PLASMA PRODUCTION AND SHAPING

Both divertor and limiter plasmas with elongations up to 2.1 havebeen formed in the device (Fig. 2). The shape and position of the plasma
is controlled by the 18 independently powered shaping coils. The low
error fields and precise field nulling which can be attained with the coilsystem allows discharges to be routinely initiated with one-turn voltagesof 11 volts or less. Owing to the small amount of field due to the ohmic
heating coil in the plasma region and the precise control afforded by theplasma shaping coils, the chosen plasma configuration can be maintained
over the entire duration of the discharge. The position, elongation andtriangularity are separately controlled, and the plasmas exhibit no verticalinstability over the range of configurations tested (K — 1.2 to 2.1).

162

LUXON et al.

D
D

O

CT
(c)
FIG. 2. MHD equilibrium cross-sections for Ohmic discharges produced in DIII-D. (a) K = 1.2
limiter discharge, (b) K = 2.1 dee shaped discharge, (c) double null divertor, (d) single null divertor.

Single null divertor configurations have been obtained with elongations K from 1.6 to 1.97 and triangularities of up to 0.35. The separatrix
flux surfaces intersect two toroidal bands of graphite armor at the bottomof the vacuum vessel (see Fig. 1). The spacing between the X-point andthe vessel bottom is controlled to an accuracy of 1 cm within a range offrom 8 to 30 cm. Currents of as high as 2.0 MA have been achieved indivertor configurations, and currents of 1 MA have been maintained for
4 sec. Values of I/aB of up to 1.9 have been attained (q* = 3.3), making
them particularly attractive for auxiliary heating. Double null divertor
configurations have also been produced, but extensive use of these configurations must await the installation of protective armor in the upper
inside corner of the vacuum vessel.
Limiter plasma configurations have been obtained with elongationsof from 1.2 to 2.1. These discharges have been run on both the insidewall graphite armor and the outside limiters. A current of 2.5 MA has
been achieved using the dee-shaped configuration of Fig. 2 with a flattop
duration of 0.5 sec. This resulted in I/aB = 1.8 with q$ = 4 . 1 andtriangularity of 6 = 0.5.
The flux necessary to reach a given plasma current is supplied byboth the ohmic heating coil (total capacity 12 V-s) and the plasma shaping
system which contributes about 1.0 V-s/MA essentially from the verticalfield. The flux required from the ohmic heating coil transformer is found tobe 2.1 V-s/MA and constant over a variety of plasma shapes and currents.
This indicates adequate flux is available to reach plasma currents of 5 MA.

163

IAEA-CN-47/A-HI-3

3.

CONFINEMENT AND PLASMA HEATING

Vessel cleanup has progressed significantly in the last nine months.
The vessel is constructed of Inconel with Inconel protective plates (35% of
surface) and graphite armor (15%) used to protect the interior surface ofthe vessel in areas of high heat flux. The limiters consist of two 0.4 X 0.8 m
shaped graphite blades located on the outside midplane. Vacuum pumping
rates of up to 10,000 tjs are obtained using two turbo molecular and two
cryo pumps. The partial pressure of non-hydrogen components stands at~ 5 x 10~9 torr with a leak rate for air of less than 5 x 10~5 torr-t/s.

Ip

VOLTAGE(MA)

0
6
GAS PUFF(torr-Ls" 1 )

0
0.4

0
1.4

0.0
1

0.0
8

'V.

•KINETICMHO
TOTAL
ENERGY(MJ)

NEUTRAL
BEAM
POWER
(MW)

0
S
T e (kV)
0200

DIAMAG

100

MHD

(ms)
0

1900

2100

2300
2500
TIME (ms)2700

FIG. 3. Time history of a beam heated discharge exhibiting H- and L-mode behaviour. The start ofthe current rise is time = 0.

164

LUXON et al.

In
85-

•
_ •

••

o
a. •»-

0 °° 0
O

°
-

o

°

5

as-

•

n

0

CQ
_J

•
U

•--

0

o = H-MODE• = L-MODE
1

2

3

4

ne(10"m"') BEFORE NBI
FIG. 4.

Characterization of the H-mode threshold for divertor discharges with Ip = 1.0 MA,

Hydrocarbons cleaned up rapidly with baking to 400°C. Baking, glow
discharge cleaning, and rapid pulse discharge cleaning are used to clean
up water and other sources of oxygen. Substantial cleanup time including
tokamak discharges is required after the vessel is exposed to air. Afterone to two months, Z€ff = 1.5 can be attained in ohmic discharges at
ne ~ 3 - 6 x 1019 m~ 3 . The principal impurity is oxygen. No gettering of
the vessel walls has been done.
Ohmic DIII-D discharges exhibit the characteristics of clean tokamak discharges. The ohmic confinement observed in DIII-D is comparable
to that seen in other larger tokamaks. The confinement in ohmic hydrogen discharges is estimated from magnetic measurements to be TE =150 ± 50 ms for densities in the range ne = 0.7 - 5 x 1019 m~3. This
produces values of TEJne which equal or exceed those expected from thescaling TE/ne = 1.92 .ft2'04 a1-04 [5]; the explicit dependence of r on n has
not been demonstrated from the limited operation to date.
Initial application of up to 6.0 MW of hydrogen neutral beam power
to single null diverted deuterium discharges (BT = 2.1 T) has resulted in
attaining regimes of good confinement commonly referred to as H-mode[6]. The time history of the parameters for an H-mode discharge is shownin Fig. 3, and the MHD equilibrium is shown in Pig. 1. The plasma
current is maintained constant, and the one-turn voltage is measured at
the surface of the plasma. The onset of the H-mode is characterized by a
rapid, to date uncontrolled, rise in the line average plasma density. The
gas puff to the plasma was held constant. The Da radiation in both thedivertor X-point region and the central plasma also drops to a fractionof the ohmic value. The central electron temperature is seen to rise, and165

IAEA-CN-47/A-m-3

the total energy rises dramatically as the global energy confinement time
returns to values comparable to the ohmic confinement.
A well defined threshold in both plasma density and neutral beam
power must be exceeded in order to attain H-mode confinement, otherwise
poor (L-mode) confinement is obtained (Fig. 4). At a current of 1 MA
the power threshold is less than 2.8 MW and the density must exceed
2 x 10 19 m~ 3 . The transition to the H-mode is delayed from the initial
application of beam power. This period is characterized by little density
rise, a sharp increase in Da radiation, and a period of poor (L-mode)
confinement.
The ion temperature profile % (r) from charge exchange recombination and the electron temperature profile Te (r) from absolutely calibrated
electron cyclotron emission are shown in Fig. 5 at 2280 ms. The line average density is determined from a single vertical CO 2 interferometer chord.
The absolutely calibrated visible bremsstrahlung radiation (dependent on
Te,ne, and Ze/f) is measured along 16 chords. These data were combined
along with the assumption Zeff = const, to obtain the set of profiles shownin Fig. 5 for the ohmic (t = 2000,n e = 2.5x 1019 m~ 3 , Zeff = 1.5), and Hmode (t = 2280, ne = 5.2 x 1019 m~3, Zeff = 2.9) phases of the discharge.

-• _,.,.
1

1

0.2

0.4

1

0.6

0.8

r/a
FIG. 5. Temperature, density, and radiated power profiles:H-mode from the discharge of Fig. 3 at
2280 ms; Ohmic at 2000 ms. L-mode is from a similar discharge (P = 4.5 MW).

LUXON et al.

166
200

•

O=H-MODE•5 150

'
.

ui

O

8

•

• = L-MOOE

00 A* H-MODE

o

<9o

9\
\

WITH
STRONG
ELM'S

AA'_ Q \AA

100 UJ

o

;
50

\

•

Ip - 1.0•

Bt - 2 . 1 0\

KAYE-GOLDSTON L-MODE

Kappa - 1.75
,

'

4

•

•

, . ! . . . •

•

6

8

10

TOTAL POWER (MW)FIG. 6. Confinement time for L-mode and H-mode discharges and a comparison to Kaye-Goldston
scaling for divertor discharges with lp = 1 MA and BT = 2.1 T.

Data for a similar L-mode discharge (n c = 1.8 x 1019 m~3, Zeff = 4.6) isalso shown. During the H-mode, the estimated peak kinetic total energy
including calculated fast ion contributions is 700 ± 140 kJ and is indicatedin Fig. 3. This results in /3 = 1.4% and (Bp - 1.1.
Tiie radiated power increases above ohmic for both H- and L-mode
with neutral beam heating. The radiated power distribution for the main
body of the plasma is shown in Fig. 5. In L-mode the radiation remains
centrally peaked, accounts for ~ 35% of the power, and is about equallydistributed between the main body of the plasma and the divertor region.
In the H-mode ~ 45% of power is radiated with the majority of the power
being radiated from the main body of the plasma.
Preliminary indications are that the scaling of confinement time withpower in the L-mode follows that expected from Kaye-Goldston L-mode
scaling [7]. Figure 6 illustrates this for a case with current fixed at 1 MA,
BT — 2.1 T and K = 1.75. However, once the H-mode is established,confinement improves dramatically reaching values of as much as threetimes the L-mode confinement and equaling or exceeding the initial ohmic(P ~ 1 MW in Fig. 6) confinement. The confinement times indicated here
are global energy confinement times based on magnetic measurements andcorrected for the time rate of change of the energy. Those discharges withthe best confinement do not reach steady state with the available beampulse durations.

IAEA-CN-47/A-m-3

167

Spikes occur in the Da radiation during the H-mode coincident withthe rapid expulsion of substantial amounts (up to 15%) of energy and particles. These losses appear to propagate from the edge into the plasma.
The discharge described in Fig. 3 only exhibited this feature after the
power was turned off. Initial evidence is that these bursts are likely associated with edge localized modes (ELM) [6]. The highest values of total
plasma energy WT occur in discharges with no Da bursts.
Additional information on the nature of the ELM can be obtained
from ICH coupling experiments carried out using cavity wave launchers
constructed by Oak Ridge National Laboratory. Satisfactory loading canbe achieved in both ohmic limiter and divertor plasmas, but with H-mode
heating the loading drops by a factor of ~ 2 from the L-mode coupling.
However, this loading is seen to increase abruptly by a factor of 2 or morecoincident with the Da bursts. These results suggest particularly low
density in the region outside the separatrix during H-mode and a sharp
rise in this density associated with the ELM.
Once H-mode confinement has been attained, it remains for theentire duration of the neutral beam pulse. H-mode confinement has beensustained for up to 740 ms with 3 MW of power applied. Burst free Hmode has been sustained for 300 ms before the heating ended. H-modes
have been achieved over the entire current range 0.7 < Ip (MA) < 2.0 andtoroidal field range 1.2 < BT (T) < 2.1 studied to date. The maximum
line average denisty attained is 8.4 x 1019 m~ 3 .

4.

SUMMARY

The first nine months of DIII-D operation have demonstrated thatthe tools necessary to carry the study of confinement and beta into new
regimes will be available in the upcoming years. Discharges with good (Hmode) confinement have been demonstrated with neutral beam heatingin the open divertor configuration. H-mode confinement times are comparable to ohmic confinement times and about three times the L-mode
confinement (rg ~ r £ H ~ 3 rjf ~ 175 ms). This result clearly shows that
this important confinement improvement can be scaled to larger devices.
It also shows that this enhanced confinement can be achieved in simplified
divertor configurations which can be more readily incorporated into larger
devices. The substantial range of divertor shapes available, including the
double null, along with the precise control demonstrated by the shaping
system, should allow the role of cross sectional shape in attaining good
confinement to be thoroughly explored.

168

LUXON et al.

DIII-D has demonstrated the tools necessary for an aggressive studyof plasma shape in attaining high values of beta. Values of I/aB ~ 2 havealready been attained in both limiters and divertors, and substantiallyhigher values are anticipated in the future. The plasma cross section canbe precisely controlled in both space and time over a range of elongations
from near circular to elongations greater than two. Triangularity can be
independently varied. With the addition of the new long pulse sources
beam heated discharge durations are anticipated to be sufficiently long toattain quasi steady state conditions.
It is anticipated that during the next year the tokamak will be operated in regimes of high current and higli I/aB (Table I) and that all of
the neutral beamlines will be brought into operation and fitted with longpulse sources capable of a total of 14 MW of power with multisecond pulse
durations. This capability should allow extension of confinement studies
to substantial values of current and power and a definitive test of the beta
limits of the machine.
ECH outside launch heating experiments in both the extraordinary
and ordinary mode are planned through early 1987. This system will
be capable of 1.4 MW of power at 60 GHz and will allow profile control
experiments using edge heating to be carried out. Installation of insidelaunch ECH with a total of 2 MW of power is anticipated in late 1987.
This will allow both edge heating and current drive experiments. The
present ICH coupling experiments will be complete in early 1987, and
2 MW of heating using cavity resonators to couple power at 30-60 MHz
is planned for late 1987.
REFERENCES
[1]
[2]
[3]
[4]
[5]
[6]
[7]

LUXON, J.L., DAVIS, L.G., Fusion Technol. 8 (1985) 441.
STAMBAUGH, R.D., in Plasma Physics and Controlled Nuclear Fusion 1984 (Proc. 10th Int.
Conf. London, 1984), Vol. 1, IAEA, Vienna (1985) 217.
BURRELL, K.H., ibid., p. 131.
KITSUNEZAKI, A., ibid., p. 57.
PARKER, R.R., et al., Nucl. Fusion 25 (1985) 1127.
WAGNER, F., et al., Phys. Rev. Lett. 49 (1982) 1408.
KAYE, S.M., et al., Nucl. Fusion 24 (1984) 1303.

DISCUSSION
Y. SHIMOMURA: Did you observe the H-mode in the hydrogen plasma?
J. LUXON: No, we have not yet had an opportunity of doing so.
M. KEILHACKER: I would like to make a comment. The recent observationof H-mode discharges in large tokamak experiments, such as JET and DIII-D, allows

IAEA-CN-47/A-III-3

169

the size scaling of H-mode confinement times to be studied. A comparison of DIII-D
and JET results with those from ASDEX seems to indicate that the global energy confinement time of H-mode discharges scales linearly with total plasma current, but has
no additional separate size scaling.
F. WAGNER: Dr. Luxon, you mentioned in your presentation that the H-mode
is often preceded by an L-phase. Having in mind the previous D-III results where
the L- and H-modes were different, from the very beginning of the beam pulse, I
would like to know whether there are still cases without a preceding L-phase. Do you
attribute the difference in H-mode development between D-III and BIG-D to thechange in the ion grad-B drift direction?
J. LUXON: At present, the onset of the H-mode is always preceded by someperiod of delay, although the duration varies. The ion grad-B drift is into the divertor
for DIII-D and away from the divertor for Doublet III. But I think that there are somany other differences that we should carry out the experiment rather than infer the
result.

IAEA-CN-47/A-m-4

PELLET INJECTION RESULTS DURING
TFTR OHMIC AND NEUTRAL BEAMHEATING EXPERIMENTS*
G.L. SCHMIDT,
S.L. MILORA1,
V. ARUNASALAM,
M.G. BELL,
M. BITTER, C.E. BUSH1, S.K. COMBS1, H.F. DYLLA, P.C. EFTHIMION,
E. FREDRICKSON,
R.J. GOLDSTON,
B. GREK,
L.R. GRISHAM,
G. HAMMETT,
R.J. HAWRYLUK, W.W. HEIDBRINK2, H.W. HENDEL,
K.W. HILL, H. HSUAN, D.W. JOHNSON, L.C. JOHNSON, P.H. LaMARCHE,
R. LITTLE, D.K. MANSFIELD, D.C. McCUNE, K. McGUIRE, D.M. MEADE,
S.S. MEDLEY, D. MUELLER,
W. MORRIS3,
E.B. NIESCHMIDT4,
J. O'ROURKE 5 , D.K. OWENS, H. PARK, A.T. RAMSEY, M.H. REDI,
J. SCHIVELL,
D. SCHURESKO1,
T. SENKO,
S. SESNIC,
F.J. STAUFFER6,
B.C. STRATTON,
G.D. TAIT,
G. TAYLOR,
J.R. TIMBERLAKE,
H.H. TOWNER,
M. ULRICKSON,
R. WIELAND,
J.B. WILGEN1, M. WILLIAMS, M.C. ZARNSTORFF, S.J. ZWEBEN

Princeton Plasma Physics Laboratory,
Princeton University,
Princeton, New Jersey,
United States of America
Abstract
PELLET INJECTION RESULTS DURING TFTR OHMIC AND NEUTRAL BEAM HEATING
EXPERIMENTS.
In Ohmic discharges, pellet injection experiments on TFTR have increased density limits andenergy confinement. Injecting multiple pellets sustains the improved confinement and increases the central density and peaked character of the profile, producing significant increases in ne(O)rE(a). A valueof (1.4 ± 0.3) x 1020 m"3-s has been achieved at a central electron temperature of 1.4 keV. Decayof the central density for strongly peaked profiles in Ohmic discharges is consistent with central particle
diffusion coefficients of 0.1 m2/s and flow velocities on the order of neoclassical velocities. Concentrations of scandium injected following pellet injection (nSc/ne) exhibit an extended rise time and thenremain roughly constant during the subsequent slow decay of the electron density. In pellet-fueled discharges heated by tangential neutral beams at high density, heating profiles vary from a fiat radial shape
to a profile that is peaked near the q = 2 radius. Global confinement is comparable to that obtainedin gas-fueled discharges with similar total input power but with heating profiles more strongly peakedon axis. Central confinement in pellet-fueled discharges exceeds values obtained in gas-fueleddischarges.
* Research sponsored by the Office of Fusion Energy, United States Department of Energy, under Contract No. DE-AC02-76-CHO-3073 with Princeton
University and under Contract No. DE-AC05-84OR21400 with Martin
Marietta Energy Systems, Inc.
1

Oak Ridge National Laboratory, Oak Ridge, TN, USA.
GA Technologies Inc., San Diego, CA, USA.
3
Balliol College, Oxford University, Oxford, United Kingdom.
4
EG&G Idaho, Inc., Idaho Falls, ID, USA.
5
JET Joint Undertaking, Abingdon, Oxfordshire, United Kingdom.
6
University of Maryland, College Park, MD, USA.
2

171

172

SCHMIDT etal.

1. INTRODUCTION
Fueling by injection of single and multiple pellets of frozen deuteriumat speeds up to 1500 m/s has been carried out in TFTR Ohmic andneutral-beam-heated discharges using gas-driven injectors developed at
Oak Ridge National Laboratory [1,2]. A carbon inner wall with a large
surface area, conditioned by extensive operation with helium plasmas,
has been used as the discharge limiter. Operation in Ohmic discharges
at n e > 2 X 10 19 m"~3 is characterized by minimal metal impurities andsignificant pumping of recycling plasma [3].
2. OHMIC DISCHARGES
2.1. Overview
Injection of multiple pellets produces increased penetration of successive pellets as the plasma is cooled by the introduction of pellet mass.
The final pellets penetrate to the plasma magnetic axis, as illustrated inFig. 1. Penetration calculations, based on a neutral gas shielding model
modified to include a Maxwellian distribution of electrons, a cold plasma
shield extending along the magnetic field lines, and a finite electron heatbath,reproduce this effect and correctly predict penetration at 3 keV.
Deep pellet penetration produces strongly peaked density profiles athigher line-average density than is attainable by gas puffing. Central
densities up to 4 x 1020 m~ 3 have been produced. This is roughly fourtimes the highest value obtained with gas puffing [4j. Plasma line-averagedensities of 1.4 x 1020 m~ 3 have been obtained, corresponding to a valueof neR/BT = 6.5 x 10 19 m " 2 ^ " 1 , twice that achieved with gas puffing.
As illustrated in Fig. 1, central electron temperatures decrease fromapproximately 3 keV to <1 keV immediately following the injection
sequence. Temperature profiles broaden, with volume-average electron temperatures decreasing by a smaller factor than central temperatures. However, plasma pressures increase with pellet injection, becoming strongly peaked during the initial reheat. Neutron rates up to
5 x 10 13 N/s have been obtained in 2.2-MA discharges, roughly four times
the maximum rate in gas-fueled discharges.
Changes in discharge MHD behavior have been associated with thesestrong temperature perturbations in smaller devices with shorter inductive time-scales [5j. Similar effects are observed in TFTR. During pellet
ablation (0.5 ms) a high-frequency fluctuation, / > 160 kHz, is observedon Mirnov loops located at the vessel wall. Following this period, a 1/1
mode, possibly associated with poloidal and toroidal propagation of thepellet density perturbation [6,7], is often observed. The mode decays on
a 1- to 10-ms time-scale ana propagates at roughly the ion sound speed.
During an extended period following the pellet perturbation, sawtoothoscillations either can dominate the MHD behavior or are suppressed.
When sawtooth oscillations are dominant, they appear shortly after thepellet event, are of large amplitude, and have a period of roughly 100 to
200 ms, compared with 20 to 50 ms before pellet injection. In discharges
associated primarily with strong pellet penetration, sawtooth oscillations

173

IAEA-CN-47/A-III-4

OJ

O

_

5

2.0

2.5

3.0

2.5

R (m)
I

I

1

1

3.0

R (m)

I

I

1

I

1

1

1

1

/

5
R

2.3

o

t-

2.4

/

J
2.5

5

I

\

wi

2.6

/

2.7
R (m)

X
2.8

2.9

3.0

FIG. 1. (a) Density profile at 3.35 s, after injection of five 2.67 mm pellets, (b) Electron temperature
profiles at 2.15 s, before injection of first pellet; at 3.13 s, before injection of fifth pellet; and at
3.35 s. (c) Measured light during pellet ablation, indicating penetration depths of the first and fifthpellets. Calculated penetration depths, allowing for a 10% uncertainty in electron temperature, areindicated by the bars.

can be suppressed for a period of 1 s. These discharges either exhibit
no mode structure or are dominated by strong, continuous m = 1 oscillations. In these cases a large and very narrow central x-ray peakforms during the plasma reheat. During this period, off-axis, sawtoothlike relaxations that do not alter the central peak can be observed. Such
events imply a broad, flat ^-profile within the plasma core. In some discharges that initially exhibit these characteristics, sawtooth oscillations
can reappear. When this occurs, the continuous m = 1 oscillations cease,the formation of the strong central x-ray emission is terminated, and thedischarge becomes sawtooth dominated.
2.2. Particle transport
Following pellet injection, particle content in strongly peaked discharges decays on a 1- to 2-s time-scale, while central density initiallydecays on a time-scale greater than 2 s.
A strong interior density
gradient forms during the first phase of this decay, as does a density
pedestal with the approximate profile shape of gas-fueled discharges.
Such a case is shown in Fig. 2. The central density profile decay underthese conditions is reproduced by calculations assuming a fixed particle
diffusion coefficient of 0.1 m 2 /s for r/a < 0.5 [D =• 0.1 + {r/a)e m 2 /s].

174

SCHMIDT et al.

2.5

2.0

(.5

1.0

0.5

2.0

2.5
3.0
MAJOR RADIUS (m)

3.5

4.0

FIG. 2. (a) Electron density profile evolution following pellet injection, obtained from Abel inversionof FIR laser interferometer data (10 channels), (b) Measured and calculated decay ofne at four radial
positionsr: ±0.08m (A), ±0.2m (B), ±0.3 m (C), and ±0.5 m (D), averaged about the plasma center.

The inward pinch term determined by this diffusion coefficient and themeasured ratio D/v evaluated at 3.7 s is less than 0.3 m/s within this region. This value is roughly only five times the neoclassical value. These
terms represent a time average of the transport. They are consistent
with results from a similar discharge based on a least-squares fit to a and0 in the equation T = an + 0 dnjdr. Data are fitted over a range oftime separately at a given radius; F is obtained from a calculation using
measured values of ne(r,t) and a computed source term. As indicated in
the figure, the initial transport deviates from the average, with density
decaying slowly at the center and more rapidly at larger radii, suggestingsome variation about the average values.
Impurity transport during this decay has been evaluated with scandium injection. Scandium concentrations {nsc/ne) within the plasmacore have been deduced from pulse-height analyzer measurements, assuming the same profile shape for scandium and electron densities within
this region. When scandium is injected following pellet injection, concentrations are found to increase over an extended period of 300 to 500 ms
after scandium injection and then to remain roughly constant during thesubsequent slow decay of the electron density.
2.3. Energy confinement
Confinement properties of pellet-fueled and gas-fueled discharges
have been compared using (1) magnetic and kinetic measurements ofthe plasma stored energy; (2) calculations of the resistive input power,
including corrections due to changes in the internal plasma inductanceobtained from measurements of the diamagnetic flux /?p,diamag(0 andthe equilibrium magnetic field A((); and (3) calculations of the global
confinement, including corrections to the input power term due to thetime rate of change in stored energy. Comparisons were made primarily at 1.6 MA, where operation with deuterium gas puffing was

175

IAEA-CN-47/A-ID-4• PELLETS
o D2 GASo He GAS51015n e (10 19 m"3)
FIG. 3. Global energy confinement in deuterium and helium gas-fueled and deuterium pellet-fueleddischarges. Plasma stored energy was obtained from /S aiamag-

1.0

PELLET SHOT
0.5

PELLET INJECTION
0 " . . . . I • + • . • I .».

. * I .+.

3
2
1

NO PELLET SHOT
. I . . , . I , , , , I , ,,

0
0.7
0.5

NO PELLET SHOT
1.5

I . . , , I . . . , I
2.0
2.5
3.0

3.5

4.0

TIME (s)FIG. 4. Time history of plasma stored energy Elot, calculated input power, and global confinement
for a deuterium gas-fueled discharge and a deuterium pellet-fueled, 1.6 MA discharge.

176

SCHMIDT etal.

limited to n e < 5 x 1019 m " 3 by "Marfe-like" phenomena and highradiative power. Operation in helium extended the density limit to~ T x 10 19 m~ 3 . As shown in Fig. 3, although the confinement times
for gas puffing saturate at n e « 4 x 1019 m~ 3 , the pellet-fueled discharges exhibit improved energy confinement at substantially higher densities. The data are compared with Goldston empirical scaling [8] atfixed Pt o t = 1.6 MW. The gross energy confinement time is expressed interms of Ohmic (TNA) and auxiliary (r aux ) heating components, l/r|, =r § A + r a \ x , where r NA = 7.1 x 10- 22 fi e (cm- 3 )a(cm) 104 iZ(cm) 2>04 g 0 - 5 ,r au x = [6.4 x 10- 8 C/2 1 - 75 a-° :}7 / p (A)/ J Ptot(W) 0S ] 2 /r £?) and C is a constant determined by the confinement regime and the plasma isotope effect. Good agreement with the pellet fueling data is obtained for an"H-mode-like" confinement factor of C = 2.66 [4j. The gas fueling confinement times are more typical of L-mode results (C = 1.33).
This improved confinement appears qualitatively similar to the
Alcator-C results [9j. However, if ion transport is neoclassical, ion heat
conduction losses represent only ~15% of the total power flow at the halfradius. Plasma convective losses and uncertainties in the radiated powerfrom within the plasma core, which is a significant part of power balance,
affect the accuracy with which the upper limit of the ion neoclassicalanomaly factor can be determined. At present, we estimate the upper
limit of this factor to be ~ 2 - 3 , and we cannot determine whether ions orelectrons play the dominant role in heat conduction losses. Calculations
were made for a representative experiment, fueled by three pellets, using
the BALDUR code with models for radiative and convective power losses
corresponding to the measured radiation and density decay and for electron transport based on dissipative trapped electron modes [10]. These
calculations simulate the measured global confinement with a neoclassicalanomaly factor of 1.
The improved confinement obtained with pellets can be sustained byinjecting multiple pellets, as shown in Fig. 4. The plasma stored energy
rises with successive pellets, the global confinement increases with thefirst pellets and then is sustained by continued injection, and the central
density increases. The increase in stored energy is partially due to higher
resistive power, but the primary contribution is a roughly 70% increase
in TB, which in this case remains above the 0.5-s level for 0.7 s.
3. AUXILIARY HEATING
Auxiliary heating on TFTR is currently produced by tangential injection of 80- to 100-keV deuterium beams. Heating powers up to 11 MW
have been used for pellet experiments. At high density, beam geometry
and attenuation result in modest central input power levels even at hightotal input power. Heating profiles vary from flat to strongly peaked near
the radius at which q = 2. Nonetheless, global confinement is comparable
to that obtained in gas-fueled discharges with similar total input power
but with heating profiles more strongly peaked on axis. This general result suggests a radial dependence of the energy confinement, which has
been confirmed by detailed analysis of results at 6 MW [4]. Application

IAEA-CN-47/A-III-4

177

of ion cyclotron heating in the coming year will provide increased central
input power levels in pellet-fueled discharges, permitting tests of possible
enhanced confinement scenarios [ll].
4. SUMMARY
Pellet injection experiments on TFTR have increased density limits
and energy confinement in Ohmic discharges. Injecting multiple pellets sustains the improved confinement and increases the central densityand peaked character of the profile, producing significant increases inn e (0)r(o). A value of (1.4 ± 0.3) X 1020 m~3-s has been achieved at a
central electron temperature of 1.4 keV. Decay of the central density forstrongly peaked profiles in Ohmic discharges is consistent with central
particle diffusion coefficients of 0.1 m 2 /s and with flow velocities on theorder of neoclassical velocities. Concentrations of scandium injected following pellet injection (nse/ne) exhibit an extended rise time and thenremain roughly constant during the subsequent slow decay of the electron
density.
REFERENCES
[1]
[2]
[3]
[4]
[5]
[6]
[7]
[8]
[9]

[10]
[11]

COMBS, S.K., MILORA, S.L., FOUST, C.R., et al., Rev. Sci. Instrum. 56 (1985) 1178.
COMBS, S.K., MILORA, S.L., "Pneumatic pellet injectors for TFTR and JET", Fusion Technology (Proc. 11th Symp. Avignon, 1986), in press.
HILL, K.W., TFTR Group, Paper IAEA-CN-47/A-IV-2, these Proceedings, Vol. 1.
MILORA, S.L., SCHMIDT, G.L., ARUNASALAM, V., et al., Plasma Phys. Controll. Fusion
28 (1986) 1435.
MILORA, S.L., FOSTER, C.A., THOMAS, C.E., et al., Nucl. Fusion 20 (1980) 1491.
HEIDBRINK, W.W., et al., Nucl. Fusion 24 (1984) 3.
PARE, V.K., COLCHIN, R.J., MILORA, S.L., BELL, S.D., FREDRICKSON, E.,
McGUIRE, K., SCHMIDT, G.L., Bull. Am. Phys. Soc. 30 (1985) 1522.
GOLDSTON, R.J., Plasma Phys. Controll. Fusion 26 (1984) 87.
GREENWALD, M., GWINN, D., MILORA, S., et al., in Plasma Physics and Controlled
Nuclear Fusion Research 1984 (Proc. 10th Int. Conf. London, 1984), Vol. 1, IAEA, Vienna(1985) 45.
TANG, W.M., BISHOP, C M . , COPPI, B., et al., Paper IAEA-CN-47/A-VI-1-2, these
Proceedings, Vol. 1.
FURTH, H.P., Plasma Phys. Controll. Fusion 28 (1986) 1305.

DISCUSSION
F.X. SOLDNER: You told us that you compared the confinement times fromthe experiment with different scalings. Which scaling law did you apply for the
H-mode in this case?
S.L. MILORA: We used Goldston scaling, with an auxiliary heating confinement time multiplier of 2.66 to account for pure deuterium operation (x 1.33) andH-mode confinement (X 2).

178

SCHMIDT etal.

A. GIBSON: Is it still true that the highest ftDTETj in TFTR is obtained with
Ohmic heating and pellet fuelling, and without additional heating?
S.L. MILORA: The two experiments exhibit nearly identical values of nerETj.
However, the deuteron concentration on axis is higher in the pellet fuelling case andconsequently its nDTETj value is still higher.
J.G. CORDEY: Do your electron temperature profiles show profile consistency
for the case of pellet injection and neutral beam injection?
S.L. MILORA: Profile consistency applies in the case of moderate density
pellet fuelled discharges with neutral beam injection. At higher densities, the electron
temperature profiles can become hollow in the centre since radiation from low-Z
impurities exceeds the input power density.
R.S. PEASE: How do the results you report depend on the current used? And
how close can you get to the value 0 P = 1?
S.L. MILORA: Similar values of the confinement time, namely 0.5-0.6 s, havebeen attained at higher currents (up to 2.2 MA). These are lower than we expected.
However, the experiments were performed with one or two pellets only and so wecannot say that we have reached the peak plasma energy. For the 1.6 MA discharges,
0 P = 0.31.

IAEA-CN-47/A-IH-5

FAST ION LOSSESAND PLASMA CONFINEMENT IN TFRDURING NEUTRAL BEAM INJECTION
EQUIPE TFRAssociation Euratom-CEA sur la fusion,
Departement de recherches sur la fusion controls,
Centre d'etudes nucleaires de Cadarache,Saint-Paul-lez-Durance, France
Presented by J. -P. Roubin
Abstract
FAST ION LOSSES AND PLASMA CONFINEMENT IN TFR DURING NEUTRAL BEAMINJECTION.
Neutral beam injection experiments on carbon limited discharges in TFR exhibit a strong ion
heating saturation with increasing additional power and negligible electron heating. These results seemto be inconsistent with the usual L-mode scaling. To identify a possible beam specific mechanism
limiting the power before complete thermalization of the fast ions, an exhaustive check of the whole
coupling chain from the neutral beam sources to the bulk plasma has been made. The coupling of theneutral beams with the plasma is analysed and shown to be as expected. Fast ion thermalization isinvestigated by different methods (active charge exchange, neutron production in D ° - D experiments).
It is concluded that a spatial diffusion mechanism causes saturation of the central fast ion density withpower at the very beginning of the slowing-down process. Finally, the heating results are interpreted,taking into account the fast ion losses. While the electron confinement appears to deteriorate, the central
ion confinement time remains constant with power.

INTRODUCTION
Recent NBI experiments [1] on carbon limited discharges in TFR, in thePadd/POhm range < 3 , exhibit a strong ion heating saturation with increasing additional power density and negligible electron heating. These results seem to beinconsistent with the usual L-mode scaling. When the classical deposition profile
computed by Monte Carlo or Fokker-Planck codes is used, energy transport analysis
shows a strong decrease of T. compared with the Ohmic level (up to a factor of threeat high power). To identify a possible beam specific mechanism limiting the power
before complete thermalization of the fast ions, an exhaustive check of the whole
coupling chain from the neutral beam sources to the bulk plasma has been made.
179

180
1.

EQUIPETFR
BEAM INPUT ACCOUNTABILITY

1.1. Power transmission into the torus
A water cooled calorimeter has been installed on the inner wall of TFR, in frontof one neutral beam port. With this calorimeter the total space and time integrated
beam power entering the torus can be accurately measured. The operation of thecalorimeter without plasma gives the transmission efficiency rjT of the injectors(ions + neutrals) when the toroidal field is off, and the global neutral transmission
efficiency r)Q, expressed as TJG = TJT-IJN (where r;N is the species averaged neutralization efficiency) when the toroidal field is on.
The values of r)T and i7G are found to be constant and independent of theextracted power and to be close to the expected values [2]: G = 0.38, r;T = 0.60 andhence ijN = 0.63. This rules out re-ionization of the neutrals before they enter thetorus and any other conceivable saturation mechanism inside the injector withincreasing power.
1.2. Capture of fast neutrals by the plasma
The calorimeter can also be operated in the presence of plasma; its contribution
is then monitored by a second calorimeter [3]. Since the actual neutral beam powerat the plasma boundary is known, the measurement of the 'shine-through' fractionof this power can be used to check if the fast neutral ionization mechanisms are theexpected ones.
The experimental shine-through fractions are shown in Fig. 1 as a function ofline density. For identical plasma conditions the results are the same for the case with
two sources in operation and for the case with four sources in operation. This showsthat the power input to the plasma grows linearly with the extracted power.
Comparison with theoretical values computed on the basis of the usual capture
cross-sections [4, 5] shows that the experimental results can be explained by a small
increase ( + 13% at n = 8 X 1019 m" 3 , E = 34 keV) of these cross-sections. A
higher increase of the cross-section (+24%) due to multi-step ionization mechanisms
has been predicted theoretically [6, 7].
In any case, under TFR conditions, the power density in the plasma centre isalmost unaffected by these corrections and hence full accountability of the beam input
is assessed experimentally. Consequently, the observed low heating efficiency andsaturation are due to fast ions and/or bulk plasma confinement problems.

2.

BEHAVIOUR OF FAST IONS

Direct information on the beam ion distribution function and its dependence on
injected power has been obtained in the following experiment.

181

IAEA-CN-47/A-III-530

25

Usual cross-sections

-20
15

10

1.5

2.5

n ] (10"'m- 4 )FIG. I. Experimental shine-through values versus line density
and comparison with theory (solid lines).

2.1. Active charge exchange measurements
2.1.1. Measurements during the stationary phase
The distribution function of central fast ions has been measured using an
auxiliary modulated doping beam which crosses the line of sight of a multi-channel
neutral particle analyser at the plasma centre. Fourier analysis of the modulated
experimental signals at the modulation frequency gives an amplitude proportional to
the fast ion density at a given energy. A comparison of the signals at various power
levels, with the other experimental parameters (Te, ne, MHD, Yeff) kept constant,
has been made [8].
It is concluded that the fast ion density does not increase with power as
expected. Since saturation appears already at the highest analysed energy (30 keV,
only 3 keV below the injection energy) and then remains constant with energy, it follows that the saturating mechanism mainly affects the most energetic ions at the
beginning of their slowing-down process and transfers them towards peripheral
regions.
2.1.2. Measurements after beam cut-off
Using the same experimental set-up and holding the doping beam constantduring the fast simultaneous cut-off of the neutral beams, information on the slowingdown kinematics of the fast ions can be obtained. The temporal behaviour of thedecrease of the energetic neutral flux is compared with a simulation based on anumerical solution of the homogeneous time dependent Fokker-Planck equation. The
experiments are in good agreement with theory over the whole range of density and182

EQUIPE TFR

E=I8 keV

4

8

12

10

20
t (ms)

FIG. 2. Comparison between theoretical (solid lines) and experimental fast neutral signals
after beam cut-off for three different densities and three analyser energies.
A: ne(0) = 1.3 x 1020m-3,
D : ne(0) = 0.5 X W2Om~3,o / n/0) = 0.3 x JO20 m~3.

power (Fig. 2) and show that the fast ions which remain in the centre of the discharge
after the beginning of their slowing-down phase (from 33 keV to ~ 30 keV) continue
to be classically coupled to the plasma. This confirms that, independent of the exact
nature of the mechanism in connection with saturation, this mechanism necessarily
acts only at the highest fast ion energy.
2.2 Neutron production in D°-D experiments
On TFR, the neutron production in D°-D experiments is largely dominated bybeam-plasma interaction. These neutron fluxes can serve as a monitor of the highest
part of the fast ion distribution function (because of the energy dependence of thefusion cross-section) in the plasma centre (because of the peaked profiles of density
and slowing-down time). Since on TFR the ion sources can be tilted in a verticalplane, the neutron production dependence can be studied for various deposition profiles at different power levels [9] and under similar plasma conditions. The beams
can be aimed on axis (5 = 0 cm) as well as off axis (5 up to a/2 = 10 cm).
For operation with a single source the neutron yields and their dependence upon
6 are in excellent agreement with numerical simulations, but a comparison of datafor operation with one, two and four sources under the same plasma conditions(T , Zeff) shows that the increase of neutron yield with power is less than expected.
Moreover, the saturation is strongest for the most peaked deposition profiles (5=0)
and thus for the highest fast ion density (Fig. 3). The numerical simulation fits thehigh power cases only with artificially broadened fast ion profiles.
This confirms the spatial diffusion of the fast ions as indicated by active charge
exchange experiments.
2.3. A possible mechanism
For the very rapid loss of beam ions a mechanism is required which cantransport the ions on a time-scale much shorter than the slowing-down time. Such a

183

IAEA-CN-47/A-HI-5

'

1

*n

source** itourct'r••••••'r,>&

TheoryJ£ysi>=\§ cm

-

/ S ^
2

i

&=0 cm"Number of operated
sources1,
1
I

FIG. 3. Theoretical and experimental evolution of neutron yield with power for
central deposition (5=0) and more peripheral deposition (8 = 10 cm).

mechanism is the fishbone instability observed in high beta experiments with perpendicular NBI.
In our low beta case, an electrostatic trapped-ion instability can occur, driven
by resonance with the magnetic drift of beam ions [10]. The target plasma is in acollisional regime and responds adiabatically.
In the simplest approximation, the dispersion relation is:

3E

co—

where co = —is the precession frequency, 2 is the toroidal mode number, andr eBRq is the beam ion distribution. With TFR parameters, we find an t=\ mode
unstable for
AE_V
nP
where AE/EB is the relative width of f.
This mode is very similar to a bump-in-tail instability. Its low frequency andlarge wavelength scale render direct observation with the existing diagnostics
difficult. It is known that this type of resonance causes a rapid outward transport ofthe ions as they lose energy to the wave [11].

EQUIPE TFR

184

a Active CX
O Neutrons Number of operated sources
I
,
I
,
2

4

6

8

FIG. 4. Ion temperature increase (*) versus power and fast ion density
calibrated for operation with one source.

3.

POWER BALANCE WITH EXPERIMENTAL FAST ION LOSSES

From the above observations it follows that the central power density used incentral energy balance analysis has to be corrected. The actual power, as given bycharge exchange or neutron flux experiments, increases only by a factor of 2.8 goingfrom one to four sources. The saturation of the ion temperature increase is quite similar to the actual power saturation (Fig. 4), and, for central ion heating simulations
using the corrected power, no decrease of the Ohmic r is required any more.
However, since classical slowing-down of the remaining central fast ions has
been proved, it follows that, for TFR conditions, the power input to electrons andions is similar. A strong decrease of ve is needed to explain the poor electron heating. No specific mechanism for such a decrease has yet been identified, but a strong
correlation of this process with the density fluctuation level has been clearly
established experimentally [12].
In conclusion, the ion energy balance is dominated by a radial diffusion
mechanism of the injected fast ions which is non-linear with input power and affects
mainly the most energetic ions. Thus, bulk ion heating saturation does not imply
enhanced ion heat conduction. On the other hand, strong confinement deteriorationof the electrons has been shown.

REFERENCES[1]

EQUIPE TFR, in Plasma Physics and Controlled Nuclear Fusion Research 1984 (Proc. 10th Int.
Conf. London, 1984), Vol. 1, IAEA, Vienna (1985) 103.
[2] KIM, J., HASELTON, H.H., J. Appl. Phys. 50 6 (1979) 3802.

IAEA-CN-47/A-III-5
[3]
[4]
[5]
[6]
[7]
[8]
[9]
[10]
[11]
[12]

185

EQU1PE TFR, Power transmission and shine-through calorimetry during NBI in TFR, to be
published in Plasma Phys.
RIVIERE, A.C., Nucl. Fusion 11 (1971) 363.
OLSON, R.E., SALOP, A., Phys. Rev., A 16 (1977) 531.
PERRAUDIN, J . - C , PhD Thesis, Orleans Univ., 1977.
BOLEY, C D . , et al., Phys. Rev. Lett. 52 (1984) 534.
EQUIPE TFR, in Controlled Fusion and Plasma Physics (Proc. 12th Europ. Conf. Budapest,
1985), Vol. 2, European Physical Society (1985) 308.
EQUIPE TFR, in Controlled Fusion and Plasma Heating (Proc. 13th Europ. Conf. Schliersee,1986), Vol. 2, European Physical Society (1986) 285.
TAGGER, M., LAVAL, G., PELLAT, R., Nucl. Fusion 17 (1977) 109.
TAGGER, M., PELLAT, R., Plasma Phys. 24 (1982) 753.
TFR GROUP and TRUC, A., Nucl. Fusion 26 (1986) 1303.

DISCUSSION
R. J. GOLDSTON: I don't know what the injection angle was for these experiments, nor do I know how the toroidal rotation speed varied with beam power in
TFR, but don't you think that many of the effects you describe could perhaps be
explained by rotation effects?
J.-P. ROUBIN: Unfortunately, we have no measurements of the toroidal rotation, but in view of the quasi-perpendicular injection angle (15°) and the high toroidal
ripple value on TFR, these effects are probably not dominant.
M.C. ZARNSTORFF: Do you have any measurements of the fast ion slowingdown spectrum off axis during central heating to show the radial fast ion motion?
J.-P. ROUBIN: No, but we have measured greater ripple losses of fast ionsthan expected from the calculations and this may be an indication of an enhanced
radial diffusion of fast ions.

IAEA-CN-47/A-III-6

PARTICLE AND THERMAL TRANSPORT, ANDRESONANT FIELD EXPERIMENTS IN TEXT
A.J. WOOTTON, R.D. BENGTSON, J.A. BOEDO, R.V. BRAVENEC,
K.L. EMPSON, K.W. GENTLE, W.L. HODGE, C.C. KLEPPER,
J. PORTER, A. MACMAHON, S.C. McCOOL, W.H. MINER,
D.M. PATTERSON, P.E. PHILLIPS, B. RICHARDS, C.P. RITZ,
T.L. RHODES, D.W. ROSS, W.L. ROWAN, J.C. WILEY

Fusion Research Center,
The University of Texas at Austin,
Austin, Texas, United States of America
D.L. BROWER, S.K. KIM, N.C. LUHMANN, W.A. PEEBLES

University of California,
Los Angeles, California, United States of America
J.S. DeGRASSIE, T.E. EVANS, G.L. JACKSON, N. OHYABUGA Technologies, Inc.,San Diego, California, United States of America
P.H. DIAMOND, T.S. HAHM, P.W. TERRYInstitute for Fusion Studies,
The University of Texas at Austin,
Austin, Texas, United States of America
Shuping FAN, Yexi HE

Institute of Plasma Physics,
Hefei, China
J.C. FORSTER, R.L. HICKOK, P.M. SCHOCH

Rensselaer Polytechnic Institute,
Troy, New York, United States of America
G. HAAS, F. KARGER

Max-Planck-Institut fur Plasmaphysik,
Garching, Federal Republic of Germany
Xuehua YU, Shaobai ZHENG

Institute of Physics,
Chinese Academy of Sciences,
Beijing, China
R.C. ISLER

Fusion Energy Division,
Oak Ridge National Laboratory,
Oak Ridge, Tennessee, United States of America

187

188

WOOTTONetal.
Abstract

PARTICLE AND THERMAL TRANSPORT, AND RESONANT FIELD EXPERIMENTS IN TEXT.
Particle and energy confinement have been studied on TEXT. Characteristics of the turbulence
are presented, and scalings of global confinement times and local transport coefficients discussed. At
the edge the measured electrostatic component of the turbulence plays a major role in particle transport.
In the interior the measured density fluctuation levels are used in drift wave turbulence models to compare measured and theoretical transport coefficients. Resonant fields are used to produce ergodic layers
and coherent island chains. Ergodic layers enhance radial electron heat flux and reduce the edge temperature, while islands allow increased limiter particle collection.

PARTICLE TRANSPORT
Transport is studied on TEXT [l], an ohmically heated tokamak withmajor and minor radii R — lm, a = 0.27m, at plasma currents Ip < 400 kA,
toroidal fields Bt < 2.8T and densities ne < 1 x 1020n?"3. The working
particle flux F, deduced from flux surface averages (FSA) of the asymmetric
Ha or Da emission [2], is characterized by the particle confinement timer p = N/r(a) ~ 10ms. The Ip and Bt dependencies are approximately
rp <x \/q{a). rp peaks at ne ~ 4 x 1019m"3.
The turbulence-driven fluxes F* were measured using electrostatic
probes. The associated confinement time, T* = TV/F-^a), agrees with rpin both magnitude and functional dependence. Poloidal asymmetries of afactor 2 are observed, but do not affect the scalings. An exception to theagreement is that r^, unlike TP, shows little dependence on plasma position.
For large radial diffusion coefficient Dr (i.e., qzDr/(eD\\) > 0.01) this could
be explained by the interaction of a localized source (following plasma displacement) and a poloidally asymmetric Dr or inward convection velocity v.
The generally anisotropic edge turbulence does not satisfy the Boltzmann relationship.
For r < a, probes and an ion beam show
Je4>{u>)n/(h{u))kTe)<L; > 1, integrated from U;/2TT = 50 to 150kHz, thedominant frequency range of the fluxes. Changing the edge potential
4> by biasing limiters changes edge confinement [3]: a 30% decrease inboth r*(a) and V(a) (i.e., increasing T^ and rp) follows a decrease in Er(25 <. r < 28cm) from ~ SOUcm"1 to ~ -SOVScm"1.
Figure 1 shows the radial fluxes F (from Ha) and F ' (from probes) [4].
For r > a = 27cm there is agreement if a parallel loss term is used toaccount for flow to the limiter. However for r < 27cm, F^ is too large,
and continuity is apparently not satisfied. This can be resolved by invoking
probe perturbations , asymmetries, or an additional inward convection witht> ~ lO 2 ^^" 1 . An ergodic region could couple the limiter to r ~ 25cm, but

189

IAEA-CN-47/A-III-6

SourceProbes^Vn theory
VTJ theory

CM
I

E
O
CM

o

0.7

1.1

1.2

FIG. I. Profiles of the radial particle flux V for Ip = 200 kA, Bl = 2 T, he = 3.5 x 7 0 " m~3.
Both experimental (from Ha and probes) and theoretical ( V« and VTJJ cases are shown.

then absolute agreement between rp and T* would not be found. Also shown
are the fluxes expected from resistivity gradient driven [5] (V?), related to
the rippling mode) and collisional drift wave [6] (V77) turbulence. The VTJ
mode, for which line tying behind the limiter is a stabilizing effect, is theleast, unlikely candidate. It correctly predicts the scalings of F with Ip, Bt
and q, but not the scalings with n £ , the observed | i r | (see below) or theinequality e4>/(kT) > ii/n. The sensitivity to ET is under investigation.
In the interior,both impurity (z) and working particle (i) profiles andtransient responses can be characterized by F = — DTVn + nvor/a, withconstant Dr and inward v0, except for working ions near the density
limit. For / / + or D + , variations with Bt, Ip and ne are approximatedby the expressions [7] DT(i), vo{i) oc l/(iieq(a)):
typically Dr(H+) ~+2++
1.4I>r(D ) ~ 1 m .s~\ while vo(H ) ~ vo{D ) ~ lOm-s"1. For impurities (Sc,Ti,Fe) the global confinement r. oc Z^m^gB^~ 25ms, with
Dr(z) oc 7»i^0-5 - In? 2 *- 1 and vo(z) oc Z^rn^B;1~ lO?^" 1 . -i»0(;) is
consistent with neoclassical predictions if reasonable functional forms for
impurity concentration n.(r) and charge are assumed. As the operationalboundaries are approached,impurity accumulation is observed [8], a resultof increasing |i>0|.

ENERGY TRANSPORT
Behind the limiter the radial energy flux qT = Conduction + Convection
is deduced from power balance, with a parallel sink included to model
flow to the limiter. With ne = 2 x 10 19 nr 3 , Xe(a) ~ lOmh~\ andConvection/ QT = 0.15 [4]. If only electrostatic turbulence is to account for

190

WOOTTON et al.
FIR Scattered Power
Up/Down
Asymmetry- Limiter
Quasi-coherent
Feature-1.2-1.2-0.6

1 .2

FIG. 2. Contours over a poloidal cross-section of scattered power S12from FIR Thomson scattering,for Ip = 300 kA, 5, = 2.8 T,nt- 3.5 x 10>9 m~3. Incremental values are 0.1; high (H) and low (L)are distinguished. Results are normalized to 1.0 at x/a = 0, yla = 1, where n/n = 0.15.

qr, then Tt/Tt ~ 1 is required, an unreasonably high value. Magnetic fluctuation levels (b™*/Bt - 1CT4 between 20 and 500kHz) do not contributesignificantly: ,\£ ~ 0.03\ e . To allow additional q without increasing F (already explained by F'), either a stationary br or nonthermal parallel flow
could be invoked. As ne is increased, qCOnveci\on/<3T increases, but increasing
experimental uncertainties prohibit scaling.
In the interior a power balance shows \ e »; l/(neq(a)), ~ 2.5 to 5 x Dr.
\c from sawtooth propagation is ~ 3 times larger. The predictions of threedrift wave turbulence models [6,9] are compared with the experimental
results, using the spatially and spectrally resolved n values obtained from
Thomson scattering. Figure 2 shows normalized contours of the frequency
integrated scattered power with k ~ 12cin"a, 5 1 2 : n/n oc ^J'S(k)dk.
There is a broadband background spectrum with Au;/ai and Ak/k > 1,
which is up/down asymmetric [10] and sensitive to both Ip direction andlimiter configuration, suggesting a role is played by toroidal asymmetries.
n/n oc fi°B^: the Ip dependence is oc /^ in the interior and oc I~^ atthe edge. A quasi-coherent feature [11] with Au>/u> < 0.2 and Ak/k < 0.5
is found in the high field region, with n/n oc I°n°Bl. As ne is increased
from 2 to 8 x 1019nj~3 the ratio of drift contributions in the electron toion direction changes from 10/1 to 2/1: a clear peak appears in the ion
direction.
A comparison between \ t experiment and theory is shown in Fig. 3.
Theoretical results, shown only where applicable, are obtained either using the FSA values of n/n derived from the contour plots of 5 (Fig. 2)
or using theoretical estimates for the saturated levels. The FSA valuesof n/n are overestimated in the confinement zone because the data is not

IAEA-CN-47/A-III-6
E l e c t r o n Thermal
102

I

Diffusivity

Exp
D-T
Co I I i s i on Ies s

101

^

191

Col I i s i

I)

10-1
10 - 20

0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1
r/o

FIG. 3. Profiles ofxjor Ip = 300 kA, B, = 2.8 T, h~e = 3.5 x 1019 m~3. Results are shown from a
power balance analysis and from three theoretical models for drift wave turbulence (collisional,
collisionless and dissipative trapped electron, DT) using estimated (upper curves) and measured (lower
curves) values of the saturated nln.

deconvolved from the instrument function, and because the assumption ofa spatially independent, fc-distribut ion overestimates the importance of thequasi-coherent feature. Disagreement both in amplitude and spatial dependence is found. Using measured n/n increases the disagreement because
rc/n(~ 1% at r — 15cm, > rq=i) is less than the usxial estimate of saturation level (3p3/Ln ~ 15% at r = 15cm). The models also fail to explain themeasured Dr.
RESONANT FIELDS
Externally imposed resonant fields have been used []2] both to create
crgodic layers (proposed as a means of extracting heat and reducing impurities) and single island chains. Coils at r = 33cm with an m — 7,
7? = 3 or 2 structure are computed (vacuum fields only) to produce anergodic region for r > 22cm when the multiple sidebands produce overlapping magnetic islands: at r = 24cm a coil current //, ~ 5kA produces FSA(\br\) ~ 1.5 x 10~37\ Edge parameters become highly asymmetric: e.g. theprobe saturation current, <x n\/T, varies by a factor ~ 10 over A6 ~ 25°
behind the limiter. Plasma potential (ion beam, probe) is modified, withET increasing from -45V-cm~1 to —lSV'cm"1 for 0.6 < rja < 1. Figure
4 shows Tt(r) for Ih — 0 and 4fc/l, and one point for //, = ikA:Te(a) is
reduced without affecting Te(0). For 0.9 < r / a < l , \ t increases from 10 to
50 771I.S"1: this increase is larger than the maximum incremental value expected for a diffusive process, A\ = ((6,,/Bt)2) Xe|| = 2077?:'5~1, suggestingWOOTTON et al.

192

Edge Electron Temperoture Profile
with 7 / 3 mode EML

-

i

8

200 -

1

o T.S. I h= 0 kA
• T.S. I h = 4 kA
T.S. I h = 7 kA
A Probes 0 kA
A Probes 4 kA

•

•
•

150p-

o
#

-22.

<u 1 0 0

50

i

o

I

--

21

.

i

22

.

I

23

.

I

.

24

I
-

o
-

•

fc .
• i

u
•

I
;

t

25

i

26

27

\ A,A.A:
28

29

radius (cm)FIG. 4. Profiles of Te from Thomson scattering (T. S.) and Langmuir probes near the plasma edge fordifferent imposed resonant field coil currents /„. lp = 280 kA, Bt = 2 T, ne = 3 x 1019 m~3.

the intrinsic Xt l' a s ^so increased. Significant experimental uncertainties,
including poloidal asymmetries, complicate this comparison. No evidence
for an increased nt (a) concomitant with the reduced Tt(a) has been found,as predicted by simple momentum balance along field lines. The large intrinsic transport (£),. -»- lm^s" 1 ) may explain this.
Particle confinement is also affected by the resonant fields. T(o) from
Ho increases for condit ions without computed island overlap, but decreases
once computed overlap starts. Impurity line emission with r < 22cm isreduced, while with r > 26cm it is increased. Predictive transport code
calculations show these emissions characterize nz, the central reduction
being explained by a combination of a reduced source and an increased
outward flow in the ergodic layer.
Coherent island structures are proposed to enhance particle collection tolimiters. I'sing the m — 7, n = 2 resonance coils, coherent island structureswith width ~ 2cm are computed. A pressure gauge inside an apertured
limiter shows ~30% increase when the limiter is inserted to 20cm, inside
an island at the O-point. Thennography demonstrates diversion of theheat flux from the leading edge to the limiter stem, where it is more readily
dissipated.
CONCLUSIONSWorking particle confinement time scalings have been found. The values agree both in amplitude (± 50%) and scaling with those deduced from

IAEA-CN-47/A-III-6

193probe measurements of electrostatic turbulence, suggesting a causal relationship. The radial profiles of the fluxes agree at and behind the limitertip. Resistivity gradient driven turbulence is the least unlikely candidate
investigated to date.
In the main plasma the working and impurity ion histories can usually
be explained using a constant Dr and inward !>0 model. For the working
gas, both coefficients oc l/(nlq{a)).
Impurity accumulation at low Bt is
explained by an increasing VQ.
Behind the limiter, at low density, energy transport cannot be explained
solely by convection. Magnetic fluctuations contribute insignificantly to the
conduction.
Within the main plasma,,\ e deduced from power balance oc 1 /(h e q(a)) ~2.5 — 5 x D r . Using spatially and spectrally resolved values of rj/ra, drift wave
turbulence models are shown to underestimate \ e and predict incorrect
scalings. At high densities an ion feature has been identified.
Resonant coils have been used to produce both ergodic edge layers andcoherent islands. Where island overlap is expected, Te(a) is reduced from100 to 30eVand interior impurity levels are reduced. Coherent island structures have been used to enhance pressure collection within an aperture
limiter.

ACKNOWLEDGMENTS
This work was supported by the U.S. Department of Energy contracts
DE-AC05-78ET-53043, DE-AS05-85ER-53193 and DE-AC03-84ER-53158.
REFERENCES
[1]
[2]
[3]
[4]
[5]
[6]
[7]
[8]
[9]
[10]
[11]
[12]

GENTLE, K., Nucl. Technol./Fusion 1 (1981) 479.
ROWAN, W.L., et al., Rep. No. 281, Fusion Research Center, Texas Univ., Austin (1986),
to be published in Nucl. Fusion.
PHILLIPS, P., et al., Rep. No. 291, Fusion Research Center, Texas Univ., Austin (1986).
RITZ, C , etal., ibid.
HAHM, T.S., et al., Rep. No. 180, Institute for Fusion Studies, Texas Univ., Austin (1985),
submitted to Phys. Fluids.
TERRY, P.W., et al., Phys. Fluids 28 (1985) 1419.
GENTLE, K.G., et al., Europhysics Conf. Abstr. IOC Pt 2 (1986) 244.
ISLER, R.C., et al., Phys. Rev. Lett. 55 (1985) 2413.
PERKINS, F.W., in Heating in Toroidal Plasmas (Proc. 4th Int. Symp. Rome, 1984), Vol. 2,
Int. School of Plasma Physics, Varenna (1984) 977.
BROWER, D.L., et al., Phys. Rev. Lett. 54 (1985) 689.
BROWER, D.L., et al., Phys. Rev. Lett. 55 (1985) 2579.
OHYABU, N., et al., Nucl. Fusion 25 (1986) 1684.

194

WOOTTONetal.
DISCUSSION

A. ROGISTER: I would like to make two points: First, the theory of therippling instability in toroidal geometry predicts an important up-down asymmetryof the spectrum of the linear density eigenfunction1 consistent with that measured
on TEXT. Theoretically, this asymmetry is a consequence of the magnetic drifts and,incidentally, there might be a connection here with the observation reported in
Dr. Wagner's paper2 on the asymmetric threshold for the transition to the H-mode
in ASDEX.
Second, you mentioned that the electron heat conduction coefficient calculated
from the measured density fluctuation (6n/n) is too small to account for the observed
transport. A similar conclusion was reached in TFR turbulence studies. However,xe is not simply proportional to (5n/n)2, but involves an integral over the poloidal
spectrum, which includes a stronger weighting factor for large kg than the integral
appearing in (6n)2. In the case of TFR, I found agreement between theory andexperiment when the proper calculation was performed.
H.P. FURTH: Towards the edge of your plasmas, you seem to find that various
theoretical transport mechanisms are too weak to account for the observed transport
rates. To be self-consistent, you should really modify transport theory to include theeffect of ionization and charge exchange on the modes that are being considered as
candidates to explain the transport. It is easy to see that in a recycling plasma theseeffects cannot be neglected unless k > d log n/dr. Have you looked into thisquestion?
A.J. WOOTTON: We have not yet applied any formulae for the edge particle
or heat fluxes from models which include these considerations.
B. COPPI: Do you have any report to make on the particle diffusion and inflow
at the centre of the plasma column?
A.J. WOOTTON: Yes. We have used an oscillating gas puff, with the
FIR interferometer system, to determine D and v0, where r = - D v n + nvr/a.
Assuming a constant D and v0, we find that both D and v0 scale approximately as
l / ^ q , ) , with typical values of D = 1 m2-s"' and v0 - 10 ms"1. However, there
are cases where the fitting of the results to the expression is poor, suggesting it isincorrect to assume that these coefficients are constant.
F. WAGNER: There may be experimental evidence that the transition from the
Ohmic linear regime to the Ohmic saturation regime depends on the edge plasma
parameters. Is the transitional density range affected by the ergodic limiter and are
there characteristic changes in the Langmuir probe fluctuation signals at thetransition?
A.J. WOOTTON: We have not made a careful study of this transition with theapplied resonant fields. Concerning the fluctuations at the edge, we have a problem.
Initial results from probes suggested no change in edge electrostatic turbulence withthe resonant fields, as long as the edge density was held constant in the comparison.
However, the discovery of large poloidal asymmetries in ion saturation current —IAEA-CN-47/A-III-6

195

only with applied resonant fields — means we must make the measurements as a function of poloidal angle before we can answer this question.
D.C. ROBINSON: What fraction of the thermal transport is accounted for byyour measured electrostatic turbulent convection, and to what extent do other transport processes, such as magnetic field fluctuations, account for the edge thermaltransport?
A.J. WOOTTON: We have only made this determination at lower densities(ne < 3 x 1019 m~3) because of the large uncertainties in the power balance estimation of the total energy transported at the edge. Here we find that perhaps 20%of the total heat flux is accounted for by convection. Behind the limiter the valuesof both T and b, in the frequency range up to 250 kHz, are too small to explain theremaining 80%. The only possible candidates would appear to be a stochastic magnetic field region, or non-concentric potential and density surfaces. The latter, if
present, would negate our agreement between total flux and turbulence driven flux.
Behind the limiter a suprathermal electron component might transport more parallel
heat than we have estimated. There is no evidence for such a component.

1

ROGISTER, A . , HASSELBERG, G., Paper IAEA-CN-47/E-IIM, these Proceedings,
Vol. 2.
2
NIEDERMEYER, H., WAGNER, F., et al., Paper IAEA-CN-47/A-II-5, these Proceedings,
Vol. 1.

IMPURITY PRODUCTIONMECHANISMS AND BEHAVIOUR DURING
ADDITIONAL HEATING IN JET
K. BEHRINGER, A. BOILEAU1, F. BOMBARDA2, G.B. DENNE,
W. ENGELHARDT, M.J. FORREST3, G. FUSSMANN4,
R. GIANNELLA2, N.A. GOTTARDI, M. VON HELLERMANN,
L. HORTON, H. JACKEL4, C. JUPEN5, E. KALLNE,
K.D. LAWSON3, G. MAGYAR, G.M. McCRACKEN3,
P.D. MORGAN, E.R. MULLER4, N.J. PEACOCK3, J. RAMETTE6,
B. SAOUTIC6, M.F. STAMP, H.P. SUMMERS7,
G. TALLENTS, A. WELLER4
JET Joint Undertaking,
Abingdon, Oxfordshire,
United Kingdom
Abstract
IMPURITY PRODUCTION MECHANISMS AND BEHAVIOUR DURING ADDITIONALHEATING IN JET.
Impurities in JET are investigated by spectroscopic and plasma edge diagnostics. Impurity
influxes are measured by visible spectroscopy, allowing identification of impurity sources and conclusions on production mechanisms. Metals and carbon in JET plasmas originate mainly from the limiters;
physical sputtering is the release mechanism. Oxygen originates from the walls, presumably by photonor neutral particle desorption. During ICRH, antenna screen material is released by a process not yet
understood. Transport studies yield D *= 1 m2/s for impurities. Ohmic and neutral beam heatedplasmas are dominated by light impurities (C, O), in particular by oxygen at high ne. Moderate metallevels (0.1% n^ during ICRH are responsible for radiation losses of = 20% P tot . These can beprevented temporarily by carbonization. Radiative power losses are 30-60% P (ot for NBI and
40-70% P tot for ICRH. Z eff ranges between 3.5 at low n e and 2.5 at higher n e . During magnetic
separatrix operation, a reduction of metals and carbon is observed. The bulk plasma radiation drops
compared to limiter discharges, but there is an additional contribution from the X-point regions. To date,no accumulation has been observed during NB counter-injection.

1
2
3

4

5
6
7

Institut national de la recherche scientifique, Quebec, Canada.
Euratom-ENEA Association, Centro Ricerche Energia, Frascati, Italy.
Euratom-UKAEA Association, Culham Laboratory, Abingdon, Oxfordshire,
United Kingdom.
Euratom-IPP Association, Max-Planck-Institut fur Plasmaphysik, Garching,Fed. Rep. Germany.
University of Lund, Lund, Sweden.
Euratom-CEA Association, Centre d'&udes nucleaires de Fontenay-aux-Roses, France.
University of Strathclyde, Glasgow, United Kingdom.

197

198

BEHRINGER et al.

1.

INTRODUCTION
Impurities in JET plasmas are studied by a variety ofspectroscopic diagnostics, from the visible to the X-ray range[1], by bolometers and soft X-ray cameras, and by charge-exchange
recombination spectroscopy (CXRS). Z e f f is derived from visible
bremsstrahlung. Hydrogen and impurity influxes, <j>uD and <|>p aremeasured at various locations in the torus by visible spectroscopy. For this purpose, line excitation rate coefficients andintensities of specific impurity lines have been computed
relating the number of emitted photons to the influx rate [2].
Surface analysis of collector probes, long-term samples andlimiter tiles provides information on the status of surfaces incontact with the plasma and on the consequences of cleaning
procedures and carbonisation. Langmuir probes are used tomeasure the plasma edge parameters. The limiter surface
temperatures are monitored by infra-red cameras [3L
2.

IMPURITY TRANSPORT AND IMPURITY CONFINEMENT
Emission shells of light impurities and low metal ionisation
stages in JET are clearly transport dominated, requiring a
diffusion coefficient D = 1 m2/s in order to explain theirposition and width [4], Information about transport was also
obtained from accidental injections of iron and nickel. At
ne=2-1019m~3, simulations using D=1 mz/s agree best with theobserved rise and decay of nickel line intensities, as shownin Fig.1 for Ni XXVI. There are indications that D decreases
with electron density and increases during additional heating.
Investigations of hydrogen particle transport [5] result insimilar transport coefficients as found for the impurities.
D=1 m V s and a drift velocity VD=-2Dr/a2 are used instandard interpretation of VUV and X-ray spectra and for
calculating radiated power losses of individual impurity species
by means of an impurity transport code [6]. The resulting,moderately peaked total ion density profiles are confirmed withina factor of two by analysis of a number of metal ionisation
stages at different radial positions (eg Ni XVII-Ni XXVII).
Sawteeth, observed on lines of higher ionisation stages, can only
be partly explained by pertinent variation of electron
temperature profiles. It must be assumed that impurities are
expelled from within the q=1 surface during the sawtooth crash.
Both the radiated power Pp ad and Zeff-1 are roughly
proportional to the square of the nuclear charge Z and theconcentration of the impurity species. Therefore, normalized,
average radiation losses per unit surface area have been defined(neglecting hydrogen radiation) where A is the plasma surface
area. P n o r m has been calculated by the impurity transport code
for a variety of JET plasma conditions and for the important JET199

IAEA-CN-47/A-IV-l
1.0

ft

\

0.8 -

\

\

#3610

6.0 s

D=0.5m2/s\

Ni XXVI 165.3 A

\
s
\
0.6 -

1
1

:
'•

\
N
N

1

0.4 -

\

•

N

o experiment

0.2 -

/

\ ^ — D=1 m2/s

A_... 0 ^ ^ - Q ^D =2m2/s
0.0

N

\

;j
,

-

\

\

i

1

0.0

0.5

•

i

i

i

i

t

i

1.0

O
i

1.5

,

,

i

,

Q

i

2.0

Time(s)FIG. 1. Ni XXVI intensity after accidental nickel injection (ne = 2 x 1019 nr3, Ip = 2.8 MA).
Simulations using different D-values are shown (VD — —2Dr/a2). Recycling at the limiters leads to a
finite level after the event.

= P nld /An. 2 (Z.«-1)(10- 3 «kW-m' 1 )0.0
23n , (1019/m3)

FIG. 2. Normalised power per unit surface area for 1986 JET limiter plasmas. The ranges of Pnorm
predicted by the transport code (D = 1 m2/s) for O, C and metals are indicated. Corona assumption
leads to much lower Pnorm values for O and C. One point each is shown for ASDEX limiter and divertor
discharges [7].

200

BEHRINGER et al.

impurities. P n o r m is found to be about 0.4-0.9*10~38 kW-m" forlight impurities and about 0.9-1.2.10"38 kW-m1* for metals. The
lower end of the range corresponds to peaked n e profiles and highT e values, the upper end to flat n g profiles and low T e values.
The assumption of corona ionisation equilibrium reduces P n o r m bya factor of 5-10 for light impurities, because the radiation
shell is then located in the scrape-off layer at low electron andimpurity ion densities. A summary of experimental PnOrm v a l u e s
as a function of n is shown in Fig.2 for JET 1986 limiter
discharges (500 data points). The vast majority of measured
values is found in the predicted range. Fig.2 proves clearly thenecessity of including transport in the plasma simulation. Two
ASDEX data points [7] demonstrate the validity of the scaling
with surface area. The detailed behaviour of Pnom with respect
to the different heating schemes will be discussed below.
Impurity particle confinement times, derived from measured
influxes and concentrations, agree roughly with calculations on
the basis of a simple diffusion model [2] and D=1 m 2 /s, if anenergy of a few eV is assumed for the neutral particles produced
at the limiters. A good correlation of impurity influxes anddensities has been observed throughout the JET operation, both asa function of plasma parameters and of vessel conditions.
Occasionally, when a steep increase in the Cr influx was
observed,due to chromium evaporation, the metal concentration in
the bulk plasma did not change [2]. These slow particles areionised in the immediate vicinity of the limiter surfaces andhave a high redeposition probability before entering the plasma.
This observation demonstrates that the usual impurity influxes ofC, 0 and metals must have a significant kinetic energy to giverise to the measured impurity concentrations.
3.

IMPURITY PRODUCTION
The impurity behaviour of JET limiter discharges was
dominated by the limiter fluxes (wall contribution = 20$ duringOH). Metal concentrations were only significant (> 0.1? n e ) if
the carbon limiters were metal-coated after accidental melting
and evaporation of wall material (Inconel droplets and £2«102lm~2
uniform coverage [8]). Recently, this has been prevented bycovering large areas of the torus by carbon protection plates. A
particular metal problem during ICRH will be discussed later.
The release of metals from the limiters can be explained by
a combination of sputtering by deuterium and by light impurities,
taking into account the respective surface coverage of the carbon
tiles (<f>M/<f>n = °"°-02)« The metal fluxes decrease with n e andincrease with I_; they are inversely correlated with the light
impurity behaviour. These trends, observed before on many othertokamaks, are due to the temperature dependence of the metal
sputtering yields. In JET, gettering of oxygen by chromium has
been observed to be a further reason for this anti-correlation.

IAEA-CN-47/A-IV-l

201

The carbon concentration in JET has been consistently at alevel of a few % n e throughout operation periods. The carbon
production rate behaves similarly to the metal release, but thedependencies are less pronounced [2]. This observation and thefact that the produced C-particles must have significant kinetic
energy suggest physical sputtering as the prevailing production
mechanism. Furthermore, during temperature excursions of thelimiter surface from 300°C to over 1000°C, no change in carbon
production rate was observed, which means that chemical
sputtering, expected to manifest itself by an increased yield at
about 600°C, is not an important process. There is some evidence
for carbon sputtering by oxygen, but only at unusually highoxygen levels [2]. The carbon yields measured (<J>C/())Q= 0.05-0.1)
can be explained by a combination of deuterium, oxygen andself-sputtering at high T ( a ) values (>100 eV).
The oxygen fluxes ana concentrations in JET varied primarily
with vessel conditions. The vessel walls constitute the mostimportant oxygen source and, once in the plasma, oxygen recycles
at the limiters. This is apparent from the tendency of oxygen
limiter fluxes to increase throughout a discharge, even if otherparameters are stationary. If the recycling occurs in the formof CO molecules, the 0 atoms must gain energy in Franck-Condon or
charge-exchange processes. Since no dependence of the oxygen
yield on plasma parameters is observed, the energy threshold forthe respective production mechanism must be very low. During 3He
discharges, carbon fluxes remained unchanged (<J>c/ij>He=0.13) while
oxygen was substantially reduced. Therefore, chemical processes
or CX neutrals must be responsible for the 0 production.
Due to the above trends, JET plasmas were metal dominated at
low electron densities, if there was significant metal depositionon the limiters, otherwise carbon was the most important
impurity. At high electron densities, metals disappeared, carbon
was reduced, and oxygen dominated. Oxygen concentrations
increased, particularly close to the density limit,_due to enhanced
plasma-wall interaction. Z e f f fell steeply with n e when metals
were important and was a weak function of n $ for metal-freeplasmas. During discharges limited by the inner-wall protection
tiles, these took over the role of the limiters leading
essentially to the same plasma behaviour.
During ICRF heating on JET, screen material (Cr, Ni) has
been observed to enter the plasma from the ICRH antennae [2,3].
The measured chromium flux from a Cr coated antenna(1016 cm~2-s~l/MW) decreased by a factor 5 during five months'
operation and was reduced temporarily (some 10 discharges) by twoorders of magnitude by means of heavy carbonisation (15$ CH^,
12 hrs).
In the course of ICRH operation, screen material was
deposited on the limiters and eroded again by the plasma, resulting in higher basic metal levels during an ICRH campaign. The
metal coverage of the limiter carbon tiles was removed by a few
OH or NBI discharges, in accord with model predictions [9]. The
release mechanism of the screen material is not yet understood.

BEHRINGER et al.

202

§
S

o
C\J ICNo < a * x

±Jo

UL

? °+ OO_

CO CO CO LLXz z z ec O
O < D *

X

00

^1
1

CV1

o

IAEA-CN-47/A-IV-l

203

4.

ICRH AND NB CO-INJECTION
Both hydrogen and impurity influxes increased during
additional heating, leading to a substantial increase in fie>unless the plasma was at the inner wall, which appears to have an
efficient pumping capability. Although the wall fluxes gained inimportance, the plasmas were still dominated by the limiters.
The ne-profile flattened during ICRH; it was unchanged or peaked
during NBI. The hydrogen and impurity particle confinement times
decreased in both cases.
Due to the higher electron densities during NBI, oxygen was
the most important impurity. The increase in the 0/C ratio iswell demonstrated by the CXRS results [10], the trends of which
are in excellent agreement with the usual VUV spectroscopic
results, while they tend to give somewhat higher carbon
concentrations. Zeff dropped during NBI, but not to a lower
level than that of OH discharges at the respective higher
densities. For constant n e , the radiated power fractionp
rad/ptot f e l 1 w i t n increasing NBI power. However, due to thepertinent ne~evolution, radiation losses remained around H5%
P tot , on average. These radiation losses were caused by light
impurities, in particular by oxygen (1-255 n § ). Due to thelatter fact and a tendency to peaked ne-profiles during NBI,
P nom -values are at the lower end of the range in Fig.2.
The metal release from the antenna screens during ICRH led
to significant metal densities in the plasma ( = 0.1$ n e , contribution to P r a d = 20$ PfO£» see also [11]) except immediately
after heavy carbonisation. Z e f f remained essentially at the OH
value before RF, La there was no benefit of the higher electron
densities. For individual campaigns, the radiated power fraction
increased somewhat with Ppp and was generally in the range 40-70$
Ptof Hydrogen minority heating led to higher metal densities in
the plasma than 3He minority heating, but the difference in P r a d
is within the scatter of the data points. Due to some metal
contribution and flatter ne-profiles, the P n o r m values for ICRH
in Fig.2 are at the upper end of the predicted range.
During combined NBI and ICRH, the consequences of the twoheating schemes essentially added. The metal densities in theplasma corresponded to the respective RF power. Basic metallevels and limiter coverage built up as during ICRH alone. Z e f f
and Pro(j/Ptot v a l u e s f o r 0 H plasmas and additional heating aresummarised in Fig.3 for 1986 JET limiter discharges. At higher
electron densities, Z e f f was about 2.5 for NBI, ICRH and combined
heating, essentially caused by oxygen and carbon.
5. MAGNETIC SEPARATRIXThe magnetic separatrix configuration has been successfully
generated in JET (X-point plasmas [12]). In this case, the top
and bottom vessel protection tiles act as X-point neutraliser
plates. Even when these plates consisted of Inconel (June 1985),

204

BEHRINGER et al.

no significant metal concentrations were detected in X-point
plasmas. With carbon plates, the carbon level was reduced
compared to the respective limiter discharges during the OH
phase. During NBI, carbon concentrations increased, in contrast
to the behaviour of limiter discharges as demonstrated again by
CXRS [10]. The behaviour of carbon and metals indicates a low
plasma temperature in front of the neutraliser plates during OH
discharges and therefore a low sputtering yield. After argon
puffing into X-point plasmas, no retention of Ar in the X-point
regions was observed, ie the same amount was found in the bulk of
X-point and corresponding limiter discharges, in contrast to the
Doublet III results [13].
The oxygen concentration was not different from limiter
plasmas (= 1.5$ n e ) . Since metal concentrations were generally
low, X-point operation resulted only in a minor reduction of Zeff
(Z e f f * 3 at n e = 1.5«1019 m"3 and I = 2 MA). The radiation
losses of the Bulk plasma were also somewhat lower, but there wasan additional contribution of the X-point regions of = 50$ Phulk*
The total radiation losses amounted to = 50$ P Q at low densities(n = 1«1019 m~3) and approached 100$ at higher densities. NBI
into X-point plasmas led to a low level of wall material in theplasma, probably due to CX particles hitting the walls. The
oxygen concentration did not change significantly during NBI.
6.

NB COUNTER-INJECTIONCounter-injection,both into limiter and X-point plasmas, hasbeen studied for a limited period in JET. The impurity behaviourof these discharges was virtually indistinguishable from thecorresponding co-injection cases (see Figs 2 and 3). Somepeaking of soft X-ray emission profiles was observed and the line
intensities of higher metal ionisation stages appeared toincrease more than the lower ones. However, the standard
analysis taking into account the respective n g - andTe-distributions resulted in the usual impurity ion density
profiles within the experimental error bars. For the present NBI
power levels (_< 7 MW) and plasma parameters, no accumulation of
impurities has been observed.
7.

CONCLUSION
The JET impurity situation is not significantly different
from that of other ungettered tokamaks with carbon limiters.
Metal concentrations in the plasma were very low during OH and
NBI discharges,due to coverage of large areas of the wall by
carbon tiles. An exception was ICRF heating, where screenmaterial entered the plasma from the antennae. This will be
prevented in the future by a carbon or beryllium coating of the
Faraday screens presently under investigation. The carbon
concentration was high at moderate electron densities, but, due
to the fact that physical sputtering is the production mechanism,
it decreased at high densities, which are aimed for anyway. At

IAEA-CN-47/A-IV-l

205high fL, radiation losses weremainly caused by oxygen. Wellbelow the density limit, P ra d /p tot was tyP i c a l l y 50$ radiatedfrom the plasma edge, while the bulk plasma was transport
dominated. A primary aim of future operation must be thereduction of oxygen levels, which could be most.easily achieved
by gettering. Accumulation of impurities has not been observed
in JET.
REFERENCES[1] BEHRINGER, K., Rev. Sci. Instrum. 57 (1986) 2000.
[2] BEHRINGER, K., J. Nucl. Mater. 145-147 (1987) 145.
[3] ENGELHARDT, W., Plasma Phys. Controll. Fusion 28 (1986) 1401.
[4] BEHRINGER, K., et al., in Basic and Advanced Fusion Plasma Diagnostic Techniques (Proc.
Workshop Varenna, 1986), CEC, Brussels (1987).
[5] CHEETHAM, A., et al., in Controlled Fusion and Plasma Heating (Proc. 13th Europ. Conf.
Schliersee, 1986), Vol. 10C, Part I, European Physical Society (1986).
[6] BEHRINGER, K.H., CAROLAN, P.G., DENNE, B., et al., Nucl. Fusion 26 (1986) 751.
[7] MULLER, E.R., Nucl. Fusion 22 (1982) 1651.
[8] BEHRISCH, R., et al., J. Nucl. Mater. 145-147 (1987) 731.
[9] McCRACKEN, G.M., et al., J. Nucl. Mater. 145-147 (1987) 621.
[10] VON HELLERMANN, M.G., et al., in Basic and Advanced Fusion Plasma Diagnostic Techniques (Proc. Workshop Varenna, 1986), CEC, Brussels (1987).
[11] JACQUINOT, J., et al., paper IAEA-CN-47/F-I-1, these Proceedings, Vol. 1.
[12] TANGA, A., et al., Paper IAEA-CN-47/A-I-1, these Proceedings, Vol. 1.
[13] MAHDAVI, M.A., et al., Phys. Rev. Lett. 47 (1981) 1602.

DISCUSSION
S. KONOSHIMA: Can you describe the entire experimental impurity profile
from the centre up to the edge by a single value of the transport coefficientD ~ 1 m2/s? And is there a significant difference in these values between the
H-mode and the L-mode, especially near the edge?
K. BEHRINGER: Within the error bars, the impurity density profiles in JET
can be described by a constant diffusion coefficient D = 1 m2/s — the value from
impurity injection — and an inward drift velocity vD ~ -2Dr/a 2 . We have no proof
that other models — for example with both D and vD varying with radius — could
not provide an equally good description.
There are indications that D increases during the L-mode. During the H-mode,particle confinement is clearly improved, but there are no systematic studies yet withrespect to the individual transport parameters.
F. WAELBROECK: You attribute the appearance of carbon impurity ions tophysical sputtering. I would therefore like to ask why there is no chemical erosion
in JET, i.e. metal-carbon bonds at the limiter surface. Furthermore, other devices,for example TEXTOR, show molecular band emission near the limiter. What
happens in the case of JET?

206

BEHRINGER et al.

K. BEHRINGER: Chemical sputtering cannot be excluded completely. If it
does take place, the respective particles must be redeposited before being ionized andare thus not detected by emission spectroscopy. The important flux entering theplasma does not originate from chemical effects.
Molecular band emission near the limiter is observed in JET, but the respective
intensities are very low.
R.J. TAYLOR: Is it possible to fabricate a limiter for JET on which norecycling of hydrogen or impurities would take place?
K. BEHRINGER: We do not know of any such limiter material, but we would
be grateful for any proposals.

IAEA-CN-47/A-IV-2

IMPURITY AND PARTICLE TRANSPORTAND CONTROL IN TFTR
K.W. HILL, V. ARUNASALAM, M.G. BELL, M. BITTER, W.R. BLANCHARD,
N.L. BRETZ, R. BUDNY, C.E. BUSH1, S.A. COHEN, S.K. COMBS1,
S.L. DAVIS,
D.L. DIMOCK,
H.F. DYLLA,
P.C. EFTHIMION,
L.C. EMERSON1,
A.C. ENGLAND1,
H.P. EUBANK,
R.J. FONCK,E. FREDRICKSON, H.P. FURTH, G. GAMMEL, R.J. GOLDSTON, B. GREK,
L.R. GRISHAM, G. HAMMETT, R.J. HAWRYLUK, W.W. HEIDBRINK2,
H.W. HENDEL 3 , E. HINNOV, S. H1ROE1,
H. HSUAN, R.A. HULSE,
K.P. JAEHNIG, D. JASSBY, F.C. JOBES, D.W. JOHNSON, L.C. JOHNSON,
R. KAITA,
R. KAMPERSCHROER,
S.M. KAYE,
S.J. KILPATRICK,
R.J. KNIZE, H. KUGEL, P.H. LaMARCHE, B. LeBLANC, C.H. MA1,
D.M. MANOS,
D.K. MANSFIELD, R.T. McCANN,
M.P. McCARTHY,
D.C. McCUNE, K. McGUIRE, D.H. McNEILL, D.M. MEADE, S.S. MEDLEY,
D.R. MIKKELSEN,
S.L. MILORA1,
W. MORRIS4,
D. MUELLER,
5
6
V. MUKHOVATOV , E.B. NIESCHMIDT , J. O'ROURKE7, D.K. OWENS,
H. PARK, N. POMPHREY, B. PRICHARD, A.T. RAMSEY, M.H. REDI,
A.L. ROQUEMORE, P.H. RUTHERFORD, N.R. SAUTHOFF, G. SCHILLING,
J. SCHIVELL, G.L. SCHMIDT, S.D. SCOTT, S. SESNIC, M. SHIMADA8,J.C. SINNIS, F.J. STAUFFER9, J. STRACHAN, B.C. STRATTON, G.D. TAIT,
G. TAYLOR,
J.R. TIMBERLAKE,
H.H. TOWNER,
M. ULRICKSON,
V. VERSHKOV5,
S. VONGOELER,
F. WAGNER10,
R. WIELAND,
J.B. WILGEN1,
M. WILLIAMS,
K.L. WONG,
S. YOSHIKAWA,
R. YOSHINO8, K.M. YOUNG, M.C. ZARNSTORFF, V.S. ZAVERYAEV5,
S.J. ZWEBEN

Princeton Plasma Physics Laboratory,
Princeton University,
Princeton, New Jersey,
United States of America

1

Oak Ridge National Laboratory, Oak Ridge, TN, USA.
GA Technologies Inc., San Diego, CA, USA.
3
RCA David Sarnoff Research Centre, Princeton, NJ, USA.
4
Balliol College, University of Oxford, Oxford, UK.
5
Kurchatov Institute of Atomic Energy, Moscow, USSR.
6
EG&G Idaho, Inc., Idaho Falls, ID, USA.
7
JET Joint Undertaking, Abingdon, Oxfordshire, UK.
8
JAERI, Naka-machi, Naka-gun, Ibaraki-ken, Japan.
9
University of Maryland, College Park, MD, USA.
10
Max-Planck-Institut fur Plasmaphysik, Garching, Fed. Rep. Germany.
2

207

208

HILL et al.
Abstract

IMPURITY AND PARTICLE TRANSPORT AND CONTROL IN TFTR.
Degassing of the TFTR graphite limiter by low density deuterium or helium discharges enables
the limiter to pump deuterium, thereby reducing recycling and improving energy confinement in neutralbeam-heated discharges. During a helium degassing sequence the hydrogen influx decreased by a factorof 20. As a consequence of degassing sequences the low density limit in 0.8 MA deuterium discharges
decreased from 1 x 1019 m"3 to 0.5 x 1019 irr 3 , the density decay time dropped from greater than
10 s to 0.15 s, and the recycling coefficient dropped from nearly 1 to less than 0.4. Z eff values in
2.2 MA L-mode discharges on the toroidal limiter with neutral-beam-heating power up to 15 MW are
between 2 and 3 if the pre-beam plasma has low Z eff (high density), but can be as high as 4.5 if the
pre-beam plasma has high Z eff (low density). Z eff values in enhanced confinement shots drop from 7
during the ohmic phase to 3 with neutral beam heating. The radiated power drops from 60-70% of total
heating power to 30-35% for beam powers from 10 to 20 MW.

I.

Introduction

Operation of TFTR plasmas on the toroidal graphite limiterat neutral beam powers up to 20 MW has provided new data on
plasma-limiter interactions at near reactor temperatures anddensities.
Discovery of a new regime of enhanced energy
confinement in beam-heated discharges [ 1 ] has improved prospects
for achieving breakeven in TFTR and has provided new insights into
the role of edge particle control. Degassing of the limiter toachieve low recycling and documentation of the effect of thedegassing on plasma confinement have been a focus of recent
experiments [2,3].
A second aspect of plasma-limiter/wall interactions isimpurity production and influence of the resulting contaminationon the plasma. Plasma impurity levels in L-mode and enhanced
confinement discharges with neutral beam powers up to 20 MW are
discussed.
Impurity transport rates have been inferred byinjecting impurities into the plasma and studying the time
behavior of their radiation.
II. Conditioning for Particle and Impurity Control
The large area toroidal limiter [2] required special
conditioning techniques to achieve satisfactory plasma operation[4]. Following installation, 130 hours of glow-discharge cleaning(GDC) and 175 hours of pulse-discharge cleaning (PDC) were
performed during a six-week period, with the vacuum vessel heated
to 150 C.
After this conditioning, discharges were severely
affected by outgassing of the limiter following a disruption.
Additional PDC and a newly developed technique, disruptive
discharge cleaning O(DDC), were used to heat the limiter surface
sufficiently (-1000 C) to effect outgassing of water vapor andhydrocarbons.
The DDC consisted of a sequence of tokamak
discharges with flat-top currents increasing from 0.6 to 2.5 MA.

209

IAEA-CN-47/A-IV-2

The flat-top current was increased progressively by 0.2 MA after aforced disruption at each current did not affect the succeeding
discharge.
When recovery was easy after a 2.5 MA disruption,
ohmic discharges at 2.2 MA had a radiatedjDower fraction less than
50? and Z f ~ < 1.5 for n g > 4 x 10^' m~3\ Following a one-week
opening or the vacuum vessel, DDC, in addition to GDC and PDC, wasagain required to eliminate excessive limiter outgassing caused by
a 2.2 MA disruption.
In contrast, only standard GDC and PDC techniques wererequired to prepare both the vacuum vessel and the movable
limiter for high power operation, prior to installation of thetoroidal graphite limiter [5]. Following 100-200 discharges, theradiated power fraction and Z f f in ohmic fiducial discharges (I= 1.4 MA, n e = 2.4 x 10 1 9 m 3 ) dropped to 60 - 70% and 2"
respectively.
The
initialenhancedconfinement
neutral-beam-heateddischarges were obtained only after degassing the graphite limiterby several tens of low density helium and deuterium discharges.
Evidently, desorption of deuterium from the normally saturated
near-surface region of the graphite enabled the limiter to retain

10'

-

1

1

1

1

1

-

r= 19 Shots
1.4 MA Helium
Conditioning

<

~

-

Shots

10\

0000

°oc

••

\

ne

•

00

°° o
°°oo

-

0

1
10

0

00

1

1

1

1

15

20

25

30

35

SHOTS IN THE CLEANUP SERIES

FIG. 1. D/3 emission intensity and line-averaged electron density during a sequence of low density
1.4 MA helium limiter degassing discharges. A decrease in the deuterium influx of a factor of 20 indicates depletion of deuterium from the surface and near surface regions of the limiter.

210

HILL et al.

2

3
4
TIME (s)FIG. 2. Line-averaged electron density and gas fueling rate as a function of time during fiducial
shots on: (a) a well conditioned limiter (solid line) and a deconditioned limiter (dashed line), and(b) an unconditioned limiter.

incident hydrogen
efficiently
and substantiallyreducedrecycling. In combination with intense central fueling by neutralbeam injection, the lower recycling produced a more peaked density
profile.
During normal enhanced-confinement operation, daily
sequences of about 10 degassing shots were used to keep plasma
performance optimized. About 20-30 degassing shots were required
to recover from gas loading caused by pellet-fueled [6] ordetached plasmas [7].
The hydrogen degassing by a series of 1.4 MA helium
conditioning discharges on the toroidal limiter is illustrated by
Fig. 1. The brightness of the deuterium DB line (n = 4-2transition), which is a measure of the deuterium influx into theplasma, decreased by a factor of 20 during this sequence. The
minimum achievable line-averaged electron density also decreased
by a factor of two, as shown, and oxygen radiation decreased by anorder of magnitude. This suggests that conditioning discharges
also remove oxygen, possibly in the form of water vapor or CO.
The intensity of carbon radiation changed little during thesequence.
Two important consequences of the limiter degassing are asmaller low-density limit and a shorter density-decay time,
constant or effective particle confinement time T D , indicated by
faster density pumpout. Both effects are illustrated in Fig. 2a,which shows the time evolution of the central-chord line-averaged
electron density and the gas input rate for 2 shots, one before
and one after a 25-shot 0.8 MA ^deuterium degassing sequence. In
contrast, a large value of T (> 10 s)
characterized thetoroidal limiter before conditioning sequences were begun(Fig. 2b).
In these fiducial discharges the gas feed iscontrolled by feedback to produce a preselected density at t = 2 s

IAEA-CN-47/A-IV-2

211

and is then shut off.
Figure 2a depicts 0.8 MA discharges
programmed to n e = 1.3 x 10™ m" 3 , whereas Fig. 2b is a 1.4 MA
discharge programmed to fie = 2.4 x 10 1 ' m~ 3 . The density decays
exponentially to a baseline which is determined by recycling. The
density decay time is determined by the intercept of the dashedlines, which characterize the initial decay rate, with the minimum
density baseline.
Decay time constants as small as 0.15 s were obtained after
applying helium degassing sequences frequently over a 1000-shot
interval. The sequences were interspersed with series of neutralbeam-heated shots to test the effectiveness of conditioning on
improving beam fueling and energy confinement. During this period
the minimum line-averaged density achievable in 0.8 MA
deuteriumdischarges decreased from 0.9 x 10 * m~ 3 to 0.55 x 10 ' m~ 3 .
The recycling
coefficientin ohmic discharges, R, isdetermined from T
and the particle confinement time, T , by theequation
T

p=itR

(D

T as measured by the absolute intensity of the Da emission
is on the order of 0.1s. Thus, the fastest density decay times,
0.15 - 0.3 s3correspond to a recycling coefficient of R < 0.5.
Helium discharges with I= 1.1 and 1.8 MA were more
effective in degassing the limiter than were 0.8 MA deuterium
discharges. A few shots in deuterium are required to expel theresidual helium, but the succeeding discharges have lower
densities than can be achieved following conditioning by deuterium
discharges.
There are indications that higher current helium
discharges are more effective than lower current discharges.
The limiter degassing effect is easily reversed by exposure
to higher density deuterium plasmas. An estimated 100 torr-liters(7.1 x 10 21 atoms) of total gas input over a few discharges will
increase the recycling coefficient to approximately 1. Deuterium
gas-fueled 0.8 MA ohmic discharges with n e as low as 1.2 x 10 °m~3 can produce a noticeable degradation.
The limiter degassing is believed to result from desorptionof hydrogen from the graphite by energetic carbon and helium ion
bombardment. Graphite can absorb hydrogen up to a H/C ratio of
0.4 at room temperature.
Spectroscopic measurements and Z f£.
values of six to seven show that the carbon density in degassing
discharges is high.
Recent measurements by Wampler et al. [8]show that bombardment of saturated graphite by carbon, helium, orhydrogen ions with energies of 3, 0.6 and 0.3 keV, respectively(consistent with the edge electron temperature and sheathpotential in these discharges) removes 5, 2 and 1 hydrogen atoms
per incident ion, the latter case resulting in replacement, with

212

HILL et al.
i

Ip==2.2 MA

1 4
z
o:

§ 3
2

oa
Range of
Neutral Beom Power (MW) "~1
Min Moii

« 0
o

I

1

4.0
7.6

4.0

• 7.6
4 10.5o 13.0

^^

a

10.5
13.0
i

1

i

1

2

3

neBEFORE NBl do'V

3

FIG. 3. Zeff during neutral-beam injection as a function of electron density before injection, sorted byranges of neutral-beam power. The values are near the pre-NBI Zeg values, but somewhat reduced
owing to neutral-beam fueling.

no net removal of hydrogenic species.
Also, the yields for
sputtering carbon atoms from graphite are 0.5, 0.09 and 0.025
atoms/ion for these three ions.
Thus the degassing process isaided by the low density, high edge temperature, and the resultant
high levels of carbon impurities.
Ill.

Plasma Impurity Concentrations in the Standard Regime

Impurity concentrations and Z ~ f have been measured by x-ray
and vacuum ultraviolet spectroscopy and visible bremsstrahlung;the total radiated power has been measured by bolometer arrays.
These measurements have documented the effectiveness of impurity
control techniques, provided quantitative data on the variation of
impurities over a wide range of plasma parameters, and provided
additional information on impurity sources and impurity production
mechanisms [9-11].
The impurity situation in ohmically heated TFTR discharges
is discussed
in Refs [9-11]. Impurity concentrations were low^Zeff = 1 " 2 ^ a t h i g h d e n s i t v > b u t w e r e significant ( Z g f f = 5-6) at
the low-density limit for 2.2 MA plasmas on either the movable or
toroidal limiter.
Discharges on the toroidal limiter had lower
Z a*, at intermediate densities than those on the movable limiter[T2]. The low-Z impurities, carbon and oxygen, ranged from about
10? of electron density (at low n e ) to 155 (high n g ) and were thedominant contributors to Zgf«f and radiated power. The ratio ofcarbon to oxygen was about 10 at low density and 1 at highdensity. Oxygen increased with density and dominated the radiatedpower at the high density limit [11]. Metals (Cr, Fe, Ni) weregenerally negligible contributors to 1Q£f and radiated power.

IAEA-CN-47/A-IV-2

213

Impurity behavior in 2.2 MA neutral-beam-heated plasmas on
the movable
limlter with beam power up to 5.6 MW has been
discussed previously [111. In a beam-power scan at constant final
density (ng = 4.5 * 10 1 ' m~3), z~~ near the end of the 0.5 s
heating pulse increased from less than 2 to about 3 as beam powerincreased to 5.6 MW.
During a discharge, metal and carbon
densities remained constant from the pre-beam to the beam-heatedphase. Thus, higher Ze£f at higher power results mainly from thelower pre-beam density (and therefore higher Zeff) required tokeep the final density constant.
The fraction of total input
power radiated decreased from about 50$ in the ohmic phase to 30$
at a beam power greater than 3 MW.
Z
eff v a l u e s i n 2 - 2 MA > L-mode beam-power scans on thetoroidal limiter also correlated much more strongly with theelectron density in the pre-beam ohmic heating phase of discharges
than with beam power or with density at the end of injection (Fig.
3). The data, measured by x-ray pulse-height analysis, have beensorted according to ranges of beam power. Low Z f*> values (< 2)are achieved during neutral-beam injection (NBI) if the ohmic
target plasma is clean (ne > 3 * 10'9 m " ^ ) , while higher Z f fvalues during NBI occur if the ohmic plasma is dirty (lowdensity).
The trend is similar to that of pre-beam Z f^ versus
iie, although shifted slightly lower, due to dilution of impurities
by neutral-beam fueling.
^eff v a l u e s measured from visible
bremsstrahlung emission were on the average 15$ higher than thevalues from pulse-height analysis for most of these shots.

The radiated-power fraction in these L-mode discharges on
the toroidal limiter decreased to about 20$ for neutral beam
powers greater than 10 MW.
IV.

Impurity Concentrations in Enhanced Confinement Plasmas

The pre-injection target plasmas in enhanced confinement
discharges are carbon dominated, with a total ZQff of six to eight
and a Zeff contribution of 0.3 - 1.5 from metals.
The carbon
concentration is about 15$ of the electron density. The deuteronfraction is estimated to be between 10$ and 20$, based on theincrease in neutron emission following a small gas puff. During
neutral beam injection, Z ^ from visible bremsstrahlung drops to
2-4 (Fig. 4 and earlier data). The earliest enhanced confinement
plasmas had Z e f f values of 2-3 during the beam-heating phase, witha negligible contribution from metals.
Subsequent limiter
conditioning and plasma operation, spanning 1000 discharges over a
period of five weeks, increased the metal levels by a factor of 10,
presumably by depositing metals removed from inconel hardware near
the limiter onto the graphite surface.
During neutral beam heating, the fueling by the beam
neutrals increases the deuteron fraction to 50-60$ of electron
density and decreases the carbon fraction to 5-8$. The fraction

214

HILL et at.
1

i

•

X

X

—

7X

o

6 ~x

Sj;

X

X

•i;

5
4 -

X

*

$

"x
*

*

-

°

111

x
D*2

a

DO

j«

-

*x$ + « f « ° j

x n9gnrff 1 B 5 x < D^c

° o o o B°

ij

3 >—

O tl

x 0.8 MA

2_

—

X

-

° 0.9 MA
+ 1.0 MA

1—
1

1

n

1

10
P H E A T (MW)FIG. 4. Zeg as a junction of total heating power in enhanced confinement discharges.

20

•

4

•

i

i

3
/

f
1

.01

10

"=

N

/

1
-

-

n

0
_

\

\

20

10

I
3
3

4
TIME (s)FIG. 5. Total radiated power, line-averaged electron density, neutral-beam power and poloidal beta,
as a Junction of time during a 1.1 MA enhanced confinement discharge with a 0.7 s beam-heating pulse.

IAEA-CN-47/A-IV-2

215

of total heating power radiated decreases from 60-70? during theohmic phase to about 30? for beam powers greater than 10 MW. For
longer pulse (0.7 s)
beams the total radiated power saturates
while electron density and 8 are still rising, as shown in
Fig. 5.
The impurity transport during beam heating in enhanced
confinement discharges appears to be similar to that during theohmic phase.
Transport rates were inferred from the time
evolution of vacuum-ultraviolet lines emitted by germanium
injected into the plasma by laser ablation. Comparison of these
intensities with numerical impurity transport code predictions[13] yielded a diffusion coefficient (assumed to be radiallyconstant) and convective velocity at the limiter radius of 0.65
nr/s and 0 - 0 . 1 m/s, respectively, in the ohmic phase, and 0.75
m2/s and 0.1 - 0.2 m/s, respectively, during neutral beam
injection. The uncertainty in the convective velocity is a factorof 2.
The impurity confinement time in both cases wasapproximately 0.25 s.

V.

Summary and Conclusions

A disruptive discharge cleaning procedure, in addition to
glow- and pulsed-discharge cleaning, was found necessary following
atmospheric exposure to remove water vapor from the large-areagraphite • toroidal
limiterandpermitsatisfactory
plasmaoperation.
Low density deuterium or helium discharges were found to beeffective in degassing the limiter, allowing it to pump hydrogenand, thus, reduce recycling and permit improved confinement inneutral-beam-heated discharges.
High carbon concentrations andf a P P e a r t° a*-d degassing by causing high edge temperatures anding to desorption of deuterium by energetic carbon ion
bombardment.
Relatively low Zef>f. values (about 2) were attained in highdensity L-mode discharges on the toroidal limiter, with beam powers
up to 10 MW, and values of 2.5 - 3 were found at beam powers up to
15 MW. The trends suggest, however, that the purity of the prebeam target plasma is a more important determinant of impurity
concentrations during NBI than is beam power or final density.
Enhanced confinement plasmas are characterized by a Z ff. ofsix to seven before neutral beam injection and two to four during
injection. Zeff decreases with beam power, presumably due to thedominance of beam fueling in these plasmas.
There is no evidence of significant changes in impurity
transport between ohmic and enhanced confinement discharges.

216

HILLetal.
Acknowledgments

We are grateful to J.R. Thompson and D. Grove for theirsupport. This work was supported by US DOE Contract No. DE-ACO276-CHO3O73. The ORNL participants were also supported by US DOE
Contract No. DE-ACO5-84OR214OO with Martin Marietta Energy
Systems, Inc.

REFERENCES[1] STRACHAN, J.D., et al., submitted to Phys. Rev. Lett.
[2] HAWRYLUK, R.J., et al., paper IAEA-CN-47/A-I-3, these Proceedings, Vol. 1.
[3] GOLDSTON, R.J., et al., Paper IAEA-CN-47/A-I-4, these Proceedings, Vol. 1.
[4] DYLLA, H.F., et al., submitted to Nucl. Fusion.
[5] LA MARCHE, P.H., et al., J. Nucl. Mater. 145&146 (1987).
[6] MILORA, S.L., et al., paper IAEA-CN^7/A-HI-4, these Proceedings, Vol. 1.
[7] STRACHAN, J., et al., in Controlled Fusion and Plasma Physics (Proc. 12th Europ. Conf.
Budapest, 1985), Pt. I, European Physical Society (1985) 339.
[8] WAMPLER, W., DOYLE, B., BRYCE, D., submitted for publication.
[9] HILL, K.W., et al., Nucl. Fusion 26 (1986) 1131.
[10] DYLLA, H.F., et al., J. Nucl. Mater. 145&146 (1987).
[11] STRATTON, B.C., et al., submitted to Nucl. Fusion.
[12] BELL, M.G., et al.. Plasma Phys. Controll. Fusion 28 (1986) 1329.
[13] STRATTON, B.C., et al., J. Nucl. Mater. 145&146 (1987).

DISCUSSION
F.C. SCHULLER: It is normally accepted that the high density limit for disruptions decreases with increasing Zeff. You quote for the high density Ohmic discharges a Zeff which is considerably lower than that in JET, whilst your density limit
is lower. Do you get 100% radiation just before disruption?
K. W. HILL: In Ohmic discharges, the high density limit in TFTR is higher thanthat in JET; however, the Murakami parameters are comparable. The good agreement in the Zeff determination between the various diagnostics, as described in thepaper, gives us confidence in our low values of Zeff at high density. At the density
limit the power radiated is about 100%. The question then arises whether the usual
Murakami/Hugill parameter coupled with Zeff determines the operating boundary
accurately. Perhaps the operational limit depends, in addition, on plasma elongationand on the toroidal field — in a manner other than that implied by the relationshipgiven by the Murakami/Hugill parameter.

IAEA-CN-47/A-IV-3

IMPURITY AND MHD BEHAVIOURIN JT-60 DIVERTOR DISCHARGES
JT-60 TEAM
(Presented by H. Takeuchi)
T. Abe, H. Aikawa, H. Akaoka, H. Akasaka, M. Akiba, N. Akino, T. Akiyama,
T. Ando, K. Annoh, N. Aoyagi, K. Arakawa, M. Araki, K. Arimoto, M. Azumi,
S. Chiba, M. Dairaku, N. Ebiswa, T. Fujii, T. Fukuda, H. Furukawa,
K. Hamamatsu, K. Hayashi, M. Hara, K. Haraguchi, H. Hiratsuka, T. Hirayama,
S. Hiroki, K. Hiruta, M. Honda, H. Horiike, R. Hosoda, N. Hosogane, Y. Iida,
T. Iijima, K. Ikeda, Y. Ikeda, T. Imai, T. Inoue, N. Isaji, M. Isaka, S. Ishida,
N. Itige, T. Ito, Y. Ito, A. Kaminaga, M. Kawai, Y. Kawamata, K. Kawasaki,
K. Kikuchi, M. Kikuchi, H. Kimura, T. Kimura, H. Kishimoto, K. Kitahara,
S. Kitamura, A. Kitsunezaki, K. Kiyono, N. Kobayashi, K. Kodama, Y. Koide,
T. Koike, M. Komata, I. Kondo, S. Konoshima, H. Kubo, S. Kunieda,
S. Kurakata, K. Kurihara, M. Kuriyama, T. Kuroda, M. Kusaka, Y. Kusama,
S. Maehara, K. Maeno, S. Mase, S. Matsuda, M. Matsukawa, T. Matsukawa,
M. Matsuoka, N. Miya, K. Miyati, Y. Miyo, K. Mizuhashi, M. Mizuno, R. Murai,
Y. Murakami,
M. Muto,M. Nagami,
A. Nagashima,
K. Nagashima,
T. Nagashima, S. Nagaya, H. Nakamura, Y. Nakamura, M. Nemoto, Y. Neyatani,
S. Niikura, H. Ninomiya, T. Nishitani, H. Nomata, K. Obara, N. Ogiwara,
Y. Ohara, K. Ohasa, T. Ohga, H. Ohhara, M. Ohkubo, T. Ohshima, K. Ohta,
M. Ohta, M. Ohtaka, Y. Ohuchi, A. Oikawa, H. Okumura, Y. Okumura,
K. Omori, S. Omori, Y. Omori, T. Ozeki, M. Saigusa, K. Sakamoto, A. Sakasai,
S. Sakata, M. Satou, M. Sawahata, M. Seimiya, M. Seki, S. Seki, K. Shibanuma,
R. Shimada, K. Shimizu, M. Shimizu, Y. Shimomura, S. Shinozaki, H. Shirai,
H. Shirakata, M. Shitomi, K. Suganuma, T. Sugie, T. Sugiyama, H. Sunaoshi,
K. Suzuki,
M. Suzuki,
M. Suzuki,
N. Suzuki,
S. Suzuki,
Y. Suzuki,
H. Takahashi1, M. Takahashi, S. Takahashi, T. Takahashi, M. Takasaki,
M. Takatsu, A. Takeshita, H. Takeuchi, S. Tamura, S. Tanaka, K. Tani,
M. Terakado, T. Terakado, K. Tobita, T. Tokutake, T. Totsuka, N. Toyoshima,
H. Tsuda, T. Tsugita, S. Tsuji, Y. Tsukahara, M. Tsuneoka, K. Uehara,
M. Umehara, Y. Uramoto, H. Usami, K. Ushigusa, K. Usui, J. Yagyu,
K. Yamada, M. Yamamoto, O. Yamashita, Y. Yamashita, K. Yano, T. Yasukawa,
K. Yokokura,
H. Yokomizo,
H. Yoshida,
M. Yoshikawa,Y. Yoshinari,
R. Yoshino, I. Yonekawa, K. Watanabe

Naka Fusion Energy Research Establishment,
Japan Atomic Energy Research Institute,Naka-machi, Naka-gun, Ibaraki-ken,
Japan

1

Princeton Plasma Physics Laboratory, Princeton University, NJ, USA.

217

218

ABEetal.
Abstract

IMPURITY AND MHD BEHAVIOUR IN JT-60 DIVERTOR DISCHARGES.
The diffusion coefficient (DA) and the inward peaking factor (C v ) of titanium are estimated fromthe experimental evolution of the titanium lines and the calculational result of a simulation code:
DA = 1.0 m 2 -s~' and Cv = ~ 0 for ohmically heated plasmas andDA = 0.6-0.8 m 2 -s~' andC v = 0 to 1 for neutral beam (NB) injection heated plasmas. No difference in titanium transport isobserved with respect to the direction (co- or counter-) of NB injection. There is no strong titanium
pinch at the centre in divertor discharges, either with or without sawtooth oscillation. — In discharges
with Ohmic and NB heating, various MHD activities are observed. The transition from single tocompound sawteeth took place at a resistive diffusion time of TR = iior*/i} a 3.5 s. It is confirmed thatthe inversion radius observed in a wide range of qeff = 2.6-6 depends linearly on l/qeff, but is independent of density, radius, plasma current or heating power.

1.

INTRODUCTION

To attain an equivalent breakeven plasma condition, impurities and MHD
activity are the most important factors resulting in a degradation of confinement
properties. A single null poloidal divertor has been installed in JT-60 in order toeliminate the impurities from the main plasma; the MHD activity in the JT-60
divertor configuration was investigated.
Plasma heating experiments with a neutral beam (NB) heating power of up to
20 MW were started in August 1986 with an injection energy of Eb = 50-70 keV,a beam pulse length of up to 3 s, by eight co- and six counter-injection units withnearly perpendicular angle (78° to the magnetic axis). Hydrogen or helium plasmas
in divertor or limiter configurations in the range of 1 < Ip < 2 MA and1.9 < B, < 4.5 T were used as the target. This paper reports impurity and relevant
MHD behaviour of ohmically and NB heated plasmas in divertor discharges.
In Section 2, quantity and species of the impurities and their behaviour aredescribed. Section 3 is devoted to the MHD behaviour of compound sawtooth andfluctuation during NB heating. These results will be discussed and summarized in thelast section.

2.

IMPURITY BEHAVIOUR

Figure 1 shows typical spectra in the wavelength range of 5 to 500 A measured
by a unit type flat field grazing incidence spectrometer in a divertor hydrogen
discharge with B, = 4.5 T and Ip = 2 MA. The major impurities were identified tobe oxygen, carbon and titanium, which were the coating materials of the first wall.
A small amount of nickel, iron and chromium was observed in divertor discharges;
molybdenum lines were also identified in limiter discharges.

219

IAEA-CN-47/A-IV-3

200

300

400
CHANNEL

500

600

700

800

NUMBER

FIG. I. Typical VUV spectrum recorded by unit type spectrometer, integrated over 20 ms for
B, = 4.5 T, Ip = 2 MA in divertor discharge.

The amount of titanium was estimated from soft X-ray measurements. In
ohmically heated discharges, the amount of titanium was around 1 X 10 ~3% and
8 x 10~3% at n,. = 2 x 1013 cm" 3 in hydrogen and helium discharges, respectively. The ratio of titanium to line averaged electron density decreased with theincrease in electron density. For NB heated hydrogen plasmas, the ratio of titanium
to electron density became about twice that in ohmically heated plasmas. And thequantities of nickel, iron and chromium were below 1/10 to 1/100 of that of titanium.
The amount of light impurities in hydrogen discharges was measured by a
Rutherford particle scattering technique, using a diagnostic helium beam with anenergy of 170 keV. This scattering technique is insensitive to metal impurities, forthe present plasma parameters, and does not distinguish between light impurity
species. The ratio of light impurity to proton density was 1.2 % or 2.4 % if the species
were assumed to be totally oxygen or carbon, respectively. Zeff as obtained by thismethod was 1.5, which was equal to the values obtained from visible bremsstrahlung
and simple Spitzer resistivity. Therefore, it was suggested that light ions were thedominant impurities in JT-60, although the main species contributing to Zeff were
not yet identified by the spectroscopic measurements. Zeff decreased with theincrease in electron density, and the Zeff values of NB heated plasmas were slightly
higher than those of ohmically heated plasmas.
The metal impurity (titanium) in plasmas in the limiter configurations is muchhigher than that in divertor configurations; this difference was exploited to intentionally inject titanium into the plasma. Specifically, when a titanium injection into adivertor discharge was needed the configuration was converted from a divertor to alimiter configuration within 50 ms, remained in the limiter configuration for 50 ms
and then again became a divertor configuration.

220

ABE et al.
Co-injection

Ctr-injection

60

6.6
t Is)

6.4

6.2

©DA=0.6r As

(b)s

_10

C V =O« 8-

C v =1.0"

Cy=1.0

6.0

6.2

GO

i

6.8

Jl
in

7.0

TiXXI
Ka line

T

1

~ 2
6.0

6.2

6.4

6.6
t (s>

6.8

6.4

6.6
I Is)

6.8

7.0

FIG. 2. Comparison of evolutions of (a) soft X-ray, (b) Ti XXI with co- and counter-NB injection.
Titanium is injected intentionally by switching from divertor to limiter configuration and back to divertor
configuration.

Figure 2 shows the time evolution of 15-channel soft X-ray diode array signals
and the Ti XXI line intensity as measured by a Johan type crystal spectrometer. In
this figure, a NB heating pulse of 7.7 to 8.4 MW, either in co- or counter-direction,
was started at 6.0 s from the start of the discharge, and the titanium injection
procedure of 150 ms as described above was started at 6.3 s. The targets werehydrogen plasmas with S . S 5 X 1019 m" 3 , Ip = 1.5 MA and Bt = 4.5 T. The
Ti XXI line intensity decreased with a decay time of ~ 200 ms if the minor disruptions were smoothed out, and there was no difference in decay time for different
injection angles. A large m = 1/n = 1 oscillation was observed as the precursor ofinternal disruptions in the discharges with counter-injection. The toroidal rotation
speed of the plasma was also calculated with MHD oscillation data, assuming an
m = 1/n = 1 mode; the results were ~ 4 X 104 m-s" 1 and ~ 8 X 104 m-s" 1 for
8.4 MW co- and 7.7 MW counter-injection, respectively.
The decay of the Ti XXI line intensity was simulated by a 1-D impurity
transport code [1]. The best fittings were obtained with DA = 0.8 m 2 -s"' andC v = 1 for the co-injection case, and with DA = 0.6 m 2 -s~' and C v = ~ 0 for thecounter-injection case. Considering that DA = 1.0 m 2 > s"' and C v = —0 for

221

IAEA-CN-47/A-IV-3

ohmically heated discharges [2], it was concluded that there was no significant difference in impurity transport among these three cases. At the same time, it was suggested that the impurity transport was not affected by a toroidal rotation of(4-8) X 104 m-s - l

3.

MHD BEHAVIOUR

In the current flat top phase, the onset of sawtooth oscillations was normallyobserved around 2-4 s after breakdown. These sawteeth were observed in the soft
X-ray, ECE (electron cyclotron emission) and line electron density signals in bothhydrogen and helium discharges. As is shown in Fig. 3(a), large m = odd successor
and precursor oscillations were observed in the local and full reconnection phases
during a compound sawtooth. The frequency of these oscillations was in the rangeof 0.2-1.2 kHz, and the density fluctuations were found to be clearly correlated tothe soft X-ray signals. The repetition times of the normal sawtooth oscillation
discharges in ohmically heated plasmas were 25-30 ms for Ip = 1 MA and
50-80 ms for Ip =2 MA, as is shown in Fig. 4, and the compound sawteeth with arepetition time of 70-140 ms were also frequently observed in discharges with a large
plasma current (Ip > 1.5 MA).

toJ r = -9.3cm

q 1.5

j r=-l6.5cm

M

/

*<*

r»-20.0cm" /•19.3 cm .'t—or»-9.3cm r—16.5 cm
r«-20.0cm-

1

v kL

U
05
6.0

6.1

6.2 6.3
Time (s)

6.4

6.6

6.7

6.8

6.9

7.0

Time (s)

6.5

/

\

r=-l6.5cm\ r—20.0cm

1.0

6.6

6.7

6.B
Time (s)6.9

7.0

FIG. 3. Characteristics ofJT-60 MHD behaviour, (a) Soft X-ray signal in ohmically heated divertor
discharge of hydrogen gas. Soft X-ray signal of (b) continuous and (c) MHD burst during NB heating.

222

ABE et al.
400

FIG. 4. Dependence of sawtooth period TS, on line averaged electron density in divertor discharge
with Ohmic and NB heating.

The transition of the normal sawtooth to the compound oscillation one wasanalysed by the 1-D tokamak code incorporating the multi-tearing instability. The
onset condition for the compound sawtooth oscillation in JT-60 given by this simulation was that the resistive diffusion time TR > 3.0 s. This value nearly agreed withthe experimental results (rR > 3.5 s) of the critical plasma current for the onset ofthe compound oscillation and was also consistent with the experimental result that thecompound oscillation was only rarely observed in helium discharges.
Except for dirty discharges, enhanced sawtooth activity as shown in Fig. 3(b)was observed in discharges with high power NB heating. Magnetic fluctuations were
measured with a magnetic probe array of 11 coils located at about 5 cm from thelimiter surface. The fluctuation level of 5Bp/Bp was higher than 0.02%, with afrequency of ~ 1 kHz. The mode was identified as m = 4 and synchronized with thecentral m = odd oscillation.
MHD bursts as shown in Fig. 3(c) were observed for a NB heating power ofover 17 MW and a beam energy of over 70 keV. The central fluctuation observedwith soft X-ray diodes was a compound of an m = odd mode with 1 kHz and a higher
mode fluctuation. In one of two cases, the m = 3 magnetic fluctuation of 4 kHz wassynchronized with the higher frequency mode of the central fluctuation, and in theother case the magnetic fluctuation was synchronized with the lower frequency modeof the central fluctuation. Considering that the toroidal rotation frequency, from
MHD measurements, was comparable to the trapped particle precession frequency
and that the central deposition of the intense neutral beam results in the generationof deeply trapped fast ions, and also considering the relatively high poloidal beta inthese discharges, it turned out that the MHD bursts were a result of mode-particle

223

IAEA-CN-47/A-IV-3o•OH plosma
NBI plasma
DIVA, eip.

D

o

rs/a

/<>

o

including\
0 / L , H/Hc 1

/
•

i

,

i

O.I

0.2

0.3

0.4

c

0.5

f /G. 5. Dependence of inversion radius rs on Hqeg, where
2ira 2 B,
Qeff

=

[-a:

— (/Jpresonance such as the fishbone [3]. However, the inferred flux of fast ions ejectedfrom the plasma by the instability was not detected with the charge exchange neutral
particle analysers, probably because of a low sampling rate and the viewing angle of
the analyser which is looking at the plasma centre and is insensitive to the ejected
fast ions. No significant effect of this instability on plasma confinement has been
observed so far, but more investigations of this instability are needed in the future.
Figure 4 shows the sawtooth repetition time TS, as a function of the line
averaged electron density for discharges with Ohmic and NB heating. No differencein TS, was observed between hydrogen and helium plasmas. Apparently, the repetition time for the ohmically heated plasmas is almost independent of the electron
density, the enhancement of the repetition time due to NB heating was longer in the
low density regime and the repetition time for NB heated plasmas approached thatof ohmically heated plasmas at fie > 5 X 10" m~3.
Figure 5 shows the normalized inversion radius of the sawtooth rs/a as a function of the inverse of the safety factor l/qeff, where rs, a and qeff are the inversion
radius, the plasma minor radius and the safety factor, respectively. The verified dataof JT-60 for discharges with various parameters are plotted in Fig. 5. The DIVA data
[4] are also plotted in the figure. Figure 5 indicates that rs/a depends only on l/qeffand is independent of all other parameters such as plasma configuration, heating
power, ion species, etc.

4.

DISCUSSION AND SUMMARIES

Clean plasmas were obtained in divertor configurations, and the major impurities were identified to be oxygen, carbon and titanium. The diffusion coefficient and224

ABE et al.

the peaking factor, were estimated to be DA = 1.0 m 2 -s"', C v ~ 0 in ohmicallyheated plasmas and DA = 0.8 m 2 -s~', C v ~ 1 and DA = 0.6 m 2 -s"', C v ~ 0
in discharges with sawtooth oscillation with co- and counter- NB injection,
respectively. The difference in the direction of the NB injection did not bring about
any significant difference in titanium impurity transport.
Considering the values of DA and C v in ohmically heated discharges without
sawteeth, it is indicated that there was no significant difference in the impurity transports between ohmically heated plasmas without sawteeth and NB heated plasmas
with sawteeth. The difference in the heating method or the existence of sawtoothoscillations does not affect the above conclusion. In NB heated plasmas, magnetic
fluctuations with 6Bp/Bp > 0.02% do not seem to enhance the impurity influx,
probably because of the sufficient thickness of the scrape-off layer.
The sawteeth had a compound structure in high current discharges and tended
to have large m = 1, continuous oscillations or large MHD bursts in NB heateddischarges. The transition from the normal to the compound sawtooth took place ata resistive diffusion time of TR > 3.5 s. The repetition time of the sawteeth inohmically heated plasmas was almost constant, for a wide range of plasma densities.
The repetition time in the NB heated plasmas was enhanced in the low plasma density
region and tended to approach that of the ohmically heated plasmas with high plasma
density. The inversion radius of the sawteeth was determined experimentally in awide range of parameters, qeff = 2.6-6, fie = (1-9) X 1019 m~3, Ip = 1-2 MA,
PNB = 5-20 MW in divertor and limiter configurations. The normalized inversion
radius rs/a depends only on l/qeff and is independent of all other parameters.
ACKNOWLEDGEMENTS
The authors wish to thank the members of JAERI who have contributed to the
JT-60 project throughout its progress. They further wish to express their gratitude
to Drs S. Mori, Y. Iso, and K. Tomabechi for continued encouragement and support.
REFERENCES
[1]
[2]
[3]
[4]

HIRAYAMA, T., et al., in Plasma-Surface Interactions in Controlled Fusion Devices (Proc. 7th
Int. Conf. Princeton, 1986) (to be published).
HIRAYAMA, T., et al., Rep. JAERI-M 86-161 (October 1986).
McGUIRE, K., et al., Phys. Rev. Lett. 50 (1983) 891.
DIVA GROUP, Nucl. Fusion 20 (1980) 271.

DISCUSSION
R.J. GOLDSTON: You discussed oscillations at high beam power which you
thought might be related to PDX fishbones. Could you tell us how their frequency
is related with the fast ion precession frequency?

IAEA-CN-47/A-IV-3

225

H. TAKEUCHI: The fast ion precession frequency is comparable to the toroidal
rotation frequency from MHD measurements.
K. McGUIRE: I noticed that your soft X-ray signal showed a very large
increase in the first second. Do you have a large influx of impurities during this time?
H. TAKEUCHI: The JT-60 discharge started with the limiter configuration,
and at 0.33 s the limiter configuration was switched to divertor type configuration.
After the switching, the impurity flux was gradually reduced by the divertor effect.

IAEA-CN-47/A-IV-4

DENSITY LIMIT AND
IMPURITY TRANSPORT INVESTIGATIONSIN DITE TOKAMAK
J. ALLEN, G.E. AUSTIN, K.B. AXON, R. BARNSLEY1,
M. DUNSTAN, D.N. EDWARDS, K.D. EVANS1,
S.J. FIELDING, D.H.J. GOODALL, W. HAN, N.C. HAWKES,
J. HUGILL, P.C. JOHNSON, G.M. McCRACKEN,
G.F. MATTHEWS, W. MILLAR, N.J. PEACOCK,
C.S. PITCHER, J. PRTTCHARD, A. SYKES,
M.F. TURNER, J.P. WEBBER
Euratom-UKAEA Fusion Association,
Culham Laboratory,Abingdon, Oxfordshire,
United Kingdom

Abstract
DENSITY LIMIT AND IMPURITY TRANSPORT INVESTIGATIONS IN DITE TOKAMAK.
The processes leading to disruption with increasing density have been studied. The discharge
remains in thermal equilibrium but there is a slow profile contraction as the density is raised becauseof a deterioration in edge confinement relative to the centre. This is only partly explained by radiation
losses. — The central confinement time of impurities injected by laser ablation has been measured. At
high density in ohmically heated discharges, it can be varied by more than a factor of five by changing
the gas feed rate. The largest values occur for refuelling by recycling only. Substantial differences also
occur with co- and counter-neutral beam injection, which are not sensitive to the gas feed rate in thiscase. These differences cannot be mainly ascribed to plasma rotation because they are not removed byswitching on the bundle divertor, which reduces the rotation speed by an order of magnitude. An
accumulation of intrinsic impurities is observed in discharges with long impurity confinement times. The
effects described are much reduced or absent at low density.

1.

INTRODUCTION

Recent experimental work on DITE tokamak has concentrated
on studies of the limit set by disruption at high density andthe confinement of impurities with and without the bundle
divertor.

The main machine parameters used were:major

radius, R=1.17m; poloidal carbon limiter radius = 0.26m;
adjustable rail limiter and divertor separatrix radius =1

University of Leicester, Leicester, UK.

227

228

ALLEN et al.

0.21m; toroidal field, B = 2.0-2.6T; plasma current, I =0.1-0.25MA; tangential neutral H beam injection, P. . ;< 1.6MW
at E = 25 keV. The working gas was generally helium forhighest density operation, with some lower density work indeuterium.
Preliminary results on the density limit were given
previously [l]« Here, we summarise the main conclusions ofmore recent work. Studies of impurity transport were intended
to complement previous work on energy and particle exhaust indiverted discharges [2] and to explain the generally lower
intrinsic impurity levels observed with both ohmic and beam
heating. Non-diverted discharges used for comparison purposes
showed some interesting new behaviour which is reported in
Section 3.
2.

INVESTIGATION OF THE DENSITY LIMIT

Disruption at high density is almost invariably precededby rapidly growing poloidal field oscillations of low mode
number. The aim of the present study is to identify theprocesses leading to their destabilisation: in particular todetermine whether there is a radial collapse of the discharge
due to thermal instability or a sudden change in transport due
to the onset of other instabilities, such as ballooning modes.
Relatively high q-values of 4-8 were employed to avoid effects
due to operation near q=3 and q=2.
In ohmically heated discharges there is a gradualnarrowing of the plasma current and temperature profiles.
A
reduction in peripheral temperature is measured by probes andMarfes [3] may appear, especially in helium. The soft X-ray
sawtooth inversion radius increases and the radial positionsof peaks in line radiation from low charge states of carbon
and oxygen ions move inwards. These last two effects areillustrated in Fig.l. For q>6 profile narrowing leads to theappearance of detachment from the limiter [4], However, there
is no rapid profile contraction due to an unstable thermal
equilibrium. Detached discharges can be maintained for times>20 -cE.
Disruption is preceded by a rapid growth (y 5 10 3 s""*) inthe amplitude of poloidal field fluctuations, usually

IAEA-CN-47/A-IV-4

> 0-8

M

J,

j,
—,

229

Sawtooth reconnexion radiusfrom soft X-ray diodes= 06^ Maxima in line radiation" ' from low-Z impurities

I <K
g 02
uj

0-1

0-2
Minor radius (m)FIG. 1. Profile narrowing in deuterium discharge at Ip = 102 kA, BT = 2.5 T, as evidenced by SXR
and impurity line radiation. Curves are model profiles consisting of truncated Gaussians with the same
total Spitzer resistance. For line radiation data, abscissa represents excitation energy.
,o - np = 1.5 x 1019 nc3;
, • - n = 2.7 X 1019 m"5.

following sawtooth relaxation. The principal mode numbers are
m=2, n=l with a smaller component at 3,2 and possibly others.
Estimated pressure gradients are at least a factor of 2 toosmall to destabilise ideal ballooning modes.
Profile narrowing may be partly explained by radiation
losses, which are relatively stronger in the periphery. A
numerical simulation [5] assuming constant thermal
conductivity, increased 100 x for q<l, and coronal radiation
losses from 0.1% Fe, 1% 0 and 2% C shows gradual profile
contraction due to increasing radiation loss as the density is
increased. The q=2 surface is typically unstable to a tearing
mode but, at low density, flattening of the current density
across the magnetic island limits its growth. At a criticaldensity the solution becomes non-linearly unstable, withoscillating and overlapping islands from 2,1 and 3,2 modes.
This model shows that radiation losses of ~ 75% at q=A canindeed produce sufficient contraction to destabilise tearing
modes to the point of disruption (overlapping of 2,1 and 3,2modes) without producing thermal instability. It is deficientin predicting a 2,1 tearing mode island at low density which
is not observed.

230

ALLEN et al.

The experimentally measured radiation loss at disruptionis ~65% at q=4 rising to ~80% at q>6. However, the limiters
receive only about 10% of the input power, the remainder being
so far unaccounted for. Line radiation from a large Marfe
indicates local temperatures in the range 5-15eV and canaccount for up to 50% of the total radiation. A substantial
fraction of the input power must be conducted to the peripheryof the discharge in this case to support the losses in the
Marfe. Attempts to raise the peripheral temperature by
applying neutral beam injection heating produce instead anincrease in radiation from the Marfe. The density limit isincreased but not indefinitely. At q=4 the Murakami parameterincreases from 2.2 to 4.0 in deuterium and from 3.8 to 5.3 inhelium, with up to 1.5 MW of neutral beam power. The Marfe
can be avoided by adjusting the plasma position in the limiter
aperture but this does not affect the density limit
appreciably. It appears to be a symptom of low edge
temperature rather than the cause of it.
In detached discharges the Marfe is usually absent.
However, probe measurements of Mach number inside the limiterradius, which show no appreciable flow in normal discharges,
reveal substantial flow along the magnetic field (up to Mach0.4) in detached discharges. These measurements indicate
large scale convection in the peripheral plasma whose flow
pattern however, could not be fully explored with the existingprobes.

3.

IMPURITY TRANSPORT IN NON-DIVERTED DISCHARGES

3.1

Experimental method

The transport of impurities was investigated using pulsed
sources of impurity, principally aluminium introduced by laser
ablation. A Bragg X-ray spectrometer [6] yielded most of thespatial, temporal and spectral survey information about
central impurities. Additional data was provided byinstruments covering the ultraviolet and visible spectral
regions and by two orthogonal soft X-ray cameras and abolometer camera.

231

IAEA-CN-47/A-IV-4

Spatial scans show that AA XII and Ai, XIII emission lines
originate from the discharge centre. We find that T (o)varies little during the period of measurement and thus therate of decay of either line, after correction for the density
variation, provides a direct measurement of the central
impurity confinement time, x.
imp
3.2

Ohmically heated discharges
A new result is the strong dependence of -r

on thedischarge dynamics at high density [7]. Fig.2 shows x.
versus "n for two density scans in helium. In one scan n
was maintained approximately constant by recycling, x•
increases with "n and reaches values in excess of 150ms. Atehigh density an accumulation of intrinsic impurities isdeduced from the soft X-ray and bolometer camera data, similar
to that seen on Pulsator [8] and more recently in B-regime
discharges on T-10 [9]. In the second scan n was ramped byadditional gas feed during the period of measurement. At low
density x
is not affected but it reaches a maximum and thenimp _decreases as n is raised. The latter dependence is similar tone (10'V 3 )FIG. 2. Variation of impurity confinement time with mean line of sight density for aluminium injectionin Ohmic Umiter discharges, o , A — rising density {dnjdt > 1.5 x 10?° m~3-s~'); • , A —constantdensity (dnjdt < 2.5 X 1019 nf3-^1)- ° . • - Al XII; A , A - Al XIII.

232

ALLEN et al.
250

n e (10 19 iTf 3 )FIG. 3. Variation of impurity confinement time with mean line of sight density for aluminium injection
in neutral beam heated limiter discharges, o, a, A — co-injection; • , • , A — counter-injection.
Q, • — 0.4 MW; a, m - 0.8 MW; A , A — 1.1 MW.

that in Alcator, which resulted in a scaling for %.
imp
— r i
independent of n L^J*
The density at which the two scansdiverge is about 50% of the disruptive limit and increases by
a factor of 1.5 when the current is raised from lOOkA to
150kA. In spite of the factor of five variation inx.
at high density, there is no indication of correspondingimpr .
changes in energy transport, as observed in T-10 \_9\; andthere is no change in the period of sawtooth oscillations,which is approximately proportional to n .
3.3 Beam heated discharges
The variation of t.
with n in discharges heated by coand counter-injection is shown in Fig.3. In this caserefuelling by the beams cannot be avoided and additional gas
Toroidal rotation speeds offeed does not affect %imp*
51
10 me~ are found, decreasing with increasing density. At
low density co- and counter-injection give similar values forT. , as observed on PLT [ill. However, at high density,
imp
counter-injection gives higher values of -u.
than
co-injection, as seen on ISX-B [l2], and t.
decreases, forboth directions, as the injection power is raised.

IAEA-CN-47/A-IV-4
4.

IMPURITY BEHAVIOUR WITH BUNDLE DIVERTOR

4.1

Ohmically heated discharges

233

The confinement of injected impurities in diverted
discharges is broadly similar to that in limiter discharges.
The highest values of x.
again occur without gas feed, which
impresults in a falling density due to particle exhaust. The
amount of injected impurity reaching the centre is not reducedby the divertor, thus the reduction of intrinsic impurity
content [2] is not explained by a screening effect. It isconsistent with a decrease in impurity generation at the walls
and limiters due to the divertor exhaust.
4.2

Beam heated discharges

At high density, diverted plasmas have somewhat larger
values of x.
for co-injection, but a substantial difference
imp
between co- and counter-injection remains. Since the divertor
field reduces the speed of rotation produced by injection by
an order of magnitude [l3] it is clear that rotation is not
the only factor producing this difference.
5.

CONCLUSIONS

5.1

Density limit

Disruption at high density is preceded by a gradualnarrowing of the temperature and current profiles, while thedischarge remains In thermal equilibrium. Disruption isimmediately preceded by a rapidly growing 2/1 mode. Profile
contraction Is partly explained by radiation losses but thereis evidence that increased peripheral transport losses also
play a role.
5.2

Impurity transport

At high density, %is strongly affected by refuellingand by the direction (and magnitude) of neutral beam
injection. With no gas feed or with counter-injection large

234

ALLEN et al.

values of x
and accumulation of intrinsic impurities areimp
observed. These effects are reduced or absent at low
density.
Application of the divertor has no significant effect onx,in ohmically heated discharges. The toroidal rotation
imp
produced by beam injection is reduced by an order ofmagnitude, however substantial differences in x. for co- andcounter-injection remain. Thus the role of rotation in
impurity transport remains to be clarified.
Although the variation in x. in high density ohmicallyimpheated discharges is produced experimentally by changes inedge conditions, the effect appears to be associated with thecentral transport of background plasma rather than with edge
processes.

ACKNOWLEDGEMENTS
We would like to acknowledge the contribution of D. Bell,R.W. Storey and the operations team in maintaining andoperating the DITE experiment and neutral beam injectors.

REFERENCES[1] ALLEN, J.M., et al., Plasma Phys. Control. Fusion 28 (1986) 101.
[2] JOHNSON, P.C., et al., in Plasma Physics and Controlled Nuclear Fusion Research 1984 (Proc.
10th Int. Conf. London, 1984), Vol. 1, IAEA, Vienna (1985) 239.
[3] LIPSCHULTZ, B., et al., Nucl. Fusion 24 (1984) 977.
[4] McCRACKEN, G.M., et al., to be published in J. Nucl. Mater. (1987).
[5] HOPCRAFT, K.I., TURNER, M.F., Phys. Rev. Lett. 56 (1986) 2372.
[6] BARNSLEY, R., et al., Rev. Sci. Instrum. 57 (1986) 2159.
[7] BARNSLEY, R., FIELDING, S.J., HAWKES, N.C., et al., submitted to Phys. Rev. Lett.
[8] ENGELHARDT, W., et al., in Plasma Physics and Controlled Nuclear Fusion Research 1978
(Proc. 7th Int. Conf. Innsbruck, 1978), Vol. 1, IAEA, Vienna (1979) 123.
[9] BAGDASAROV, A.A., et al., in Controlled Fusion and Plasma Physics (Proc. 12th Europ.
Conf. Budapest, 1985), Vol. 9F/1, European Physical Society (1985) 207.
[10] MARMAR, E.A., RICE, J.E., ALLEN, S.L., Phys. Rev. Lett. 51 (1983) 455.
[11] SUCKEWER, S., et al., Nucl. Fusion 24 (1984) 815.
[12] ISLER, R.C., et al., Nucl. Fusion 23 (1983) 1017.
[13] HAWKES, N.C., PEACOCK, N.J., Nucl. Fusion 25 (1985) 971.

IAEA-CN-47/A-IV-4

235

DISCUSSION
K. IDA: Is the impurity transport observed in DITE neoclassical transport? Ifso, can you use the neoclassical theory to explain the long confinement time for impurity injected into the plasma with constant density and the difference in impurity
transport between the co-injection case and the counter-injection case?
P.C. JOHNSON: Some of the observed effects can be explained using arguments based on neoclassical theory, for example the switching on and off of a temperature gradient screening term in the Ohmic high density cases. However, we do nothave profile data of sufficient accuracy for a detailed comparison of experiment withneoclassical theory. The results with central beam heating suggest that processes
other than neoclassical ones are important.

IAEA-CN-47/A-IV-5

STUDIES OF
EDGE PLASMA PHENOMENA IN T-10
K.F. ALEXANDER, L. DIETRICH, H. GROTE,
K. GUNTHER, J. LINGERTAT
Akademie der Wissenschaften der DDR,
Zentralinstitut fur Elektronenphysik,
Berlin, German Democratic Republic
E.L. BEREZOVSKU, A.V. CHANKIN, V.M. CHICHEROV,
S.E. LYSENKO, V.A. MEDVEDEV, A.Yu. PIGAROV,
V.I. PISTUNOVICH, S.V. POPOVICHEV, V. VERSHKOV
I.V. Kurchatov Institute of Atomic Energy,
Moscow, Union of Soviet Socialist Republics
Abstract
STUDIES OF EDGE PLASMA PHENOMENA IN T-10.
A scoop limiter equipped with Langmuir and heat flux probes in its scoop channels has been
operated in T-10. Results concerning the properties of the scoop plasma and the neutral gas pressurein the pump duct are reported together with a Monte Carlo simulation of the neutral component which
fairly well agrees with the observations.
A 3-D model of the scrape-off layer (SOL) in tokamaks has
been developed and evaluated with respect to plasma density, taking into account the multiplicity ofconnection lengths due to the complex interplay of the q-dependent magnetic topology and the particular
limiter geometry (rail limiter in T-10). It turns out that such a SOL is highly non-uniform in accordancewith experimental data from direct Langmuir probing, limiter thermography and a post mortem investigation of limiter damage. Results of a 2-D Monte Carlo calculation (geometry of T-10) of the distribution of neutrals originating from the limiter by recycling, considering the effect of backscattering, show
good agreement with corresponding measurements and, furthermore, can explain the discrepancy
between the neoclassical and the observed magnitude of ion heat transport: it is the neutrals that are
the carriers of the additional efflux of ion energy in question.

1. Further results of measurements with a scoop limiter
Fig.ia shows the sketch of a scoop limiter with instrurrentation which has been used in T-10 to study the scoop plasma andthe neutral gas in the pump duct in dependence on dischargeparameters. Details of the experiment and main results havealready been reported [ 1 ] .
In typical T-10 regimes the electron density, n e , in the
scoop channels on the electron and ion drift sides increases withgrowing rTe (Fig.i.b) while the electron temperature T e decreases
down to about 5 eV, which leads to an only weak n*e-dependence ofJ s , the sum of saturation currents of the probes
at the entrancesof both channels (0s<*« n e f7^), for rfe > 10 19 m"3. With the two
237

238

ALEXANDER et al.

pairs of probes at different minor radii (Ar = 1 crn) i t waspossible to determine the decay length A (e-foliing length forJj.) . Fig.1c,d shows a strong e-i-side asymmetry of A depending on
the direction of the magnetic field. A is always larger on thatside of the limiter where the plasma influx comes from the outerside of the torus, in accord with an argument given in [1] about
the preponderance of plasma escape to the outer torus side.
By comparison of the signals from the outer and inner probes
in each channel i t turned out that the electron temperature islargely constant along a channel. The mean value of the floating
potentials of the probes on both sides of the neutralizer plate
was not much different from the limiter potential, indicating
ambipolarity of the plasma flow in the channels. The heat flux
probes (thermocouples) at the neutralizer plate generally d i s played an average power input equal to 7J s T e /e.
As to the gas pressure, p^, in the pump duct, the finding [1]that Pj depends non-linearly on the plasma influx into the channel can be reproduced by Monte-Carlo calculations. The plates are
the source of the neutral gas. In the equilibrium the production
rate of neutrals balances the plasma flux into the channel$o = necsi where Cs is the sound velocity. The trajectories ofneutrals were simulated in two dimensions using the code DDC-83[4-,5], which takes into account the processes of dissociation,
charge exchange, stepwise ionization, backscattering, and reemission. The calculated dependence of the gas pressure in the pumpduct on the plasma influx $ o is shown in Fig.ie for two values ofthe plasma temperature. Especially at low temperatures (T = 5 eV)
the gas pressure rises considerably faster than # o above a c r i t i cal value of <$o. This can be qualitatively explained by the factthat a great deal of neutral gas near the plates and inside theduct consists of molecules produced by reemission. The moleculesinteract strongly with electrons even at relatively low temperatures. The atoms produced by dissociation of the molecules are
emitted isotropically and return in part to the plates or go into
the duct, thus increasing the pressure p^. At higher temperatures(T > 7 eV) ionization of atoms and dissociation of molecules with
ion production become important. Consequently, a smaller numberof neutrals relative to $ 0 can reach the pump duct and the pressure rise is not so steep.
More r e a l i s t i c a l l y , i t was assumed that the temperature
decreases with increasing $ o as observed experimentally. The
dependence of p<j on $ o was calculated with the assumption thatthe plasma temperature decreases 1from 320 eV to 4 eV 9whiletheplasma density increases from ^xiO ' .T;~ up to 1.5X10"* m' 3 . The
results (Fig.ie) fairly well match the experimental data.
2. The inherent structure of a r a i l - l i m i t e r scrape-off layer(SOL): a 3-D approach to plasma densityofIn view of growing experimental evidence for the occurrence
a non-uniform structure of the SOL in toksmaks [ 2 , 3 , 6 , 7 , 8 ] ,239

IAEA-CN-47/AIV-5|2[21mm g 3.2-•°

•ff Heat flux probes(graphite)

Langmuir probes
o graphite, •tungsten

"

/

•

°aa

- BtlJ

0

A

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BH3

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2.4

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1 22
3$ o ((10 2 3 slm 2 )FIG. 1. Results of scoop limber operation in T-10. (a) Limiter design (section in toroidal direction),(b) Electron density in the two scoop channels (O electron side, • ion side) and sum of saturation currents of channel entrance probes (ofor B,\ \JP, • for B,\ \JP) as a function of line averaged density
nt in the discharge (B, = 2.1-2.5 T, JP = 200-250 kA, a = 28 cm;. (c), (d) Radial decay lengthreferring to saturation current as a function of ne for normal (c) and reversed (d) magnetic field (opensymbols: electron side, closed symbols: ion side). Symbol shape indicates different discharge regimes(B, = 2.2-3.0 T, Jr = 200-350 kA, a = 26.5-28.0 cm), (e) Gas pressure in scoop limiter pump ductas a function of plasma influx at channel entrance. Experimental data and model calculations.

the symmetry-breaking effect of limiter configurations is ofparticular interest. While an aperture limiter leads to a relatively simple two-dimsnsional situation in cylindrical geometry[2], the case of a single r a i l limiter as used in T-10 can only
be mat by a 3-D approach with properly taking account of themagnetic topology in dependence on the safety factor g(i").
2.1. Theory
Fig.2 illustrates the problem. The "reference cross-section"(RCS) shown on the left is defined as the minor cross-section at
the toroidal position opposite to the limiter, ie 180° away fromi t . The different connection lengths, L(l , of flux tubes give rise
to a poloidal variation of plasma parameters in addition to theradial one, and the numerical treatment of cross-field transport
requires a sufficiently fine grid in polar coordinates. Alongeach field line, however, there will be no great variation within
half a revolution round the torus, ie up to the limiter section,
because the flux tubes usually have a length of several revolutions: L|, = L0W = 27rR0W. This may justify our taking the whole
toi*us circumference along a field line (««Lo), beginning andending at the limiter section, as the elementary interval Ax forthe treatment of parallel transport along the coordinate X. Thus,
each point in the RCS stands for a flux tube segment of length
Lo, and a complete flux tube is described by a sequence of points

ALEXANDER et al.

240

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IAEA-CN-47/A-IV-S

241

In t h e FiCS wich a r e found by means of t h e r o t a t i o n a l t r a n s f o r m a t i o n corresponding t o <j(f). On t h e o t h e r hand, a 2-D map of, eg,
t h e d e n s i t y i n t h e RCS, as shown i n Pig.2 through F i g . 4 , d e -scribes the entire 3-D field of density owing to the rotationaltransformation which has to be applied according to the toroidal
position under consideration.
Aiming at the structural aspects first of a l l , we have only
solved the continuity equation by time evolution to obtain thestationary plasma density n:

Lo = 2<rrtf 0 ,The velocity, U(X), of parallel flow can be expressed b y n ( x ) ,using an analytical solution to the momentum balance equation,
bun ,
bt
nm
with the boundary condition u(±L,,/2) = ± C S , setting Vxu = 0 (aswell as du/dt = 0). The occurrence of sufficiently large r,Sregions of uniform L w (Pig.2) warrants and requires this approximation by contrast with a traditional approach of complete neglect of the momentum sources. Additionally, we have made use ofT(x)<* ["(*)]'
with the standard choice j»; = 5/3 (adiabatic
expansion) and jfe = 1 (isothermic expansion). As to the stagnation point temperatures Ti0 and Te0 , equality and constancy withrespect to r and $ has been assumed for the sake of simplicity.
2.2. Results, discussion, and experimental evidence
Pig.2 visualizes what turns out upon assuming, for the sakeof demonstration, that^(r) is constant, (i) There is a fundamental symmetry which shortens the calculation, (ii) The distributions on both sides of the limiter are asymmetric with respect tor9", show mirror symmetry to each other, and are simply exchanged
by Bt-field inversion, (iii) In the limiter shadow area the average n and the associated effective decay length X (apparent SOL
thickness) are smaller than the corresponding total SOL averages
owing to a bias in statistical weight: each flux tube bundle
appears exactly once in the limiter area, but counts just N times
in the total average; however, the larger N the greater A and n.
Additionally, the general character of the density profile n(y)for each flux tube enhances this iirtportant effect. It implies a"self-protection" of the limiter (SOL acts as a "by-pass") andshows that probing the SOL by means of the limiter itself(thermography) may be misleading (accountability of power; see[10]). All these statements remain valid v/hen taking into accountthe actual dependence of q on r (cf the following examples).

ALEXANDER et al.

242

1

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243

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244

ALEXANDER et al.

The high degree of non-uniformity of the SOL exhibited by thedensity map in Fig.3 has been confirmed by a Langmuir probe experiment on T-10 (different probe positions by f i e l d i n v e r s i o n ) .
The only adjustable parameter in the calculation i s %.o. The valuethat proved appropriate (4.5 nm) corresponds to £ r = 0.2 m z /s for
T = 8.5 eV, which just equals the Bohm coefficient (B t = 2.5 T ) .
In e a r l i e r reports [ 6 , 7 ] on thermographic measurements on
T-10 r a i l limiters we have pointed out the existence of a s h i f t
between the temperature peaks on both sides of the l i m i t e r . Thisphenomenon has now been studied systematically by evaluating thetemperature p r o f i l e s along s t r a i g h t lines p a r a l l e l to the ridgeof the span-roof-like l i m i t e r surface [7] on both the electron
and ion d r i f t s i d e s . In the minor cross-section these l i n e s arehorizontal chords whose distance from the l i m i t e r edge i s called"depth". Fig.4 shows an example (a) compared with the t h e o r e t i c a l
prediction ( b ) . Although both data are i n t r i n s i c a l l y inconmensura t e , the s i m i l a r i t y i s s t r i k i n g (a smaller value of Ao has beenused to simulate the influence of temperature decay). There i s an
obvious mutual mirror symmetry of the asymmetric peaks.
In a uniform exponential-decay SOL the profiles in question
should read 4T(z) oc exp[-(z-z o ) /2aA]. We have used t h i s expression complemented v/ith a t h i r d - o r d e r asymmetry term to f i t theexperimental profiles as v/ell as the t h e o r e t i c a l ones, mainly t oobtain a unique measure of the s h i f t : the difference of zo values
for both of the l i m i t e r s i d e s . Fig.4c,d shows an example together
v/ith the corresponding density map ( e ) . The e n t i r e set of data
examined (q(a) between 2 and 3.3) has shown t h a t the theory in
any case yields the correct sign of the s h i f t for a small depth(always positive for normal field direction and vice versa) and,thus, also accounts for previous r e s u l t s of a post-mortem i n vestigation of a T-10 r a i l limiter [11] which has shown the samee-i-side shift of limiter load by ablation and cracks.
I t can be concluded that the phenomenon of asymmetry in question i s actually due to the kind of SOL structure our modelaccounts for. Remaining discrepancies can c l e a r l y be a t t r i b u t e dto the experimental uncertainty of q(a) on the one hand and theshortcomings of the model on the other hand: ( i ) A complete twofluid theory in terms of n, U, T'L, T e is needed, ( i i ) True t o r o idal geometry must be used, ( i i i ) Cross-field motions (eg, ExB)
nust be considered, (iv) The effect of loop voltage should be
included. None of these requirements i s too d i f f i c u l t to meet.
3. Toroidally localized neutral p a r t i c l e d i s t r i b u t i o n s
The limiter as the main sink for the plasma efflux i s at thesame time the main source of neutrals by recycling. This leads tostrong inhomogeneities in the s p a t i a l d i s t r i b u t i o n of neutrals
both in the plasma column and in the SOL. Taking into account theactual geometry of T-10, we have simulated the n e u t r a l p a r t i c l etransport in the v i c i n i t y of the poloidal aperture limiter by aMonte-Carlo procedure. Since a large percentage of neutral atoms

IAEA-CN-47/A-IV-5

245

F/G. 5. Neutral particle density in T-10. (a) Calculated radial distributions at position of poloidal
aperture limiter. Na: limiter produced atoms (dashed line: analytical solution); nm: molecules from gas
puff; N™: molecules from gas puff; N™: atoms produced by dissociation of gas molecules,(b) Experimentally determined radial distribution of atoms. ND: from Dp radiation in limiter section(N: calculated for limiter produced neutrals). N": from charge exchange spectra 90° away fromlimiter section, (c) Experimental temperature and density profiles typical for T-10 Ohmic discharges,(d) Calculated contributions to radial ion energy efflux due to twice the neoclassical ion heat conduction(PJ and charge exchange neutrals (PcJ, which sum up to total flux corresponding to net energy input
in ion component (Pei).

leaving the limiter are backscattered neutralized ions, the penetration depth of these neutrals into the plasma core is considerably larger than i t is in the case of neutrals originating fromcold gas molecules desorbed from the wall or puffed into thechamber. Results for both cases ("fast" neutral atom source and"slow" molecular source normalized to the sane global particleflux), referring to one of the standard regimes of T-10 [9], aregiven in Fig.5a. The Monte-Carlo calculations were used to fit
neutral density profiles obtained with the one-dimensional modelof consecutive generations [12] by varying the energy of the
primary neutrals. An energy of 20 eV gives a good overall fit to
the Monte-Carlo profile for limiter-produced neutral atoms. This
profile is confirmed by comparison with experimental data for thelirniter section (Fig.5b), while the low values ireasured far fromthe latter are in keeping with the calculated density peak at thelimiter to decay steeply in toroidal direction.
4. Ion heat transport by charge-exchange with neutrals
An important effect, ordinarily not considered in energy
transport calculations, is the additional heat conduction causedby the charge-exchange interaction of neutral atoms with plasmaions. This process can contribute substantially to the ion heattransport in the outer region of the plasma column, considering
the enhanced penetration depth of fast primary neutrals comingfrom the limiter via recycling. A thorough investigation of ion
temperature measurements in the presence of kinetic convective

246

ALEXANDER et al.

ripple transport of ions [13] revealed that the ion temperature
profiles in T-10 have a more pronounced peak than thought before.
The ion heat conductivity derived from these profiles deviates
strongly from the neoclassical scaling for r > 0.7O. . This behaviour can be a t t r i b u t e d to the additional heat conduction by
neutral atoms.
The plasma energy balance has been calculated for the e x p e r i mental data in Fig.5c. Assuming neoclassical ion heat conductivi t y multiplied by Kan = 2 and including the heat transport by
neutral atoms based on the above-mentioned Monte-Carlo c a l c u l a t i o n s , one obtains the curves in Pig.5d which shows that theapparent high anomality of the ion heat conductivity for r > a / 2
is actually heat conduction by n e u t r a l s .

REFERENCES
[1]
[2]
[3]
[4]
[5]
[6]
[7]
[8]
[9]
[10]
[11]
[12]
[13]

VERSHKOV, V.A., GRASHIN, S.A., CHANKIN, A.V., in Plasma-Surface Interactions inControlled Fusion Devices (Proc. 7th Int. Conf. Princeton, NJ, 1986), Report 7.5 (in press).
SAMM, U., et al., ibid., Report 1.6 (in press).
WAN, A.S., et al., ibid., Report 1.3 (in press).
IGITKHANOV, Yu.L., et al., Dokl. Akad. Nauk SSSR 278 (1984) 338.
KUKUSHKIN, A.S., PIGAROV, A.Yu., USSR Contribution to the INTOR Phase-Two-A
Workshop, Group A, p. 20.
LINGERTAT, J., et al., in Plasma Physics and Controlled Nuclear Fusion Research 1984 (Proc.
10th Int. Conf. London, 1984), Vol. 1, IAEA, Vienna (1985) 265.
GUNTHER, K., et al., in Controlled Fusion and Plasma Physics (Proc. 12th Europ. Conf.
Budapest, 1985), Vol. 2, European Physical Society (1985) 567.
BOGEN, P., et al., ibid., p. 527.
ALIKAEV, V.V., et al., ibid., p. 160.
LINGERTAT, J., et al., ibid, p. 559.
HILDEBRANDT, D., et al., in Controlled Fusion and Plasma Physics (Proc. 10th Europ. Conf.
Moscow, 1981), Vol. 1, European Physical Society (1981) J-12.
DNESTROVSKY, Yu.N., KOSTOMAROV, D.P., Numerical Simulation of Plasmas, SpringerVerlag, Berlin, Heidelberg (1986).
BEREZOVSKIJ, E X . , et al., Nucl. Fusion 23 (1983) 1575.

DISCUSSION
H. WOBIG: The limiter introduces inhomogeneities not only into the density
but also into the electric potential. This may lead to additional convective motion in
the vicinity of the limiter. Did you take into account this effect?
K.F. ALEXANDER: No, we did not take this effect into account in the present
work, but we will do this in future. In the experiments with the scoop limiter, differences in the floating potential of several volts between the electron and ion drift sides
were observed.

IAEA-CN-47/A-IV-5

247

R.J. GOLDSTON: In your model of the scrape-off plasma you don't seem toneed asymmetric sources to explain an asymmetric scrape-off plasma. In some auxiliary heating experiments, asymmetries have been attributed to the physical loss
mechanisms. Could you perhaps comment on this and tell us what has been observed
during ECH on T-10?
K.F. ALEXANDER: Our model assumes a poloidally and toroidally
homogeneous plasma source feeding the scrape-off layers. The asymmetries
produced by the rail limiter discussed here are only due to the topology of open fieldlines of different connection lengths within the scrape-off layers.
During ECH, additional asymmetries are certainly observed, for instance inlimiter thermography. I would refer you in this connection to the paper by
J. Lingertat which he presented at the 1984 Conference1.

1

LINGERTAT, J., et al., in Plasma Physics and Controlled Nuclear Fusion Research 1984
(Proc. 10th Int. Conf. London, 1984), Vol. 1, IAEA, Vienna (1985) 265.

IAEA-CN-47/A-IV-6

ALT-I PUMP LIMITER BEHAVIOUR
AND EDGE PLASMA FLOWSDURING BIASING AND ICRF HEATING
IN THE TEXTOR TOKAMAK
R.W. CONN, D.M. GOEBEL, W.K. LEUNG, G.A. CAMPBELL,
B. LaBOMBARD, A.K. PRINJA

Center for Plasma Physics and Fusion Engineering,
University of California1,
Los Angeles, California, United States of America
W.L. HSU, G.J. THOMAS

Sandia National Laboratories2,
Livermore, California, United States of America
K.H. DIPPEL,K.H. FINKEN,
R. REITER,
G.H. WOLF,
H.L. BAY,
G. BERTSCHINGER, W. BIEGER, P. BOGEN, W. BRUSSAU, Y. CAO,
H.G. ESSER, G. FUCHS, B. GIESEN, E. GRAFFMANN, H. HARTWIG,
E. HINTZ, F. HOENEN, K. HOTHKER, A. KALECK, H. KEVER, L. KONEN,
M. KORTEN, L. LI, Y.T. LIE, A. POSPIESZCZYK, A. ROGISTER, G. ROSS,
D. RUSBULDT, U. SAMM, J. SCHLUTER, B. SCHWEER, H. SOLTWISCH,
F. WAELBROECK,
G. WAIDMANN,
P. WIENHOLD,
J. WINTER,
R. YAMAUCHI

Euratom-KFA Association,
Institut fur Plasmaphysik,
Kernforschungsanlage Jiilich GmbH,
Jiilich, Federal Republic of Germany
T. DELVIGNE, P. DESCAMPS, F. DURODIE, M. JADOUL, R. KOCH,
D. LEBEAU, A.M. MESSIAEN, D.I.C. PEARSON, P.E. VANDENPLAS,
R. VAN NIEUWENHOVE,
G. VAN OOST,G. VAN WASSENHOVE,
R.R. WEYNANTS

Laboratoire de physique des plasmas —Laboratorium voor Plasmafysica,
Association Euratom-Etat Beige —Associatie Euratom-Belgische Staat,
Ecole royale militaire — Koninklijke Militaire School,
Brussels, Belgium

12Work supported by the United States Department of Energy, under Contract
No. DE-RG03-85-ER51069.
Work supported by the United States Department of Energy, under Contract
No. DE-AC04-76DP00789.

249

250

CONN et al.
Abstract

ALT-I PUMP LIMITER BEHAVIOUR AND EDGE PLASMA FLOWS DURING BIASING AND
ICRF HEATING IN THE TEXTOR TOKAMAK.
Experiments have been performed in the TEXTOR tokamak, using the ALT-I pump limiter withstrong electrical biasing and ICRF heating. With negative biasing (up to - 4 5 0 V) of ALT-I, the flux
to ALT-I increases moderately (about 10%) at the front face of the limiter but very significantly (upto 250%) in the throat. The pressure and removal rate increase by a factor of two or more. Simultaneously, total D o light from the surface of other recessed limiters and the liner decreases by a factor ofup to two. Some shots show no change in the loop voltage signal, while others show an increase. With
positive bias (up to +450 V), generally opposite effects occur. The flux and pressure at ALT-I decrease,
while D o light at the liner increases by a factor of two. The O VI signal likewise jumps and the loop
voltage increases by 80% in 0.5 s. The net current drawn to the power supply ends at 120 A for negative
bias and about 220 A for positive bias. With up to 1.6 MW of ICRF heating, negative biasing of ALT-I
causes a decrease in recycling at the antenna. However, no change in plasma energy confinement scaling
using ALT-I relative to previously reported results was found in a systematic study.

1.

INTRODUCTION

Core plasma density control and heat removal are primaryreasons for investigating pump limlters[1-3] in fusion
devices.
As part of the ALT-I program [4] in the TEXTOR
tokamak [5,6], pump limiters of several designs and materials
have been used in experiments to explore the physics of plasma
exhaust [7-10], impurity and helium flows [10], the nature of
the plasma edge [11], the performance of different limiter
materials [12,13], and pump limiter operation during ICRF
heating [14].
In this paper, we report experiments in which the ALT-I
pump limiter is electrically biased both positively (to +450V)
and negatively (to -450V) with respect to the other limiters
and the liner.
Early work on the role of DC radial electricfields in plasma transport was reported by Taylor et al. [15]and recent experiments using limiter biasing were performed in
the PDX [16] and TEXT [17] tokamaks. The experiments in TEXTOR
were performed during ohmic and ICRF heated (up to 1.6MW)
discharges.
General ICRF heating studies in TEXTOR arereported in references [18] and [19], We describe here studies
of ICRF heating with ALT-I operated as a pump liraiter or, with
its exhaust channels closed, as a regular limiter.
Limiter
biasing during ICRF heating will also be discussed.

2.

LIMITER BIASING EXPERIMENTS

2.1.

EXPERIMENTAL ARRANGEMENT AND DIAGNOSTICS

The
generalTEXTOR
diagnostics are described

experimental
arrangement
in references [5 f 6 ] ,
The

and
ALT-I

251

IAEA-CN-47/A-IV-6

IRdet.

CRYOPUMPS (3)7000 l/s

QUAD.
MASS. SPECT.
TMP

FIG. 1. ALT-I pump limiter, showing position of Langmuir probes, Da viewing positions, the pumping
chamber, and position of pressure measurements.

llmlter has the design first reported In ref. [4] and is madeof a carbon-carbon composite weave. The design and location ofvarious diagnostics are given in Fig. 1. For the experiments
reported here, ALT-I limits the plasma to a radius of 42 cm.
Approximately 90° toroidally from ALT-I are three main limiter
segments made of carbon, each subtending approximately 30°
pololdally and located top, bottom and midplane outside. They
are retracted to a minor radius of 48 cm, i..e. 6 cm radiallybeyond ALT-I. The ICRP antennas are two half-turn loops located
40° toroidally from ALT-I in the opposite direction of the main
limiters. The protective carbon limiters of the antennas form afull poloidal ring beginning at 48.8 cm. The liner is at 55 cm
and the tokamak is generally conditioned by carbonization[20].
Experiments were also performed with ALT-I at otherradial locations, but we restrict the discussion here to thegeometrical arrangement just described.
Key diagnostics which will be referred to in the paper
include D - monitors viewing the liner, the ICRF ring Hmiters,
and the upper main limiter.
In addition, D - emission ismonitored at the face of the ALT-I head and at the deflector

252

CONN et al.

ALT-I NEGATIVE BIAS (-450V)

ALT-I POSITIVE BIAS (450V)

23707
(a)
A

•

A

2

•

2 •

/

f

BIAS

23707
(c)
0.8-

0.8^\

J V - - 2*f

X

0.4 -

0.4"

BIAS
0

I
TIME(s)FIG. 2. Core plasma density, pressure in the pump limiter and flux into the pump limiter throat, asa function of time. Negative or positive bias to ALT-I is applied from 1 s to 1.5 s. The dashed curve
indicates signal values without bias.

253

IAEA-CN-47/A-IV-6

ALT-I POSITIYE BIAS (450V)

ALT-I NE.OATIVE BIAS (-4S0V)
10

23707
<•
D-ALPHAHEAD OF
ALT

2

A i

1 .
"

23707
D-ALPHAMAIN
LIMITER(f)
23693
D-ALPHATHROAT
OF ALT

A.U.

(b>
23707
D-ALPHATHROAT
OF ALT

23693 (g)D-ALPHAMAIN
LIMITER

(c)
• 2
E>

BIAS

BIAS

3

FIG. 3. Da signals viewing various locations in ALT-I, at the main limiters, and at the liner, as aJunction of time. Bias is applied from 1 s to 1.5 s. The dashed lines indicate the signal values without
bias.

CONN et al.

254

ALT-I NEGATIVE BIAS (-450V)
23707

a,LT-l POSITIVE BIAS (450V)
(a)

2

2

I1

1

I.

o

23693 (e)H»WT-" —*•

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BIAS

BIAS

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160

F/G. 4. Loop voltage, O VI and C V signals, and energy confinement time as a function of time.
Bias is applied from 1 s to 1.5 s. The dashed lines indicate signal values without bias.

IAEA-CN-47/A-IV-6

255

plate where plasma terminates inside the pump limlter. Plasma
fluxes are measured with Langmuir probes and the neutralpressure is measured in the pump limlter chamber.
2.2,

RESULTS FOR OHMIC DISCHARGESKey results from various diagnostics are summarized in
Figs 2, 3 and 4, where signals are plotted versus time during
a discharge. Generally, the gas puff is on until 0.8 s, thebias is applied from I s to 1.5 s, the plasma current isconstant from 0.5 s to 1.6 s, and the discharge generally
terminates in a "soft landing" at 3.0 to 3.5 s [5,6].
Fig. 2 shows that the central line-averaged density
increases at bias turn-on for both positive and negativebias. The relative increase is largest with positive bias andin each casethe increase appears to be due to impurity
introducton.
(See Fig. 4, to be discussed shortly.)For
negative bias, the flux into the throat of the pump liralter
increases sharply at bias turn-on and both the pressure signal
and the plasma removal rate in the pump limiter increase by 50
to 100%. (See Fig. 2.) By contrast, we see from Fig. 3 thatthe D - signals at the other llmiters and particularly at theliner decrease.
The D - signal at the front face of ALT-I
increases, but by a smaller amount than the signals in thethroat. The increase in the CV signal shown in Fig. 4 appears
to be due to carbon from the limiter and correlates with thedensity rise.
These results indicate that during negative biasing, ion
flow is preferentially to ALT-I and, in particular, into thethroat.
A measurement by Pospleszczyk and Ross [21] of thedensity profile in the plasma boundary layer, taken at 1.4 s
into the discharge and while bias is on, is shown in Fig. 5.
During negative bias, a density bump occurs between 45 and 48
cm, suggesting that cross-field transport is inhibited as theplasma approaches the main limlters (48 cm). The main liraiters
are the first objects radially beyond ALT-I and are positive
relative to it. The impedance to cross-field flow appears to
be related to a radial potential barrier.
Generally, opposite effects occur during positive bias.
Following turn-on of positive bias, decreases are observed in
the Dfl- signals at the ALT-I face and at the deflector plate.
No bump in the boundary layer density occurs (see Fig. 5) andthe density in front of the main limiter at 48 cm is somewhat
lower than with no bias. The D - signals at the other limiters
do not change much, but four of the five D - monitors at theliner show increases of a factor of 2 or more. We do find thatthe liner D - signals strongly depend on toroidal location.
Some signal changes are illustrated in Fig. 6, where A (signal)
corresponds to the signal value just after bias minus the value
just prior to bias, normalized to the value just before bias.

256

CONN et al.

o
>
tn

a

44

48
RADIUS (cm)52

FIG. 5. Edge plasma density profile at 1.4 s into discharges, with positive bias, negative bias orno bias applied to ALT-I. The radial position of ALT-I and the main limiters are shown.

The increased flux to the liner is correlated with aprompt increase in the OVI and CV signals in the plasma edge,
indicating that there is a flux-induced release of oxygen andcarbon.
The loop voltage shown in Fig. 4 increases linearlyover the 0.5 s positive bias time, suggesting impurity
accumulation. The value of global energy confinement,
T , correspondingly decreases, apparently due to increased
levels of low Z radiation.
For negative bias, the oxygen
signal increases slightly, but the loop voltage and confinement
time show little change. We deduce that the source of oxygen
impurity is desorption from the liner, not from ALT-I. During
negative bias, the source of carbon must be ALT-I.
The I-V trace of current and voltage at the bias power
supply, shown in Fig. 6, indicates that the current limits to a
net value of 110A for strong negative bias and about 220A forstrong positive bias.
The floating potential is 50-90V
relative to the wall.
The 110-120A current level during
negative bias is within a factor of 2 to 3 of the estimated ion
current diffusing from the core plasma for this particluar
arrangement of liraiters.
This estimate assumes T is 0.1s,
consistent with results reported earlier [22].
A possible picture to explain the edge plasma flows during
biasing is to view the plasma potential as following thepotential applied to ALT.
(This is similar to the observation
reported by Taylor et al. [15].) The system then effectively
acts as a double probe of unequal areas, with ALT-I as one
electrode and the liner as the other. The net current in the
power supply circuit is determined by the plasma potential in
the scrape-off layer plasma relative to the two electrodes.

257

1AEA-CN-47/A-IV-6

(a)

•BIAS
•CURRENT (A)

200

•

100
600

V——^

BIAS
VOLTAGE (V)

200

-200

-400

•

400

600

-100

BIAS
VOLTAGE (V)
200

-400

-0.2 "

-0.4

•

400

•

600

•

A (SIGNALVSIGNAL
ALT HEAD
D-ALPHA• A (SIGNAL)/SIGNALALT THROAT
D-ALPHA

(d)

0.8

•
• •
-400

-400

- p

0.4

' A(SIGNAL)/SIGNAL•D-ALPHA-MAINLIMITER
BIAS
VOLTAGE (V)

•

-200

200

•400

600

-0.4

1.2

(e)

-600

.-400

•

• •

-20^

0.8

' L (SIGNALVSIGNAL
D-ALPHA
•UNER

0.4

•
200

.

1

BIAS
VOLTAGE (V)
400

600

'

FIG. 6. (a) Current through the power supply versus bias voltage applied to ALT-I.
(b, c, d, e) Relative changes in Da signals at various locations. A(signal)/signal is the value justafter turn-on of bias, minus the value just prior to bias, normalized to the value before bias.

258

CONN et al.

During negative bias of ALT, the plasma potential decreases
relative to the wall, creating a potential barrier between theplasma edge and the main limiters at 48 cm and the liner at 55
cm.
The density bump reported in Fig. 5, and the decrease in
the D - signals at the other liraiters and liner are consistent
with this potential barrier.
For positive bias, the plasma
potential increases relative to the wall, but is now below the
ALT potential.
The 220A of current in the circuit isdetermined by the amount of current that flows across themagnetic field in the edge and reaches the main limiters andliner, and not by the random electron flux in the edge near
ALT.
Overall, we conclude that plasma flows in the boundary
are significantly modified by plasma potential changes, without
requiring excessive currents in the power supply. In general,
limiter biasing is a specific technique to alter wall recycling
in tokamaks and a means to alter the plasma flux to various
limiters and the liner.

3.

PUMP LIMITER AND ICRF HEATINGFirst results on ALT-I behavior during ion cyclotron
resonance heating (ICRF) [18,19] were reported earlier [14],Here, we report a systematic comparison of ICRF heating using
ALT-I as a pump limiter (PL mode) or, with its throat closed,
as a regular limiter (RL-mode).
Particular emphasis is given
to the influence, if any, of pump limiter operation on coreenergy confinement time and of negative limiter biasing.
The analysis of energy confinement time is that of
Messiaen et al. [18], in which the global energy confinement
time, T_, is determined from
E
E

dia

where E^j^ is the plasma energy determined from the diamagnetic
loop signal, P Q H is the ohmic heating power during the ICRF
heating, and P^p is the RF power radiated into the tokamak
vessel.
A reasonable estimate of the RF power coupled to theplasma is 85% of the RF power delivered by the generator.
Results are given in Fig. 7, where T is plotted versus thetotal heating power, PQ1j + P I C R H » evaluated during a typical
0.5 second RF heating pulse. The open circles are values ofT F for ALT-I in the PL-mode, while solid circles denote T for
RE-mode operation. No systematic difference is found. This isso, even though operation in the PL-mode reduces the degree ofcore density rise which normally accompanies the ICRH pulse.
Though not shown here, T- u follows an Alcator-like scaling with

259

IAEA-CN-47/AIV-6
1

i

1

'

ALT-MODEo PUMP LIMITER

100

A REGULAR LIMITER
\

80

\

\

\
\

60

-

40 -

20

-

OHMIC POWER
w/o RFn0.0. 1, .

i

0.4

0.8

1.2

1.6

2.0

P T0T (MW)FIG. 7, Global energy confinement time as a Junction of total input power, determined midway throughan ICRF power pulse. Open circles are for ALT-I as a pump limiter and triangles are for ALT-I as aregular limiter (with its throat closed).

density. No difference in !-„ scaling is found when operating
ALT-I in either the PL- or RE-modes.
RE-m
Negative limiter biasing during ICRH heating is found to
reduce the recycling plasma flux at the RF antenna and,
operationally, fewer breakdown/interruptions in the antennas areobserved. Nevertheless, no differences in global T_, are foundduring ICRF heating and limiter biasing.

4.

CONCLUSIONS

Plasma edge flows can be strongly affected by limiter
biasing. Plasma exhaust is increased (by a factor of 2 in thiswork) during negative bias, while flows to the liner, otherliraiters and the RF antennas generally decrease.
No strongchange in energy confinement is observed.
The strong exhaust
increase follows from a bump in the edge density profile which
may be due to a radial potential barrier.
Positive bias
increases flow to the walls, decreases flow to the pumplimiter, decreases plasma exhaust, and appears to induce
impurity accumulation.
No bump is found in the edge radial
density profile.

260

CONN et al.

During ICRK heating, negative biasing of ALT-I causes a
decrease in recycling at the RF antenna and results in fewer
antenna breakdowns/Interruptions.
However, no change in coreplasma confinement is observed, relative to earlier results[18,19], when using ALT-I as a regular or a pump limiter.

ACKNOWLEDGEMENT
The authors gratefully acknowledge the help of the TEXTOR
operations team and the support staff of UCLA, Sandia Labs andthe ERM, Brussels.

REFERENCES[1] MIODUSZEWSKl, P.J., J. Nucl. Mater. 111&112 (1982) 253.
[2] CONN, R.W., J. Nucl. Mater. 128&129 (1984) 407.
[3] GOEBEL, D.M., Fusion Technol. 10 (1986) 761.
[4] CONN, R.W., et al., in Fusion Technology 1982 (Proc. 12th Symp. Jiilich, 1982), Vol. 1,
Pergamon Press, Oxford (1983) 497.
[5] WAIDMANN, G., BAY, H.L., BERTSCHINGER, G., et al., in Plasma Physics and Controlled
Nuclear Fusion Research 1984 (Proc. 10th Int. Conf. London, 1984), Vol. 1, IAEA, Vienna(1985) 193.
[6] WOLF, G., et al., Plasma Phys. Controll. Fusion 28 (1986) 1413.
[7] PONTAU, A.E., et al., J. Nucl. Mater. 128&129 (1984) 434.
[8] CAMPBELL, G.A., et al., The Influence of Plasma-Neutral Interactions on the Behaviour ofa Pump Limiter, Rep. PPG-958, Univ. of California, Los Angeles (1986).
[9] FINKEN, K.H., et al., J. Nucl. Mater. 145&146 (1987).
[10] DIPPEL, K.H., FINKEN, K.H., GUTHRIE, S.E., et al., in Plasma Physics and Controlled
Nuclear Fusion Research 1984 (Proc. 10th Int. Conf. London, 1984), Vol. 1, IAEA, Vienna(1985)249.
[11] GOEBEL, D.M., et al., Langmuir Probe Measurements in the TEXTOR Tokamak during ALT-I
Pump Limiter Experiments, Rep. PPG-945, Univ. of California, Los Angeles (1986).
[12] McGRATH, R.T., et al., J. Nucl. Mater. 145&146 (1987).
[13] DOYLE, B.L., McGRATH, R.T., PONTAU, A.E., Nucl. Instrum. Methods Phys. Res.,
Sect. B (1987).
[14] LEUNG, K., et al., Rep. PPG-968, Univ. of California, Los Angeles (1986).
[15] TAYLOR, R.J., JAMES, B.W., JIN, S.X., et al., in Plasma Physics and Controlled Nuclear
Fusion Research 1982 (Proc. 9th Int. Conf. Baltimore, 1982), Vol. 3, IAEA, Vienna (1983) 251.
[16] SHIMADA, M., et al., Bull. Am. Phys. Soc. 30 (1985) 1439.
[17] PHILLIPS, P.E., et al., J. Nucl. Mater. 145&146 (1987).
[18] MESSIAEN, A., et al., Plasma Phys. Controll. Fusion 28 (1986) 71.
[19] VANDENPLAS, P.E., DELVIGNE, T., DESCAMPS, P., et al., Paper IAEA-CN-47/F-I-4,
these Proceedings, Vol. 1.
[20] WINTER, J., J. Nucl. Mater. 145&146 (1987).
[21] POSPIESZCZYK, A., ROSS, G., TEXTOR Group, private communication.
[22] SAMM, U., et al., in Fusion Engineering (Proc. 11th Symp. Austin, TX, 1985), Vol. 1,
IEEE-CAT. No. CH2251-7, IEEE, New York (1986) 66.

IAEA-CN-47/A-IV-6

261

DISCUSSION
R.J. HAWRYLUK: In your experiments, the confinement time in ICRF heateddischarges was unaffected by the pump limiter. Did limiter biasing affect the confinement time?
R.W. CONN: At present, we have no evidence of improved energy confinement using limiter biasing with ICRF heating. However, no systematic experimentson this point have yet been undertaken.
A. GIBSON: On the basis of your results, would you care to offer a prognosis
for the future of pump limiters?
R.W. CONN: Pump limiters have been shown in earlier work to be effective
in controlling plasma density and also to have efficient exhaust capability. The key
issue ultimately will be to achieve improved confinement, such as is now possible
with divertors. The results presented in this paper demonstrate that plasma edge flows
are strongly influenced by limiter biasing. This can be combined with material conditioning and pump limiter exhaust capability to produce low recycling operation — a
key to achieving better confinement. Also, we have heard at this Conference that such
improved confinement (H-mode) with a limiter has been achieved on JFT-2M. Given
the inherent simplicity and low cost of limiters, the future for pump limiters is
reasonably bright.

IAEA-CN-47/A-V-1

CURRENT DENSITY PROFILESIN THE TEXTOR TOKAMAK
H. SOLTWISCH1, W. STODIEK2,
J. MANICKAM2, J. SCHLUTER1
1

Euratom-KFA Association,
Institut fur Plasmaphysik derKernforschungsanlage Julich GmbH,
Jiilich, Federal Republic of Germany

2

Princeton Plasma Physics Laboratory,
Princeton University,
Princeton, New Jersey,
United States of America
Abstract

CURRENT DENSITY PROFILES IN THE TEXTOR TOKAMAK.
Current density and hence safety factor profiles have been measured by the Faraday rotation
method in the TEXTOR tokamak. The profiles of sawtooth dominated discharges appear marginallystable for all modes and have stationary values of q on axis less than unity, in some conditions as low
as 2/3. The profiles assume a roughly universal shape which is determined in the central region by thestability of the m = 1 tearing mode and in the outer region by the general heat transport and thermal
stability of the plasma column.

1.

INTRODUCTION

The tokamak magnetic structure for low |5 plasmas is determined by "external" parameters: magnetic axis major radius, toroidal magnetic field strength, total plasma current and an "internal" parameter: the poloidal field distribution, adjusted tofulfil macroscopically MHD (at least near) stable equilibrium.
In general, the radial temperature profile is determined bybalancing the ohmic heating with the plasma heat loss. The thermal instability will cause the current channel to shrink if theplasma edge is set by a "virtual" limiter. This shrinking process is hatted, leading to a stationary state, through clampingof the local current density by additional local tra/isport introduced by sawtooth activity near the plasma axis. The sawtooth activity is connected with instability of the m=l, n=l internal tearing mode, originating near the q=l radius.
263

264

SOLTWISCH et al.

Under steady state conditions with sawtooth activity theradius of the q=l surface and the current density profile (the"internal" tokamak parameter) are thus determined by the dependence of the stability of the m=l, n=l tearing mode on the current density profile.
The experimental investigation of this relationship is the
subject of this paper.
The recent development on TEXTOR Tokamak of the Faraday rotation method
for poloidal magnetic field measurements has
enabled us to follow the evolution of the current density
distribution and in particular to determine the quasi-stationary
profiles, especially near the plasma axis, for various "external" tokamak parameters. A direct measurement of the current
density in this area is preferable to, for instance, inference
from plasma conductivity, because of the intricacies of the sawtooth activity.
The TEXTOR tokamak plasmas considered are ohmically heated,circular with aspect ratio of 3.5-4, and have an effective chargeof 1.5-3. The "external" parameters varied are toroidal field
and total plasma current.
Theoretically, tearing mode stability is determined by analyzing the observed profiles for m > l modes with a cylindricaltearing mode code, which is sufficient for this problem. The
m=l, n=l mode, ideally unstable in cylindrical geometry,? requires toroidal analysis, which has been carried out with the
Princeton resistive PEST code.
The experimental results indicate that the plasma current
density increases and the column radius decreases in time, until
the q=l radius is well inside the plasma column. As a generaltrend, this has been derived from conductivity measurements with
soft x-ray techniques permitting simultaneously the determination of electron temperature and Ze.ff profiles. A direct determination of greater accuracy was carried out with the Faraday
rotation method. Under these conditions the value of q on axis
is measured to be considerably less than unity.
Such profiles have been found experimentally stable, thatis, no modes of any m or sawtooth activity are observed. In manycases these profiles are hindered from further shrinkage by centrally located impurities. If the profile shrinks further (forinstance by loss of impurities from the center), predominantly
the m=l mode, with accompanying sawtooth activity, develops
periodically, and now the sawtooth mechanism, on the average,
becomes the controlling mechanism for the central conductivity.
The exact theoretical criteria for the onset of the m=l instability, as function of the q-profile and aspect ratio,are not

IAEA-CN-47/A-V-1

265

yet clear in d e t a i l . 9 S t a b i l i t y i s favored by a plateau i n thecurrent density d i s t r i b u t i o n , suggesting t h a t , in t o r o i d a l geometry, s t a b i l i t y can be obtained through nonlinear saturationand p r o f i l e m o d i f i c a t i o n . This is supported by the observationthat m=l, n=l modes exist with s t a t i o n a r y , large amplitudes f o rhundreds of plasma r o t a t i o n periods.
For p r o f i l e s with q=2.1
at the l i m i t e r and q(0)=0.66, these plateaus could be resolvedin our experiments, due t o t h e i r dimensions r e l a t i v e t o the apparatus. We f i n d these p r o f i l e s to be near marginal s t a b i l i t y ,using the PEST code, in agreement with the experiments.
We observe that in stationary conditions in sawtoothingdischarges, current density and q - p r o f i l e assume a unique shape.
This can he expressed in reduced coordinates (depending only on
the " e x t e r n a l " tokamak parameters R, By and I ) as a result of
the c r i t i c a l e f f e c t of the m=l t e a r i n g mode on t r a n s p o r t :The universal p r o f i l e i s overall tearing mode marginallyin the c y l i n d r i c a l approximation.

stableThus the p r o f i l e s may be interpreted as f i n e - s c a l e heat
conduction p r o f i l e s , i n t r i n s i c a l l y marginally stable to higher-mt e a r i n g modes, clamped only by the very c r i t i c a l m=l mode,through sawtooth a c t i v i t y , t o the external tokamak parameters.
With the q=l radius f i x e d , the t o t a l p r o f i l e and thus the " i n t e r n a l " tokamak parameter i s then determined.

2.

ACCURACY OF THE FARADAY ROTATION METHOD TO DETERMINE THE
VALUE OF q AND j AT THE MAGNETIC AXIS

To prove experimentally the s t a b i l i t y of the m=l tearing
mode, i t is necessary t o show that s t a b i l i t y exists f o r q(0)considerably less than one. This requires s u f f i c i e n t spatial r e solution of the polarimeter diagnostic inside the q=l radius t oderive the value of q on axis by Abel inversion. For larger q=l
r a d i i t h i s i s possible in TEXTOR with the e x i s t i n g probing beamspacing of 10 cm. For plasmas with high q at the l i m i t e r and asmall q=l radius, the e f f e c t i v e number of probing beams can beincreased by s l o w l y moving the plasma r e l a t i v e t o the beamsystem by about _+_ 1/3 beam distance. Even f o r small q=l r a d i i wef i n d generally agreement between the two methods. At present obtainable accuracies f o r q and j are given in Fig. 1 and areabout _+_ 15% on a x i s . Thus, at t h i s time the question of the s t a b i l i t y of the m=l mode can be unambiguously decided only f o r cases of q-values considerably less than u n i t y .

SOLTWISCH et al.

266

i
(%)

(%) f/fV

3-

1

/
\

.I

///

s

f?l
<J

)z

8.'
\

.S

\
\
^ ^ \ \

.

\v

--t

V

(d)

location oprobing be/ \S
Z _UID-V ) r

s

tut

•§

^> i
acy

-

u1

O

(

I

(b)

\

Cn

~*

"i

f**

*~

i^

beyond

1
<^>

•

so

K

o

(6ep) xi

.1

IAEA-CN-47/A-V-1

267

(a)

r

s

\

ms

/700

, \

A

000 ms2-

200 ms

I

0-

J

"A
-^
\

170

-

• 200

FIG. 2. (a) Electron density, (b) safety factor and (c-e) current density profiles at the end of current
ramp-up (200 ms; / = 625 kA-s'1, I = 300 kA) and in quasi-stationary conditions (700 ms, 1000 ms;
I = 335 kA, BT = 2 T). A non-destructive disruption occurring between 700 ms and 1000 ms changes
the discharge characteristics from a non-sawtoothing plasma of relatively high metal content to a cleaner
sawtoothing discharge. Safety factor and current density are plotted versus normalized flux surface
radius r/a to allow a comparison with linear tearing mode stability theory in cylindrical geometry.
Dashed curves represent computed stable profiles with A'-values as listed.

3.

DEVELOPMENT OF THE CURRENT PENSITY PROFILE IN TIME

An example of the development of the current distribution
profile in time is shown in Fig. 2 for nearly constant total
plasma current. Under the influence of the thermal instability
the distribution contracts until the stationary profile with q
on axis of about 0.6 is reached. The values of the stability
criteria A1 of linear tearing mode theory in cylindrical approximation are given in the inserts for the individual modes, indi-

268

SOLTWISCH et al.
(a)

(c)

_6

• 22 =66

r
j

J
130

3/4
m/n

200

\

\

1

100

\

i ?
150

4

5

6
190

- 8.49- 2.53-10.20
-5.88
-4.77

Y\

\

\
1
\

3/4
2/2
4/3
3/2
2/1

_ 2/1.

i

7
210

0.8

0.4

Rlcml

r,Q

(d)

y
/
/
/

— -

770

t(ms)

0.4

0.8 r/Q

FIG. 3. Quasi-stationary low-q discharge (I = 476 kA, BT = 1.7 T, q(a) = 2.33).
(a) Electron density distribution, (b) Fluctuations of line-integrated electron densities along the chords
indicated in frame (a) (ne in units of 6.6 x 1014 cnf2). (c) Current density distribution, (d) Safety factor
profile presented as in Fig. 2.

cated by the arrows (note that the m=l, n=l mode cannot be treated by this method). During the developmentin time the profiles
always show small negative values of A1,indicating marginal stability. It must be concluded that at the time of 700 msec the
m=l mode is also stable because no m=l or sawtooth activity isobserved,but the q=l radius is clearly at r/a = 0.3. This is accompanied by a peaked electron density profile indicated in frame a of Fig. 2. At 1000 msec the j-profile has further steepened, and sawtooth activity connected with the marginally unstable m=l mode occurs with the accompanied flattening of theelectron density profile.

269

IAEA-CN-47/A-V-1
(a)
• ISO

\

\

\

/
\

1

/
130

1

2

1,
170

150

5

6

7
755

R (cm)

765

(d)

(c)

t (msl

(e)

600

/

•

->
\

400

\
\
\

• soft X- ray d ita

n0.4

0.8,

—1—10.4

0.8

r/a

FIG. 4. Low-q discharge (1 = 485 kA, BT = 1.57 T, q(a) = 2.1) suddenly developing a large
m = 1 mode after about 500 ms of stationary behaviour (frame (b) showing the fluctuations of lineintegrated electron densities at the chord positions indicated in frame (a)). Profiles about 25 ms before
this event: (a) electron density, (c) safety factor, (d) current density, showing a pronounced shoulder
at the q = 1 radius, and (e) electron temperature calculated from the j-prqfile using neoclassicalresistivity with Ut = 1.50 V and Zeg = 2 (dots represent average temperatures in the time interval
400-800 ms measured by soft X-ray pulse height analysis).

4.

PLASMAS WITH q AT LIMITER NEAR 2

In Fig. 3, a current distribution with a limiter q-value
near 2.3 is shown. The profile is maintained by the m=l instability-induced sawtooth activity. The m=l mode must be marginallystable. The A' values of the higher-m modes derived for the cylindrical model are all negative. In this case the q-value on
axis is so low that even without the plasma shift method theaccuracy is sufficient to definitely exclude a value of q(0)=l.
A case with q at limiter of 2.1 is shown in Fig. 4. The
current distribution is held stationary by the increased transport caused by sawtooth activity initiated by marginal instabil i t y of the m=l mode. The q=l radius has moved out to r/a ~ 0.5.

270

SOLTWISCH et al.

2.0

R(m)

FIG. 5. Current density profile stable to the m = 1 resistive mode in toroidal geometry(a/R0 = 0.46/1.75, BT= 1.49 T) with a finite $p of 0.25 according to the Princeton PEST III code.
Full line: experimental profile of Fig. 4, used to calculate the required input information q(*p); dashedline: stable profile after self-consistent equilibrium recalculation.

The q(n) is about 0.6 +_ 15%. Frame e of Fig. 4 gives an indication of the electron temperature from soft X-ray measurements(dots). The solid curve represents the temperature profile derived from the current density profile using neoclassical electron
conductivity.
For this steep profile the nonlinear stability behaviour ofthe m=l mode is so strong that an actually measurable plateau
occurs in the j-profile. The precarious stability situation isevident by the rapidly growing m=l mode (at 767 msec) which (incontrast to the small local m=l modes which may produce sawtooth
activity without affecting the current distribution) produces,by total reconnection, a complete disruption of the plasma
column.
The current density profile of Fig. 4 has been analyzed
with respect to tearing mode stability with the PEST code intoroidal geometry . Figure 5 shows the result. The solid curveis the experimental profile which was entered into the code asa q(i|i)-profi1e. The dotted line represents as a check the recomputed j-profile for the recalculated plasma equilibrium. The
computed values of A1 for the 1/1, 4/3, 3/2 modes are allslightly negative, indicating marginal stability including the
1/1 mode. As expected, stability cannot be obtained in cylindrical geometry.** From the experimental results as well as from

271

IAEA-CN-47/A-V-1

IlkAl B,ITI q ( Q )
t
2
3
4

5
6

485
473
476
340
202
191'

q(O)

1.S8

2.10

0.67

1.70

2.33

0.74

2.51

3.30

0.72

1.70

3.17

0.78

1.70

5.46

0.80

2.00

6.34

0.88

= 1.75m; a = 0.4-6 IT

FIG. 6. Superposition of normalized current density profiles measured in a variety of quasi-stationary
sawtoothing Ohmic discharges (see insert). In reduced co-ordinates the profiles resemble roughly auniversal profile.

this first theoretical result we conclude that the m=l, n=l resistive mode can be linearly or at least nonlinearly stable intoroidal geometry. The dependence on geometry and other profile
effects is under investigation. 11

5.

PROFILE SHAPE

The experimental profiles show a dependence of the effective minor radius on the plasma current and current density on
axis. With the current density clamped inside the q=l radius,
the profile outside the q=l radius is also fairly well fixed by
the restriction of the thermal transport. It is thus convenient
to display the experimental profiles for various external tokamak parameters in appropriate coordinates. Figure 6 shows thatthey can roughly be represented by one profile. This effective
minor radius dependence requires the effectiveness of the virtual or radiation limiter, which in the investigated cases was
assured by the presence of a small amount of light impurities(carbon). Theoretically the situation around the q=l radius isstill not completely clear, and thus the q=l radius cannot be
determined precisely because the m=l stability criteria are notsufficiently well known to warrant numerical calculations.

272

SOLTWISCH et al.

SUMMARY
In stationary state plasmas with sawtooth activity, q(0) is
found to be considerably below unity. In most cases, but not exclusively, sawtooth activity is not connected with complete reconnection, which would increase q(0) to near unity. Experimentally, the m=l, n=l tearing mode is intermittently stable or atleast saturated at an undetectable level. Theoretical estimateswith the PEST code confirm these results in the cases investigated. The effect of current density clamping near the axis results in a general shape of the stationary, sawtooth dominated,current distribution and determines the effective minor radius.
ACKNOWLEDGEMENT
This work was in part supported by U.S. DoE Contract No.
DE-ACfl2-76-CHO-3O73.
REFERENCES
[1]

[2]
[3]
[4]
[5]

[6]
[7]
[8]

[9]
[10]
[11]

BISHOP, A.S., HINNOV, E., in Plasma Physics and Controlled Nuclear Fusion Research (Proc.
Conf. Culham, 1965), Vol. 2, IAEA, Vienna (1966) 673.
GRISHAM, L.R., BOL, K., Nucl. Fusion 18 (1978) 315.
YOSHIKAWA, S., CHANCE, M.S., Bull. Am. Phys. Soc. 31 (1986) 14.
FURTH, H.P., ROSENBLUTH, M.N., RUTHERFORD, P.H., STODIEK, W., Phys. Fluids
13 (1970) 3020.
VON GOELER, S., STODIEK, W., SAUTHOFF, N.R., Phys. Rev. Lett. 33 (1974) 1201.
SOLTWISCH, H., Rev. Sci. Instrum. 57 8 (1986) 1939.
ALLADIO, F., MAZZITILLI, G., TUCILLO, A.A., VLAD, G., in Controlled Fusion andPlasma Physics (Proc. 12th Europ. Conf. Budapest, 1985), Vol. 9F, Part I, European PhysicalSociety (1985) 138.
LACKNER, K., private communication. (This code was provided by Dr. Karl Lackner,
Max-Planck-Institut fur Plasmaphysik, Garching, and installed by A. Kaleck, Julich.)
FURTH, H.P., RUTHERFORD, P.H., SELBERG, H., Phys. Fluids 16 7 (1973) 1054.
GRIMM, R.C., DEWAR, R.L., MANICKAM, J., JARDIN, S.C., GLASSER, A.H.,
CHANCE, M.S., in Plasma Physics and Controlled Nuclear Fusion Research 1982 (Proc. 9thInt. Conf. Baltimore, 1982), Vol. 3, IAEA, Vienna (1983) 35.
MANICKAM, J., Phys. Rev. Lett. 51 (1983) 1959.
JAHNS, G., Nucl. Fusion 22 (1982) 1049.
MANICKAM, J., Bull. Am. Phys. Soc. 31 (1986) 1431.

DISCUSSION
A. GIBSON: Can you clarify whether your observed universal current density
profile also holds good for discharges with large additional heating, especially in thecase of off-axis heating?
H. SOLTWISCH: So far we have not made any systematic studies of the current
profiles during ICRH on TEXTOR.

IAEA-CN-47/A-V-1

273

R.J. GOLDSTON: We have a ten-channel interferometer on TFTR, and when
we carry out Abel inversion, we find significant in-out asymmetries that cannot be
explained by a picture in which ne = n e (^). We attribute this result to asymmetries
in the scrape-off plasma. Of course, our limiter geometry is different from yours, but
I wonder whether there are any uncertainties due to effects of this sort which caninterfere with your measurement of q?
H. SOLTWISCH: For standard Ohmic discharges, especially at moderate ne,we have no indication that ne varies considerably on a flux surface. Lithium beammeasurements of ne in the scrape-off layer usually match the ne profiles obtained byinterferometry quite well.

IAEA-CN-47/A-V-2

STABILITY AND CONFINEMENT STUDIESIN THE PRINCETON BETA EXPERIMENT (PBX)*
M. OKABAYASHI, K. BOL, M. CHANCE, P. COUTURE1,
H. FISHMAN, R.J. FONCK, G. GAMMEL, W.W. HEIDBRINK2,
K. IDA3, K.P. JAEHNIG, G. JAHNS2, R. KAITA, S.M. KAYE,
H. KUGEL, B. LeBLANC, J. MANICKAM, W. MORRIS4,
G.A. NAVRATIL5, N. OHYABU2, S. PAUL, E. POWELL,
M. REUSCH, S. SESNIC, H. TAKAHASHI
Princeton Plasma Physics Laboratory,
Princeton University,
Princeton, New Jersey,
United States of America
Abstract
STABILITY AND CONFINEMENT STUDIES IN THE PRINCETON BETA EXPERIMENT (PBX).
Values of /3t in excess of 5% were achieved in the PBX tokamak by producing plasmas withindentations > 20%. Despite the high beta equilibria achieved, the plasmas still remained within the firstregime of stability. Comparisons of data with theoretical analyses were made in order to evaluate the
MHD stability properties for plasmas close to the Troyon-Gruber limit at all q. At the low q and highl3t, the plasmas were limited by hard disruptions whose characteristics were generally consistent withthose expected from ideal external kink modes. At higher q, a wide variety of internal and edge MHD
modes were found to be associated with loss of both thermal and fast ion energy. This MHD activity
included fishbones, internal sawtooth-like disruption, Edge Relaxation Phenomena (ERPs), coupled disruptions, and the 'tangential equivalent' of fishbones. Most of the activity was associated with internal
m/n = 1/1 modes. At low plasma density, the tangential instabilities observed in discharges with puretangential and mixed tangential and perpendicular neutral beam injection, led to fast ion losses that werecomparable to those associated with fishbones in purely perpendicular injection.

* Work supported by the United States Department of Energy, under Contract
No. DE-AC02-76CHO-3073.
1
2
3
4
5

Institut de la recherche d'Hydro-Qudbec, Canada.
GA Technologies Inc., San Diego, CA, USA.
University of Nagoya, Nagoya, Japan.
Balliol College, University of Oxford, Oxford, UK.
Columbia University, New York, NY, USA.

275

276
1.

OKABAYASHI et al.
Introduction

Recent theoretical results [1,2] indicate that the beanshaped configuration can lead to high-8 t plasmas in the second
stability regime for both ballooning and internal kinkmodes. The PBX [3] has achieved values of 8 t in excess of 5%
in plasmas that have indentations >20%. While the (3. values
are greater than those previously achieved on conventional
tokamaks, the plasmas still reside within the first regime ofstability, as defined empirically by the relation £S < 2.58 ,where 8 = u I /(a . ,B. ), with a ., the midplane half radius
c
o p
mid t
mid
[4,5].
Extensive theoretical and experimental analyses havebeen carried out in order to evaluate the MHD phenomena
associated with the present operational 8 t limit. This paper
summarizes the results of these studies and, in addition,compares the MHD properties and energetics of plasmas withpurely perpendicular injection to those with purely tangential
injection. The final topic is a brief discussion of the confinement and transport properties of H-mode plasmas.
The experimental conditions near the Troyon-Gruber 8 t
limit are summarized in Fig. 1. While Fig. 1(a) contains thedata points over all experimental conditions, the data in
Fig. 1(b) and 1 (c) were limited, for purposes of clarity, todischarges with 0fc > 1.5 8 •At high 8 C , discharges wereobtained with indentations of 18 to 22%; here the indentationi
is defined as i = d / ( 2 a . . + d ) , where d is the radial
depth of the indentation. At lower 8 , the highest attainable
indentations were progressively less, and the edge q values
were higher.
Along the 0 t boundary, the various MHD phenomena andplasma characteristics that were observed depended on thevalue of 8 • At the highest 8 , plasmas were limited by hard
disruptions exhibiting little MHD activity until just before
the disruption. At lower values of 8 (0.8 to 1.3), 8 t saturation, sometimes with subsequent 8 t collapse, was observed.
The 8t~saturation phase of the discharge was associated withinternal and external MHD bursts, while the collapse phase wasassociated with a more continuous, saturated amplitude mode in
addition to ERP's.
Much, but not all, of the MHD activity
during both phases of these discharges exhibited internal m/n= 1/1 activity.
Examples of high- and low-8 discharges areshown
in Fig. 2(a)(0=5.3%,
8 = 2 . 5 , i^ = 20%) and
Fig. 2(b)(8 t = 2.8%, 8 C = 1-0, i °= 10%). "ihe high-8 t
configuration could be produced only by ramping the current at
a rate of >1.5 MA/sec.
For the 8 t collapse discharges, thecurrent was held relatively constant during the neutral-beampulse.

277

IAEA-CN-47/A-V-2

,

, .

,

I , I

'

• i '

, ,

I , ,

, .

I I , .

, I

, , ,

(a)

'

'

'

I '

'

'

'

'

l

(b)

0.2-

0.1-

10-

(c)

*" *

*'

5-

p > I.5&

0

0.5

1.0

1.5
fie

2.0

2.5

3.0

FIG. 1. Range of experimental parameters, (a) Achieved @t versus Troyon-Gruber parameter $c,
including all experimental conditions; (b) indentation in versus /3C only for /3 > 1.5 0C; (c) qedge versus/3C only for fi > 1.5 |8C.

278

OKABAYASHI et al.
(a)

1.2

1.6
X (metres)

1.2

1.6

2.0

X (metres)

FIG. 2. Examples of discharges with (a) high @c (high 0, and low q with large current ramp) and(b) low /3C (medium /3( and medium q with low current ramp). Equilibrium was calculated at
t = 560 ms for high Qc (a) and t = 575 ms for low £ c (b).

2.

Discharges Near the (3. Limit

In an attempt to delineate the mode or modes responsible
for the experimental 0. limit, theoretical 0. limits were
calculated using equilibria based on the experimental conditions (Fig. 3 ) . Calculations were performed in which thefollowing parameters were matched to the experimental measurements: coil and plasma currents, poloidal flux, diamagnetic
signal, and a total pressure profile taken to be proportional
to the electron pressure profile measured by Thomson scattering. In some cases these pressure profiles were cross checkedagainstthe pressure
profilesdetermined
by
the
time
dependent, generalized geometry transport code, TRANSP. Threecases at various |3 were chosen as representative.
In thiscalculation q(0) was fixed at 0.8.
The calculations focusedon three instabilities: the internal and external kink and the279

IAEA-CN-47/A-V-2

10-, , , I I , , , . I , , , , I , , , I I I I I I I• Non-Disruptive* Disruptive

I

Ballooning Limit

Internal
Kink Limit 4

5-

FIG. 3. Comparison of experimentally achieved j3; values and onset of various ideal MHD instabilities.
Cases chosen as representative are: (1) 0C = 2.5; j3 = 5.3%; in = 20%; (2) &c - 1.5; 0 = 3.1%;
in = 17%; (3) Pc = 1.0; 0 = 2.8%; in = 10%.

high-n ballooning modes.
For high f} and hence highindentation, the internal kink mode is completely stable andthe marginal f3 for ballooning modes is between 8 and 9%.
However, without a conducting shell, the external kink goes
unstable at 3.5%, but with a conducting shell at one plasma
midplane half-width, k>w/a . = 2, the marginal 0 for stability is increased to 4.5-5%, close to the observed value.
At slightly lower 0 c (S c => 1.5, 8 t = 3.1%, i n = 17%), therelative stability of the three modes is similar to that at
higher p c . At the lowest Q , with the lower indentation (6= 1.1, f$t = 2.7%, i n = 10%), the internal kink is unstable,
indicating a more peaked pressure profile than that assumed inprevious theoretical studies [2]. With these pressure and q
profilesconsistent withthe experimental
results, theexternal kink and ballooning modes can be destabilized atrelatively low values of 6 (1.5 to 2.5%).
3.

MHD Activity

3.1High B c
High «($ cc- 2.5) discharges were produced by employingca large current ramp of >1.5 MA/sec for a period of up to 140
msec during neutral-beam injection, resulting in discharges

280

OKABAYASHI et al.

similar to the one shown in Fig. 2(a). The plasma parameters
for this discharge were I = 570 kA, n (0) = 6.0 x 10 1 3 cm~ 3 ,T (0) = 2.0 keV, and T±(.0f = 4.5 keV at t = 570 msec [6]. All
discharges at these high 0 terminated in a hard disruption.
The disruption, at the highest 0 (and lowest q ) , was characterized by rapid growth (150 ysec) of magnetic activity
leading to a simultaneous loss of plasma energy and current.
At slightly lower 0 , growth of magnetic activity led first to
the partial (30%) loss of plasma energy with no loss ofcurrent[7], but with several msec delay to the loss of bothcurrent and the remainder of the energy. The amplitude of the
m/n = 1/1 component was smaller in the case of the shorter
delays.
The delay period was determined to be a period of
non-oscillating exponential growth of an n = 1 mode locked inphase with a specific toroidal orientation, presumably due tosome field misalignment.
The poloidal mode structure of the magnetic activity
agrees well with that calculated from PEST for the externalkink with some of the ideal MHD constraints relaxed [8].
Strong coupling across the discharge results in a spectrumof m components, including an m/n = 1/1 component near thecenter.
The amplitude of this component is theoretically
expected to depend strongly on the exact value of [q(0) - 1 ] .
At still lower 0 c (1.5 to 2.0), the internal m/n = 1/1
crash was accompanied by a smaller energy loss, and it did notnecessarily lead to a current termination [7].
3.2

Low ftc
At values of ft « 1 U n * 10%), stability calculations
indicate that the plasmas may be unstable to external orinternal kink or ballooning modes depending on the specific 0.
value (Fig. 3 ) .
Experimentally these plasmas exhibited avariety of internal and edge MHD phenomena, some of which wereassociated with losses of thermal and/or fast-ion energy. An
example of this is shown in Fig. 4; for this discharge four
neutral beams were injected(two perpendicular and twoparallel) with a total injected power of 5.2 MW.
The diamagnetic signal shows 0. increasing and then saturating at
about 590 msec and finally collapsing at about 640 msec.
During the time of 0 t saturation, there were various MHD modes
excited: internal disruptions with m/n = 1/1 precursors (582,590, 598, and 620 msec), fishbones (606, 608, 610, and 615msec), and edge relaxation phenomena — ERP's [9-12], causing
energy loss predominantly from the outer plasma region (e.g.
592 msec). Associated with the collapsing phase in the plasma
3 t was a continuous, saturated amplitude mode that exhibited
an m/n = 1/1 internal structure and, depending on the exact

IAEA-CN-47/A-V-2

CHARGEEXCHANGE
FLUX(A.U.)

281

n ^
u

^

-5
MIRNOV
SIGNAL(A.U.)
Ha-SIGNAL(A.U.)

2

j e -

550
TIME(ms)

FIG. 4. Characteristics of low @c discharges. A 0 r is diamagnetic flux loop signal, and 0t is determinedfrom A<t>j and flux loop measurements.

discharge, either an n = 1, m = 2 or a higher-m coupled mode
structure as deduced from the magnetic probe measurements.
The plasma stored energy dropped even more dramatically whenthe mode locked (t = 667 msec) and, after locking, grew
exponentially.
3.2.1(Jt-Saturation Phase

While the m/n = 1/1 precursor to the internal disruption
during the $ -saturation phase is similar in some respects(frequency and growth rate) to that of a typical sawtooth,
other features point to their marked difference. Figure 5(a)shows the amplitude and phase of the precursor oscillations asmeasured by the horizontally viewing soft X-ray array.
Therelative amplitude of the oscillations, s'/s, exhibits a broad
structure and shows a gradual phase change across the plasma
center unlike the typical sawtooth which exhibits a clear 180°
jump in phase [Fig. 5(b)].
Here s'/s is defined as the ratioof the precursor amplitude s to the average X-ray signal

282

OKABAYASHI et al.

(a)

-200

0.03

30

0
Z(cm)

-30
—|

1

1

1

1

-

1

0

—

0

o

0

1

1

-1

1

1

1

1

0

0

1

1

1

—1—'— 6

—

T

o

(b)

-

# o'_
(o)

0

101

-30

0

o

0

0 —i
200

0

0

0
0

o

—

o

o

o

0 -<

o

o

9
1

Q
_l

0
Z(cm)

1

1

1

30

FIG. 5. (a) Amplitude and phase of precursor with two perpendicular and two parallel injectors(at t = 620 ms). Solid lines are theoretical values predicted with PEST code, (b) Behaviour of typicalsawtooth.

0.4 -FIG. 6. Change of soft X-ray radial profile before and after mode crash, (a) Crash of internal mode
producing ERP with several ms delay; (b) crash of internal mode having additional inversion aroundq = 2 surface; (c) crash with ERP event; (d) typical sawtooth crash.

IAEA-CN-47/A-V-2

283

amplitude s. Another difference between the internal disruptions shown here and the typical sawtooth was the profile ofthe heat redistribution at the time of the "crash," as shownin Fig. 6.
In the figure, the As/s for various types of
events are shown, where As/s is the drop in the X-ray signal
at the crash normalized to the average value. While the sawtooth exhibits a relatively large drop in the plasma centerand an inversion near the q = 1 radius (curve d ) , one type ofinternal disruption exhibits only a small, but broad, drop in
the center, with large positive As/s beyond (curve a ) . The
normalized amplitude As/s did not decrease within the field ofview. Occasionally, an additional inversion was observed nearz » 40 cm, close to the estimated location of q = 2 surface asshown on curve b.
The "coupled disruption" events [curve(b))] coincided with spikes in the H /D emission. The typeof crash that does not exhibit a second inversion (curve a)tended to cause the ERP event (curve c) within a few msec,
most likely due to the development of a large temperature
gradient at a larger minor radius following the internal disruption.
The large temperature gradient did not develop in
the second type of internal disruption (curve b) because of
the double inversion; therefore, this "coupled disruption"
event did not lead to an ERP.
There is also a possibility
that coincident with the coupled disruption, a small ERP
occurred which might have been masked in our measurement by
the q = 2 disruption. The energy loss evaluated from the diamagnetic flux signal at the ERP events was 3 to 4% of thestored energy.
This value was consistent with the thermalenergy loss estimated from the change of the soft X-ray
signal.
To test the relation between the observed internal
oscillations and the ideal MHD prediction, the soft X-ray
array signals were calculated using the simulation from theideal-MHD code, PEST, assuming that the emissivity was proportional to the integral along the sight of the square of thepressure.
As shown by the solid line in Fig. 5(a), theory
predicts a broad phase shift presumably due to the strong
coupling to the higher-m components. This compares well withthe experimental observation.
Furthermore, it was recently
pointed out [13] that pressure-driven internal modes are
unstable at high Bfc/S , even for q(0) > 1 if the shear is
reduced around the magnetic axis.
Under that condition themode tends to couple to the plasma periphery; the experimental
profiles were not far from the onset condition.
3.2.2

8t-Collapse Phase

The actual collapse of g t occurs
steady-state portion of the discharge.

during an otherwise
Associated with the284

OKABAYASHI et al.

beginning of the loss of plasma and beam energy during thistime was a continuous mode, as observed on various magnetic
pick-up coils, with a frequency of up to 15 kHz and locked tothe plasma rotation. This continuous mode grew after several
internal disruptions and/or fishbones. The internal structureof this mode was similar to that of the internal disruption
precursor, exhibiting m/n = 1/1 behavior.
Dependent on the
precise discharge, however, the strength of the internal
m/n = 1/1 varied.
For discharges in which the continuous
m/n = 1 / 1 was dominant internally, the magnetic loops observed
higher m (4 to 6)/n = 1 components, indicating a strong
toroidal coupling of the internal mode.
However, fordischarges in which the internal m/n = 1/1 was not dominant(e.g., higher amplitude m > 2 components observed on the soft
X-rays), the magnetic loops indicated m/n = 2/1 structure of
the continuous modes.
During the continuous modes, theexcitation of ERP's occasionally coincided with a decrease infrequency (at constant Bg signal amplitude) of the mode.
The continuous oscillations were observed to slow down
during the discharge, corresponding to a slowing down of theplasma rotation.
Once locked, however, the n = 1 mode grew
(it is difficult to identify m numbers due to slow rotation),as observed from a toroidal array of B loops, and was associated with a large loss of energy. As was the case in the lowq disruption, the mode locked in phase with the same toroidal
orientation
orientation mentioned earlier in connection with the high-S t
disruption.
3.3 Tangential Beam-Driven Instability at Modest Indentationand Low Density
With purely tangential injection for plasmas with 3 C= 1.0 to 1.2 and low indentation, a loss of fast ionsassociated with bursts of low-frequency (20 to 30 kHz) andhigh-frequency
(150
to
220
kHz)
MHD
activitywasobserved [14]. As shown in Fig. 7 the fast-ion loss, detectedon both the neutron and the neutral charge-exchange detectors,
caused a significant power loss only in low density (n < 1.5
x 10 cm" ) plasma, where there was a correspondingly highratio of beam ion pressure to plasma thermal pressure. The
fast-ion losses occurred during periods of either low plus
high-frequency activity or high-frequency activity alone. The
range of high frequencies was comparable to the transit
frequency of tangentially moving beam ions, indicating aresonant interaction between the fast particles and the mode
similar to fishbone loss [15].
The greatest particle loss
occurred
duringthe combined
low- and high- frequency
events.
The low-frequency portion exhibited a growing
m/n = 1/1 structure and led to a "crash" that had a much

285

IAEA-CN-47/A-V-2-0.3^1(mWb)-0.4

NEUTRONFLUX
1CHARGE 4
EXCHANGEFLUX (A.U.) 0
MIRNOV
SIGNAL(A.U.)

0-5

Hq-SIGNAL(A.U.)CENTRAL X-RAY DIODE
(Z=0)
.,—
^
~
E06E DIODE(Z = 27cm)- ^ ~ ^ ^

•

*^V_-

525

530

535

540

545

550

TIME(ms)

FIG. 7. Characteristics of parallel injection.

1.6

~1

1—

T"

1.4

o,»-

2ii

I
1.8

,

Unstoble

1.2'1.0 -

Stable0.8
0.6
0.4 _.
0.6

1

0.8

,

I

1.0

,

I

1.2

.

L

1.4

1.6

L
2.0

2.2

FIG. 8. Comparison of onset dependence of fast ion loss on /3 , (J, for perpendicular andparallel injections.

286

OKABAYASHI et al.

smaller central drop and much broader inversion radius than atypical sawtooth, but was similar to that of the internal
disruption cases. The average energetic ion confinement time,
inferred from the drops in the neutron emission, differedlittle from that associated with the fishbone instability
during purely perpendicular injection.
In addition, thevalues of $. and (3 for which the tangential and fishbone
modes go unstable are similar (Fig. 8 ) .
4.

Confinement and Transport in H-ModePBX plasmas limited by a magnetic separatrix internal tothe vessel rather than by a material surface were produced byusing a flat or only slightly increasing (<1 MA/sec) plasma
current.
These plasmas, with indentations up to 18%,
exhibited H-mode transitions typical of the observed diverted
plasmas in ASDEX and PDX: a sudden drop in the H /D signal
associated with a continual rise in plasma density at constantgas puff rate.
Both effects indicate an increase in theeffective particle confinement time.
However, what did notnecessarily follow the H-mode transition in PBX was a sustainable enhancement of the energy confinement time.
For
instance, at the lower toroidal fields (<1.2 T ) , energy
confinement times increased by 50% above the nominal L-mode
value at the time of the transition, but decayed back to the
L-mode values within 50 to 100 msec. One, but not the only,
cause of the H-mode confinement deterioration in many of thelower toroidal-field discharges was radiative loss due tocentral accumulation of metallic impurities.
In thesedischarges, central radiative power densities of up to 1 W/cm^
were observed, and this power density was consistent with that
expected
assuming
neoclassicaltransportofthe
impurities [16].
At higher toroidal fields, a sustainable confinement
enhancement was achieved, as shown in Fig. 9.
The steadystate confinement time after the H-mode transition is plotted
as a function of B in Fig. 9(a). At 1.0 T, the T E values are
similar to those of L-mode plasmas, while at 1 .« T, the T_
JCJvalues are approximately 60% higher than the L-mode, and scaleas (I K 1 ' 2 )/9, which is comparable to the good H-mode plasmas
produced in PDX.
The ability to sustain the enhanced T_, atthe higher B
value is associated with an increase in theratio of the divertor to midplane recycling, as is seen in
Fig. 9(b).
The transport properties of these H-mode plasmas havebeen investigated using the TRANSP code, in particular toaddress
the
issue
of
the
relationshipbetweentheexperimentally deduced x± profile and the neoclassical one.

287

IAEA-CN-47/A-V-2

1.5

TF(Tesla)

2.0

FIG. 9. Dependence of H-mode confinement time on toroidal field.

An analysis of T.-profile data on Dili has indicated that theion thermal diffusivity is neither neoclassical-likeinmagnitude nor in form [17]. The Dili results indicated thatwhile X' w a s °f t n e order of the neoclassical value in thecenter of the discharge, it could be as much as ten times theneoclassicalvalue
halfwayout, andthat
ion
thermalconduction, underthese
conditions, could
dominatetheelectron and ion power balance.
Radial ion temperature profiles from charge-exchange
recombination of O^ + [18] have allowed the further investigation of this topic in these PBX H-mode plasmas. An exampleof the results of these analyses is shown in Fig. 10. Figure10(a) shows the T. and T profiles, while Fig. 10(b) shows thecorresponding difrusivities. The T. profile labeled ax- wascalculated in the traditionally standard fashion: the neoclassical multiplier a was adjusted in order to match themeasured central T., and it was assumed to be constant acrossthe plasma in order to determine the remainder of the T.
profile [19]. For the case shown, a = 0.5 when the ChangHinton neoclassical x- i-s assumed [Fig. 10(b)]. The second T.
profile is the measured one, and this was used directly todeduce x.; from the ion power balance.
The shaded region
corresponds to the band of X' deduced is this fashion. Rather
than plot a y. from a single power balance calculation,

288

OKABAYASHI et al.
3.0

Tii.xf)(a)^

FIG. 10. (a) Calculated (—— ) and measured () 7} profiles; (b) thermal diffusivity.

several calculations were performed/ each with a T. profile
varied randomly within the experimental uncertainties.
The
shaded area, then, is the envelope of x- ' s a s determined fromthese various calculations. Because of uncertainties in T.(r)0 cm
and T (r), the v. (r) inside r= 5 cm and beyond 20
midshould be ignored.
It is clear from Fig. 10(b) that the two x^ profiles [0.5x x^h and x-; (^) ] are not dramatically different in magnitude
unlike the Dili result. In fact, the computed x- ^ s less than
the neoclassical value in the inner half of the dischargerather than greater. In addition, to within the uncertainties
in the data and the calculation, the difference in the two xprofiles is not reflected in the T. profiles; the T. (r)determined from the assumption of 0.5 x x- is consistent withthe measured profile.
The reason for this is that in the

IAEA-CN-47/A-V-2

289

discharges analyzed, the ion thermal conduction loss neverdominates and is responsible for, at most, 30% of the totalenergy loss.
5.

Summary

An extensive study was carried out to evaluate the MHD
behavior of plasmas residing along the first stability boundary. At the highest f3 , the achievable $. is limited by hard
disruptions produced by external kink modes or substantial
energy loss due to the large internal disruption. At lower 8
and lower plasma indentations, various MHD modes, all with
m/n = 1/1 internal activity, grow and are associated with lossof thermal and fast-ion energy. The growth of these modes is
consistent with the onset of internal kink modes. In addition
to the internal modes, edge phenomena, such as ERP's and theedge inversions in the coupled disruptions, lead to thermalenergy loss.
Purely tangential injection excites lowfrequency modes similar to fishbones, but which rotate withthe plasma, and high-frequency modes that appear to be responsible for fast-ion loss through some resonant process.
The
presence and locking of the continuous mode clearly indicatesthat non-ideal MHD effects must play an important role during
the saturated MHD activities in the 3 -collapse phase.
Finally, transport analyses of H-mode plasmas with theobserved T.(r) profile did not indicate a large difference
between measured and neoclassical x- values.

REFERENCES
[1]
[2]
[3]
[4]
[5]
[6]
[7]
[8]
[9]
[10]
[11]
[12]

CHANCE, M., JARDIN, S., STIX, T., Phys. Rev. Lett. 51 (1983) 1963.
MANICKAM, J., GRIMM, R., OKABAYASHI, M., Phys. Rev. Lett. 51 (1983) 1959.
OKABAYASHI, M., et al., in Plasma Physics and Controlled Nuclear Fusion Research 1984
(Proc. 10th Int. Conf. London, 1984), Vol. 1, IAEA, Vienna (1985) 229.
TROYON, F., GRUBER, J., Phys. Lett. A 110 (1985) 29.
SYKES, A., TURNER, M.F., PATEL, S., in Controlled Fusion and Plasma Physics (Proc. 1 lth
Europ. Conf. Aachen, 1983), Pt. II, European Physical Society (1983) 363.
BOL, K., et al., Phys. Rev. Lett. 57 (1986) 1891.
TAKAHASHI, H., et al., Rep. PPPL-2391, Princeton Plasma Physics Laboratory (1986).
OKABAYASHI, M., et al., in Proc. Int. School of Plasma Physics, Varenna, Italy, 1985 (to be
published).
KAYE, S., BELL, M., BOL, K., BOYD, D., BRAU, K., et al., J. Nucl. Mater. 121 (1984) 115.
FONCK, R.J., BEIRSDORFER, P., BELL, M., BOL, K., BOYD, D., et al., in Heating in
Toroidal Plasmas (Proc. 4th Int. Symp. Rome, 1984), Vol. 1, ENEA, Frascati (1984) 37.
WAGNER, F., BECKER, G., BEHRINGER, K., CAMPBELL, D., EBERHAGEN, A., Phys.
Rev. Lett. 49 (1982) 1408.
McGUIRE, E., et al., in Plasma Physics and Controlled Nuclear Fusion Research (Proc. 10th
Int. Conf. London, 1984), Vol. 1, IAEA, Vienna (1985) 117.

290
[13]
[14]

[15]
[16]
[17]
[18]
[19]

OKABAYASHI et al.
MANICKAM, J., et al., private communication (1986).
HEIDBRINK, W., et al., Phys. Rev. Lett. 57 (1986) 835.
HEIDBRINK, W., et al., Rep. PPPL-2405, Princeton Plasma Physics Laboratory, Princeton, NJ(1986).
KAITA, R., et al., Plasma Phys. Controll. Fusion 28 (1986) 1319.
IDA, K., et al., Rep. PPPL-2313, Princeton Plasma Physics Laboratory (1986).
GROEBNER, R., PFEIFFER, W., BLAU, F., BURRELL, K., FAIRBANKS, E., et al., Nucl.
Fusion 26 (1986) 543.
GAMMEL, G., KAITA, R., FONCK, R.J., JAEHNIG, K., POWELL, E., Rev. Sci. Instrum.
57 (1986) 1800.
KAYE, S.M., GOLDSTON, R., BELL, M., BOL, K., BITTER, M., et al., Nucl. Fusion 24(1984) 1303.

DISCUSSION
G. BECKER: When studying ballooning stability in ASDEX discharges, it was
found that the results depended appreciably on the beam contribution to the pressure
profile and the time evolution of the current density profile. Did you include theseeffects in your ballooning mode analysis?
M. OKABAYASHI: Our MHD analysis was carried out with the equilibria
matched with the current and pressure profiles evaluated, first, from Thomson scattering Te(r) and ne(r) data, and, second, from the output of the TRANSP analysis.
The ballooning mode was stable for both profiles.

IAEA-CN-47/A-V-3

STABILIZATION OF THE TEARING MODE
T.C. HENDER, D.C. ROBINSON, J.A. SNIPES
Euratom-UKAEA Fusion Association,
Culham Laboratory,Abingdon, Oxfordshire,
United Kingdom

Abstract
STABILIZATION OF THE TEARING MODE.
Various methods of stabilizing MHD instabilities in tokamaks are examined and compared withexperiments. The intrinsic stabilizing effect of average curvature on the tearing mode is discussed, withparticular reference to JET. Stabilization by RF current drive/heating and induced helical fields is
considered. It appears possible to stabilize all gross resistive instabilities in the tokamak at moderate
beta.

1.

INTRODUCTION

The m = 2 tearing mode in the tokamak has been shown both theoretically [1]and experimentally [2] to play a dominant role in causing the major disruption. Control of this mode may lead to suppression of the major disruption and to an extensionof the operating regime of the tokamak (higher I, N, beta). Control of the m = 1
mode could permit more peaked pressure profiles, better control of the m = 2 mode
and avoidance of the problems associated with the sawtooth collapse.

2.

STABILIZATION BY PROFILE AND PARAMETER CONTROLIn this section we consider the intrinsic stabilizing effect of the average curvature on the tearing mode [3], which is particularly pronounced on JET because of its
tight aspect ratio, high conductivity temperature and high beta potential. We also consider the stabilizing effect of local RF current drive/heating modifications to theequilibrium current profile.
The studies discussed in this section are conducted using the linear stability code
FAR [4, 5]. This code solves the compressible (or incompressible) resistive MHD
equations in full toroidal geometry with no ordering assumptions.
2.1. The m = 2 tearing mode
The tight aspect ratio and the high degree of shaping in JET lead to strong toroidal coupling between the m = 2 harmonic and its resonant side-bands (m = 3, 4 ...
291

292

HENDER et al.
1

11„-————,

2.0-

1.5-

^. —

/
/

/

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/
0.5-

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*

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'

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FIG. 1. m = 2 and m = 3 toroidal side-band islands. The solid ellipse is the liner location, thebroken ellipse is the plasma-vacuum interface and the squares are the locations ofpoloidal coils.

for q0 > 1). Figure 1 shows the magnetic island structure in a constant toroidal
angle plane for JET shortly before a density limit disruption when the n = 1 modeis growing linearly and is not rotating. The poloidal field strength at the liner is
8IVB3 = 1%. In this case, q varies between 1.08 at the axis and 4.6 at the plasmaedge. The dominant m = 2 islands and the toroidal side-band m = 3 islands areclearly visible. The toroidal coupling is weakly stabilizing; decreasing the aspectratio at constant q (which strengthens the coupling) lowers the growth rate. A farstronger stabilization, however, arises from the average curvature effects which
increase with beta [3]. As the magnetic Reynolds number (S) increases (and thegrowth rate y decreases) the relative importance of the stabilizing average curvatureterms (DR < 0) increases. Figure 2 shows how the n = 1 growth rate varies

293

IAEA-CN-47/A-V-3

5x10

10

2x10

5x10

10

2x10

5x10

FIG. 2. Growth rate versus S for JET, with q = 1.08 (1 +(solid line is growth rate, broken line is frequency).

(p/0.508)Ywith S for a typical JET discharge with b/a = 1.4, j80 = 0.7% and
q = 1.08 (1 + Go/0.508)8)w. This case is moderately unstable in the cylindricallimit (A' = 4.7). The growth rate of the tearing mode is strongly reduced withincreasing S and becomes overstable at high S (as predicted in Ref. [3]). For valuesof S consistent with the flat-top in JET ( ~ 3 X 107) this case is approximately marginally stable to the n = 1 mode. The strength of the average curvature terms also
increases with beta (since DR oc (3) and complete stability to the n = 1 mode occurs
at j80 = 1-1% when S = 105, for this case. It should be noted that, in part, thisstabilization is probably also due to a decrease in A' as beta increases.
This inherent stability to tearing modes may be related to the absence of m = 2
oscillations and the low level of coherent MHD activity (5Be/B9 ~ 3 x 10~5) which
is observed for the majority of the time in JET discharges with q^ > 2.3. During
the q and density limit pre-disruption phase, in which strong m = 2 activity(6BS/B9 ~ 1 %) is observed, the temperature (and therefore S and beta) at the q = 2
surface decrease. Since the critical A' is proportional to (a2 T2/3o/Ro)1/3, thetearing mode is rapidly destabilized by falling |80 and S, in agreement with theexperimental observations.
For cases in which m = 2 activity is observed on JET, we have compared thetheoretically predicted and the experimentally measured phase and amplitude variations of the poloidal fluctuations (8Be) at the liner. Figure 3 shows a comparison of
the amplitude variation of 83$ with poloidal angle between the experimental
measurements on JET and the predictions from the FAR code. The theoretical results

294

HENDER et al.

1.0

-•

JET

I

'

1

(a)
0.5

-E

0

m
2

1.0

0.5

--

/

\

/

il -

\
(b)

\

i

FAR

I

'

(1)

-

1
120

(3)
240

360

Coil angle

FIG. 3. Fluctuating poloidal field amplitude variation for JET.
(a) RF shot with q^ ~ 5, (b) OH shot with qt ~ 7.0, (c) NBI shot with q$ ~ 6.5; with FAR
predictions for increasingly peaked current profiles: (1) (1 + (iji/OA)4)'1'', (2) {1 + ty/OA)2)'2,(3) (1 - t/0.85), q0 = 1.08 at $0 = 0.8%.

reproduce the strong in-out ballooning asymmetry shown by the experiment and indicate that the in-out ratio becomes stronger for more peaked current profiles. Between
the liner and limiter surfaces (in the FAR calculations) we have imposed theequilibrium constraints P = Jz = 0 and, to match experimental observations, useda high resistivity (5 X 104 relative to its central value). The locations of the liner,
the plasma-vacuum interface and the poloidal field coils are shown in Fig. 1 for thesame case as shown in Fig. 3, curve (2).
We have also investigated the effects of localized heating and/or current drive,
which allows us to tailor the current profile near the q = 2 surface. A possible
scheme for JET is to use ECRH downshifted by 60 GHz, which leads to a localized
unidirectional driven current [6]. We have added a Gaussian perturbation to the toroidal equilibrium current to simulate the driven current (51). We find that the stabilization is very sensitive to the radial location of the current perturbation and that it is
optimized when it is centred on the q = 2 surface; this is in agreement withexperimental results from T-10 [7]. From examining a range of q-profiles we conclude that 8I/I ~ 5 % should be sufficient to stabilize the m = 2 tearing mode in
JET, in all but the most extreme cases. Ray-tracing and Fokker-Planck calculations

295

IAEA-CN-47/A-V-3

0

5

10
Aspect15
Ratio

FIG. 4. Stabilization of the resistive m = 1 mode by lowering the aspect ratio when q' = 0, S = 10s

indicate that such a current can be achieved with 2 MW of ECRH [6]. It should be
noted that the intrinsic stability of the tearing mode in JET means that cylindricalestimates of 6I/I necessary to stabilize a given case would be much larger.
2.2. The m = 1 sawtooth mode
Because of the different physics of the toroidal m = 1 mode, stabilization techniques that are applicable to m > 2 instabilities will not necessarily work for them = 1 mode.
At |3 < |8crit, with monotonically increasing q-profiles and q0 < 1, the idealinternal kink mode is stable [8]. In these cases, however, the generally large valuesof A' (—100) result in a resistive kink instability with y « S" 1/3 , a result which has
been confirmed numerically [9]. The resistive m = 1 kink mode has been proposed
as providing an explanation for the sawtooth [10], though experiments on JET
suggest a much faster scaling [11]. The rapid growth of the resistive m = 1 kinkmode means that the average curvature terms do not stabilize this mode. With a tightaspect ratio (which makes the ideal energy strongly stabilizing) and high S it ispossible for the m = 1 mode to have a tearing mode scaling (7 <x S~3/5) [9].
However, extreme values of S (~10 8 ) and aspect ratio (~1.5) are required toachieve this. Also the average curvature terms, which are stabilizing when
7 oc S~3/5 (provided p' < 0 and q' > 0 at q = 1), tend to be weak because thetoroidal curvature vanishes at q = 1 and also the local aspect ratio is generally quite
large.
It is possible to reduce the growth rate, and in some cases to achieve complete
stability to the resistive m = 1 mode, by reducing q' near q = 1. In practice, thismay be achieved by RF current drive or indeed it may be the profile which the plasmachooses in some cases [12]. Figure 4 shows the growth rate of the n = 1 mode as

296

HENDER et al.

a function of aspect ratio for the profile q = 1 — 0.2 (1 — (p/0.58)2)3, with
S = 105 and 0 = 0. For this q-profile there is a very strong stabilizing effect when
the aspect ratio is decreased. This stabilization is a toroidal effect. For the same
q-profile in the cylindrical limit the m = 1 mode is destabilized when the aspect ratio
is reduced (i.e. the ratio of length to radius). This is partly because in the cylindricallimit the ideal internal kink mode is unstable at zero beta [8]. This method of stabilizing the m = 1 resistive mode appears to be sensitive to the form of the q-profile; forother q-profiles, with q' = 0 at q = 1, we have not been able to achieve complete
stability.

3.

STABILIZATION WITH HELICAL FIELDSIn this section we examine the effects of resonant fields induced by saddle coils
external to the plasma, as studied on Pulsator [13] and TOSCA [14]. Because of thecomplex boundary conditions we use in this case the reduced tokamak MHD equations. Single helicity non-linear calculations show that if we apply a (2, 1) helicalcurrent such that its x-points and the 0-points of an existing (2, 1) plasma island arealigned, then the island width first decreases. However, the final saturated state issuch that the x-points of the saddle coil and of the plasma island are aligned and,because of the influence of the saddle coil, the island width increases slightly. We
can exploit the initial decrease in island width by using a pair of orthogonal coils toactively stabilize the (2, 1) mode. This scheme seems particularly attractive for alocked or slowly rotating ( - 4 0 0 Hz) (2, 1) mode such as that which precedes the
disruption in JET. We have also examined the effect of adding a static saddle coilcurrent to a toroidally rotating plasma. For a (2, 1) saddle coil perturbation which
gives vacuum islands that are significantly smaller than the intrinsic saturated plasma
island, the island continues to rotate and has a slightly enhanced width. For larger
saddle coil currents the plasma island locks to the phase of the imposed perturbationand the island width is slightly increased. Experimentally it has been observed onmachines such as Pulsator [13] and TOSCA [14] that large (2,1) saddle coil currents
will lock the mode and may cause a disruption (and indeed the mode lock observed
on JET may be caused by a similar phenomenon, with the mode locking to small
inhomogeneities in the external fields). Lower currents, however, have been
observed experimentally to stabilize the mode without phase locking. It appears that
this stabilization must be due to effects not included in our simulation (quasi-linear
modifications to the resistivity, toroidal coupling, etc.).

4.

CONCLUSIONS

Various methods for stabilizing MHD instabilities in the tokamak have been
examined. For the m = 2 tearing mode the average curvature (i.e. beta) is strongly

IAEA-CN-47/A-V-3

297

stabilizing — an effect which is particularly pronounced at tight aspect ratio and highconductivity temperatures. The average curvature terms are also stabilizing in the
non-linear Rutherford regime [15]. This stabilization effect, however, depends
strongly on the details of the MHD equations within the resistive layer and may
therefore be affected by kinetic effects.
Local flattening of the current (and q) with RF current drive/heating can stabilize the resistive m = 1 and m = 2 modes. For the m = 1 mode this is a purely
toroidal effect.
The m = 2 mode can be stabilized with an applied (2, 1) helical field, using
an active feedback system of two orthogonal coil sets. A static saddle coil perturbation, with the induced saddle coil island widths comparable to the intrinsic plasma
island, will phase lock a rotating mode; this is in agreement with experiment.

ACKNOWLEDGEMENTS
The authors are indebted to the ORNL MHD Group for providing them witha copy of the FAR code and for their continued collaboration and help. The authors
would also like to thank R. J. Hastie for numerous useful discussions and suggestions.
This work was performed under a Task Agreement between the JET Joint
Undertaking and Culham Laboratory.

REFERENCES[1] WADDELL, B.V., et al., Phys. Rev. Lett. 41 (1978) 1386.
[2] ROBINSON, D.C., McGUIRE, K., Nucl. Fusion 19 (1979) 115.
[3] GLASSER, A.H., et al., Phys. Fluids 18 (1975) 875.
[4] HENDER, T.C., HASTIE, R.J., ROBINSON, D.C., submitted to Nucl. Fusion.
[5] CHARLTON, L.A., et al., J. Comput. Phys. 63 (1986) 107.
[6] O'BRIEN, M., et al., in Controlled Fusion and Plasma Heating (Proc. 13th Europ. Conf.
Schliersee, 1986), Vol. 2, European Physical Society (1986) 270.
[7] ALIKAEV, V.V., ARSENTIEV, Yu.I., BAGDASAROV, A.A., et al., in Plasma Physics andControlled Nucler Fusion Research 1984 (Proc. 10th Int. Conf. London, 1984), Vol. 1, IAEA,
Vienna (1985) 419.
[8] BUSSAC, M.N., et al., Phys. Rev. Lett. 35 (1975) 1638.
[9] HASTIE, R.J., HENDER, T.C., CHARLTON, L.A., HOLMES, J.A., CARRERAS, B.A.,
submitted to Phys. Fluids.
[10] WADDELL, B.V., et al., Nucl. Fusion 18 (1978) 735.
[11] EDWARDS, A.W., et al., Rep. JET-P(86)12, Culham Laboratory, Abingdon (1986).
[12] SOLTWISCH, H., SCHLUTER, J., STODIEK, W., MANICKAM, J., Paper IAEA-CN-47/
A-V-l, these Proceedings, Vol. 1.
[13] KARGER, F., WOBIG, H., CORTI, S., et al., in Plasma Physics and Controlled Nuclear Fusion
Research 1974 (Proc. 5th Int. Conf. Tokyo, 1974), Vol. 1, IAEA, Vienna (1975) 207.

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[14]

ELLIS, J.J., HOWLING, A.A., MORRIS, A.W., ROBINSON, D.C., in Plasma Physics andControlled Nuclear Fusion Research 1984 (Proc. 10th Int. Conf. London, 1984), Vol. 1, IAEA,
Vienna (1985) 363.
KOTSCHENREUTHER, M., et al., Phys. Fluids 28 (1985) 294.

[15]

DISCUSSION
M. PORKOLAB: It is possible that we have already observed the effect you
propose to use for stabilization of the sawteeth, namely to flatten the current profilenear the q = 1 surface. In contrast to results for ASDEX, the inductance <•increases during lower hybrid current drive in ALCATOR C. Nevertheless, the sawteeth are stabilized. We believe that the current is driven just outside the q = 1 surface and that it flattens the current profile there, while reducing the current density
in the outer layers, thereby increasing the self-inductance. The total current remains
constant. This observation is described in my paper1.

1

Paper IAEA-CN-47/F-II-2, these Proceedings, Vol. 1.

IAEA-CN-47/A-V-4

EXPERIMENTAL AND THEORETICALSTUDIES OF CIRCULAR CROSS-SECTION
HIGH BETA TOKAMAKS*
G.A. NAVRATIL, Y. BARANSKY, A. BHATTACHARJEE,
C.K. CHU, A.V. DENIZ, A.A. GROSSMAN, A. HOLLAND,
T. IVERS, X.L. LI, T.C. MARSHALL, M.E. MAUEL,
S. SABBAGH, A.K. SEN, J. VAN DAM1, X.-H. WANG
Plasma Physics Laboratory,
Columbia University,
New York, New YorkM. HUGHES, M. PHILLIPS, A.M.M. TODD
Grumman Corporation,
Princeton, New Jersey
United States of America
Abstract
EXPERIMENTAL AND THEORETICAL STUDIES OF CIRCULAR CROSS-SECTION HIGH
BETA TOKAMAKS.
The paper presents the first results from Columbia University's high beta tokamak, HBT,
demonstrating the production of plasmas with poloidal beta of the order of the aspect ratio,/3p ~ R/a = c~'. Results from a theoretical study of equilibrium, stability and transport in aconceptual second regime experiment, SRX, are also presented. Methods to access the second regime
and the advantages of large aspect ratio, second regime operation for tokamak fusion reactors are
discussed.

1. Introduction
Conventional tokamak experiments have been limited to levels of volume
average beta, </2>, less than fic = 0.035 /p(MA)/oBpwhich corresponds to thetheoretically predicted onset of external kink instabilities and internal ballooning
modes[l,2]. Although it has been shown theoretically that these instabilities
restabilize at higher beta under certain conditions[3-5], no tokamak experimenthas operated in this "second stability" regime. The second regime is characterized
by large values of the poloidal beta, ePp=* 1, and, in this paper, we reportexperimental and theoretical results relevant to these high /3_ plasmas. The* Work supported by United States Department of Energy, Grants Nos. DE-FG02-86ER53222
and DE-FG02-86ER53236, and by Grumman Corporation.
1

Present address: Institute for Fusion Studies, Austin, TX, USA.

299

300

NAVRATIL et al.

o

I

I

I

l \ j l

I

I

I

I

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CVJ

2

I

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d

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I—d—I

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d

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I

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in

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d

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d

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d

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q
d

q
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(LU)Z

in

q
d

1

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IAEA-CN-47/A-V-4

301

experimental results were observed with the HBT tokamak, and they demonstrate
the existence of tokamak equilibrium with ej5p ~ 1 and </J> > f5c for as many as
20 Alfven transit times (tA = qRIVA).
To explore the physics of high (5pplasmas more completely, the theoretical work has described tokamak
equilibrium, stability and transport for a conceptual second regime experiment(SRX) sustained with auxiliary heating.

2. HBT Experiment
The HBT experiment is a pulsed, high-beta tokamak replacing the earlier
Torus II device[6]. HBT can produce very high beta plasmas (</J> < 30%, Bt <0.4T) by rapidly reversing the toroidal field (reverse bias),or lower beta plasmas(</2> < 4%, Bt < 0.7T) by rapidly increasing the toroidal field (parallel bias). A
quartz vacuum chamber surrounds the plasma, allowing fast programming(<2)isec) of the applied magnetic fields. We have operated primarily with circularcross-section plasmas, maximizing pp. By adjusting the magnitude of the initial
toroidal field, the ohmic coil current or the neutral gas pressure, the plasma's
inverse aspect ratio (0.1 < e < 0.3), cylindrical safety factor (0.5 < qcyi < 9) andcollisionality can be systematically changed. Typically, at low to medium
densities of helium or hydrogen ions, we find <«,-> ~ lO^nr 3 , <Te> = <T{> ~25eV, R = 0.24m, a = 0.05m, and </$> = 10%. For these plasmas, both external
and internal ideal MHD modes are predicted to be unstable. The plasma
parameters are measured with a multi-point Thomson scattering system, a twochannel CO2 laser interferometer, infrared bremsstrahlung tomography[7], andinternal and external magnetic probes[8].
2.1. Observation of high j3p plasmas in HBTThe first experiments with HBT have documented plasmas with highpoloidal beta, and they illustrate the outward-shifted poloidal flux surfaces andnon-concentric profile of the parallel current density expected in ideal MHD highbeta equilibria.
Fig. 1 illustrates the results of these measurements for both reverse bias andparallel bias formation methods. The contours of constant poloidal flux andparallel current density are reconstructed from internal magnetic probe
measurements. They clearly indicate shifted poloidal flux surfaces and nonconcentric current density profiles. These measurements can be compared to the
expected magnitude of the Shafranov shift, A, by estimating the poloidal betafrom the condition for major radial force balance

iMlf

=

^ V J Pn<8«/fl) - 1 +Vi + Pp + Hi)/2]-2xRIpBv

(1)

302

NAVRATIL et al.

In Eq. (1), the plasma size (/?, a), mass (M), diamagnetism (//,•), current (Ip), andthe applied vertical field (5V) can be measured as a function of time to determine
the poloidal beta. For the reverse bias case (Fig. la), Eq. (1) gives £/3p=> 0.8
and, for the parallel bias case (Fig. lb), ePp~ 0.75. The expected Shafranov
shift for both examples is of the order of A = 0.2a = 0.01m, comparable withmeasured shifts. When plasmas with lower /L are produced, nearly concentric
flux surfaces are observed as expected[8].
(a) Reverse Bias:f(b) Parallel Bias

\
V":

4

o

7\
CQ.

2 V
0
5
_4
Shafranov q

COCylindrical q

CO

0
150
^J^V 3

V ^/Inductive

S

l>100

< p> = 4%
Plasma Kinetic

50
0
31

H

Plasma Boundary
Magnetic Center

18
rr

24

"cd

17
60

80

100

120

140

160

95

115

135

155

175

195

lisec
FIG. 2. Time evolution of (a) a reverse bias discharge made with 80 mtorr He fill pressure, and(b) a parallel bias discharge with 22 mtorr H2 fill pressure, showing the transition from low e@p tohigh e/3p. For each discharge, the four plots illustrate: (i) the evolution of e@p and (|3)/<3C,(ii) the cylindrical safety factor (qcy, = 2wa2B//t0IpR) and the Shafranov safety factor(°i = Qcyi{l + t2 [1 + 0.5(1/2 + ftp)2}}), (Hi) the total inductive and kinetic stored energies, and(iv) the position of the magnetic center and outer edge of the plasma.

IAEA-CN-47/A-V-4

303

2.2. Investigation of the transition from low to high ef$p
Eq. (1) can also be used to calculate the time evolution of the dischargeparameters, allowing the investigation of the transition to high epp from low e/L,
as summarized in Fig. 2. For the first example (Fig. 2a), the toroidal field isreversed at t = 70}isec, inducing several MA of poloidal currents and coupling
300J into the plasma. After ionization and heating, Wp = 0.37'5PpR\i^p2 ~ 100J,<P> ~ 4/Jc and e(5p ~ 0.5. As the discharge continues, both the inductive stored
energy (Wj = 0.5L/A where L is the plasma's inductance) and Wp gradually
decrease, although Wj is observed to decrease more rapidly than Wp,leading to anincrease in epp. After 30|isec, the safety factor, q, has increased to 2 and epp has
increased to 1. The poloidal flux contours shown in Fig. la correspond to theplasma shown in Fig. 2a at t = 95jisec.
Fig. 2b shows the parallel bias example. With this formation method, arapid compression of the plasma occurs at the same time as the plasma current is
rapidly increased. Although this method couples less total energy to the plasma
as does reverse bias, by reducing the gas fill pressure and substituting H2 for He,
parameters similar to the previous case can be achieved. As shown in the figure,the initially high current dissipates shortly after formation,raising q from slightly
less than one to more than 2.5. Thereafter, the plasma current and energy decay
with a time-constant of =60jasec while maintaining a high beta configuration (ef}p« 0.8, </3> = 2 fic) for more than 80nsec. With Bt = 0.4T and <«,-> = 2 x1021m"3, this corresponds to 20 Alfven transit times.
For both of the examples shown in Fig. 2, the plasma was observed to
evolve without macroscopic instability and to terminate when the resistive decayof the vertical field, Bv, was unable to center the plasma within the vacuum
vessel. Although these experiments establish that high e/L plasmas can be
maintained for times long compared to zA, we do not yet know whether theseequilibria are stable as the plasma's collisionality is reduced. Investigating MHD
stability at higher temperatures and reduced collisionality remains a current
research activity of the HBT experiment.
3.

Large Aspect Ratio, Second Regime Tokamak

When operating in the second stability regime, a large aspect ratio tokamak
can provide significant advantages for a fusion reactor[9]. With large aspect ratio
and circular cross-section, a tokamak can access the second stability regime atlower <fb> than a conventional tokamak. This is seen by examining a simple
estimate of ePp shown in Eq. (2)

p>

(2)

304

NAVRATIL et al.

TABLE I. TARGET PARAMETERS FOR THE SRX EXPERIMENT<p>- -- 3.68%
ePp = 1.16

Rla = 9= 4.2
Q0 = 1.01

5=0
a = 0.175 m(b/d)cn=2 = 0.15-0.2(Wa)cn=3 = 0.05-0.1

qa=t UP a u r = 2-3MW/ p = 51.2kA(Z?/a) c n > 3 = ooq(Y)' ^ 2 . 5

BT= IT(£>/a)cn=1 = 0.25-0.3

To minimize the <p> required for second regime access (ePp ~ 1), we should
reduce e, choose a circular cross-section ( K = 1), and operate with high q. The
results of equilibrium, stability and transport calculations for a second regime
experiment (SRX) based on this concept are described below. In addition,
extensions to the theories describing transition into the second regime from low
beta using either energetic particles or toroidal rotation are discussed.
3.1. Equilibrium and Stability Boundaries
Ideal MHD stability analysis has defined the target parameters for thesecond stable region of SRX, and these are shown in Table I. Low toroidal mode
number, n, ballooning modes, driven by the pressure gradient in the low shear
region, require a close fitting conducting shell, 0.5 < (b/a)cn~^ < 0.10, forstability [10]. Here,(6/a) is the shell displacement from the plasma edge
normalized to the plasma radius. Raising the axis shear to balance high and low ninternal instabilities would increase the critical wall position. The fact that modes
with n > 1 set the minimum requirement, and that (b/a)c is observed to increase
rapidly for all modes when epp > 1, indicate that the pressure gradient rather thanthe parallel current is the dominant instability driver.
Current or safety factor, q(y/), programming has been suggested as amethod to improve second region access[10]. We have demonstrated stable
access for all ideal MHD modes in SRX when q0 < 3.1, qa = 4.1, (b/a) > 0.05,
p(\j/) oc ^2.5 ( anc j q(yfy oe y/ 4.0 Again, stability of the « = 3 mode defines theminimum wall position as ePp is increased. The minimum wall requirement
applies only in the middle of the unstable region. Restabilization, as describedabove, is again observed at higher ePp. Even in the absence of a completely
stable access path, current programming can significantly reduce the ePp required
for second region operation (e.g. epp < 0.7 for q0 „, 2). This technique could
then be used in conjunction with the other mechanisms discussed in Section 3.3.

305

IAEA-CN-47/A-V-4

3.2. Transport Modeling and Power Requirements
A flux surface averaged transport code[ll] has been used to estimate theauxiliary heating power necessary to obtain the second regime SRX design point
plasma of Table I with neutral beam injection. The transport model includes
neoclassical coefficients supplemented by a microinstability model of anomalous
diffusion[12]. Similar models have been used to match PBX[13] and TFTR[14]results. Electromagnetic effects (e.g. ballooning modes), which are expected toplay an important role at high <fi>, are not included in the present transportP'-2.00E+06 •1.00E+06
0.00 0.10 0.20 0.30 0.40 0.50 0.60 0.70 0.80 0.90

1.00

•1.00E + 06
0.00 0.10 0.20 0.30 0.40 0.50 0.60 0.70 0.80 0.90

1.00

FIG. 3. Profiles qfp': (a) calculated from a 1.5-D transport code with 5 MW of neutral beam power,and (b) corresponding to the stable second regime design point of Table I.

306

NAVRATIL et al.

model. However, the analysis does provide an estimate of the required power for
an SRX plasma with a stable access path to the second regime. The confinement
times obtained from this code are similar to those predicted by L-mode scaling.
Shown in Fig. 3a is the p' profile generated by the transport code and in Fig. 3b
the p' profile for the stable design point equilibrium. The corresponding pressure
profiles for these two cases are nearly identical. Our results suggest that a device
based on the design parameters of Table I can exceed j5c by a substantial factor(>4) and that the target parameters can be achieved with a neutral beam power of
2 to 3 MW.
3.3. Stabilization Mechanisms for Second Regime Access
We have examined and extended two proposed stabilization schemes fortransition from the first to the second regime: (1) energetic particle stabilization,and (2) induced toroidal rotation[9]. The original work on energetic particle
stabilization showed that highly energetic particles introduced into the badcurvature region of a tokamak could improve the stability of the system against
ideal ballooning modes[16]. We have extended this work to include finite aspectratio effects in an analytic equilibrium with scalar pressure and an anisotropic
energetic particle pressure. Even without aspect ratio corrections, an additional
region of ballooning instability exists beyond the so-called "drift reversal line"(see Ref. 16). When the energetic particle beta is comparable to the plasma beta,
finite aspect ratio effects can completely remove the original ballooning unstableregion, but the new instability region persists. This may limit the magnitude ofthe local p' in a high beta plasma stabilized using this technique.
A WKB theory has been developed for n = °° ballooning modes in an
equilibrium with large rigid toroidal rotation, and sufficient conditions forstability, which extend earlier work[17], have been obtained. For large flow
velocities on the order of Mach 1, a possible stabilization mechanism is provided
by the centrifugal separation of the surfaces of constant mass density andpressure. In the central part of the equilibrium, the density gradient caused by thecentrifugal force may offset the destabilizing effects of the pressure gradient.
4.

Fusion Reactor Applications

In order to quantify the differences between a conventional first regime
reactor and a large aspect ratio second regime reactor, zero-dimensional
calculations based on a model developed by Uckan and Sheffield[18] have beenused to directly compare the two configurations. By limiting the magnetic field at
the toroidal field coil to 12T, limiting the neutron wall loading to 5MW/m2, andconsidering "Starfire-sized" reactors with net electrical output of 1.2GW, very
similar operating parameters and costs were observed for first regime reactors

IAEA-CN-47/A-V-4

307

with elongation and triangularity sufficient to raise fic > 6% (K => 2, 8 ~ 0.4) andfor second regime reactors with large aspect ratios (R/a > 6). However, the large
aspect ratio reactor can operate (1) at half the field strength at the toroidal fieldcoils, allowing the use of NbTi superconductors, and (2) with 16 times lessinductive energy stored in plasma current,reducing the potential for disruptioninduced damage. The high </J> and magnetic field utilization of the large aspectratio second regime reactors also permit cost effective compact reactors (with net
electrical output of the order of 300MW) and may also make a reasonably sized
tokamak reactor burning advanced fuels feasible. Indeed, by increasing </fc> to
30% and B, to 10T, </3>2£,4 can increase by a factor of 150 over that needed for
DT fusion. The zero-dimensional reactor model has also been used to estimate
the confinement needed for DT ignition by extrapolating present empirical scalinglaws to high beta and large aspect ratio. As in first regime reactors, it is observedthat only a limited subset of these scaling laws predict ignition for reasonably
sized reactors, and these have been used to specify the confinement which must
be observed in the conceptual SRX experiment to demonstrate trends leading to
ignition[19].

REFERENCES
[1]
[2]
[3]
[4]
[5]
[6]
[7]
[8]
[9]
[10]
[11]
[12]
[13]
[14]
[15]
[16]
[17]
[18]
[19]

BERNARD, L.C., HELTON, F.J., MOORE, R.W., Nucl. Fusion 23 (1983) 1475.
TROYON, F., GRUBER, R., SAURENMANN, H., SEMENZATO, S., SUCCI, S., Plasma
Phys. 26 (1984) 209.
MERCIER, C , in Plasma Physics and Controlled Nuclear Fusion Research 1978 (Proc. 7th Int.
Conf. Innsbruck, 1978), Vol. 1, IAEA, Vienna (1979) 701.
CHANCE, M.S., JARDIN, S.C., STIX, T.H., Phys. Rev. Lett. 51 (1983) 1963.
COPPI, B., CREW, G.B., RAMOS, J.J., Comments Plasma Phys. Controll. Fus. 8 (1983) 11.
CHU, C.K., DENIZ, A., GEORGIOU, G.E., GROSS, R.A., GROSSMANN, A.A., et al.,Nucl. Fusion 25 (1985) 1109.
HOLLAND, A., NAVRATIL, G.A., Rev. Sci. Instrum. 57 (1986) 1557.
DENIZ, A.V., MARSHALL, T.C., IVERS, T., LI, X., MAUEL, M., Phys. Fluids 29 (1986)
4119.
NAVRATIL, G.A., MARSHALL, T.C., Comments Plasma Phys. Controll. Fus. 10 4 (1986).
TODD, A.M.M., et al., Paper IAEA-CN-47/E-III-ll, these Proceedings, Vol. 2.
BATEMAN, G., Proc. International Centre for Theoretical Physics, Trieste, 1985.
TANG, W.M., Rep. PPPL-2311, Plasma Physics Lab., Princeton Univ., NJ (1986).
JARDIN, S.C., et al., Rep. PPPL-2358, Plasma Physics Lab., Princeton Univ., NJ (1986).
REDI, M., Rep. PPPL-2368, Plasma Physics Lab., Princeton Univ., NJ (1986).
KAYE, S.M., GOLDSTON, R.J., Nucl. Fusion 25 (1985) 65.
ROSENBLUTH, M.N., et al., Phys. Rev. Lett. 51 (1983) 1967.
HAMEIRI, E., LAURENCE, P., J. Math. Phys. 25 (1984) 296.
UCKAN, N.A., SHEFFIELD, J., Rep. ORNL/TM-9722, Oak Ridge National Lab. (1985). Alsoin Tokamak Startup (KNOEPFEL, H., Ed.), Plenum Press, New York and London (1986) 45.
MAUEL, M.E., submitted to Nuclear Fusion.

308

NAVRATIL et al.

DISCUSSION
D.C. ROBINSON: In your HBT experiments with €/3p ~ 1, was the discharge
characterized by central q values greater or less than one?
G.A. NAVRATIL: We have produced discharges with central q greater thanone, but for those discussed here the central q was near one.
D.C. ROBINSON: Could I also ask how you account for the absence of strong
MHD instabilities in your plasmas with e/3p ~ 1, given such central q values?
G.A. NAVRATIL: We have computed the stability of equilibria similar to theseand expect them to be unstable. Two effects which may reduce the growth rate ofthese modes come from the collisional nature of these plasmas and the presence ofa conducting halo plasma with finite pressure.

IAEA-CN-47/A-V-5

DISRUPTION CONTROL EXPERIMENTS
USING LOCAL MODULAR
MULTIPOLE FIELD COILS
K. YAMAZAKI, K. KAWAHATA, Y. ABE, R. AKIYAMA,
T. AMANO, R. ANDO, Y. HAMADA, S. HIROKURA,
K. IDA, E. KAKO, O. KANEKO, Y. KAWASUMI,
S. KITAGAWA, T. KURODA, K. MASAI, K. MATSUOKA,
A. MOHRI, M. MUGISHIMA, N. NODA, I. OGAWA,
Y. OGAWA, K. OHKUBO, Y. OKA, K. SAKURAI, M. SASAO,
K. SATO, K.N. SATO, S. TANAHASHI, Y. TANIGUCHI,
K. TOI, T. WATARI
Institute of Plasma Physics,
Nagoya University,
Nagoya, Japan
Abstract
DISRUPTION CONTROL EXPERIMENTS USING LOCAL MODULAR MULTIPOLE FIELD
COILS.
In the JIPP T-IIU tokamak, major disruptions have been successfully suppressed by using compact local sets of I = 3 modular multipole field coils with an m = 3/n = 2 dominant helical component.
It is found that stabilization can be attributed to the excitation of a 'mini-disruption' induced by the 3/2
helical field perturbation before the catastrophic stage that leads to a major disruption. Because of thequick but mild ergodization of the m = 2/n = 1 magnetic islands by this mini-disruption, temperature
flattening is impeded and a two-step disruption that would lead to disruptive current termination is
avoided. This system is easily applicable to larger tokamaks because of compact coil utilization, slow
pulse application and quick stabilization response. It has also been demonstrated that this coil system
can be used for edge plasma control with an ergodic limiter configuration.

1.

INTRODUCTION

The usefulness of the superposition of weak external helical field perturbations
on tokamak plasmas for stability studies [1] and transport studies [2] has alreadybeen widely recognized. The stabilization of the m = 2/n = 1 island by an
m = 2/n = 1 resonant helical field (m and n are the poloidal and toroidal mode
numbers) with gradual current profile modification or fast in-phase feedback schemes
has been demonstrated. Stabilization by a non-resonant helical field has also been
tried with the production of an ergodic region outside the q = 2 surface. In contrast
to these previous experiments, in our experiments we investigated the effect ofergodization inside and/or outside the q = 2 surface. We found a new, quick stabilization process through excitation of a small disruption ('mini-disruption') induced by
309

310

YAMAZAKI et al.

the helical field. In contrast to a configuration with complicated purely helical coils
or saddle coils, the coil configuration used here is rather compact. With this coil set,
tests of both core ergodization for disruption control and edge ergodization forergodic limiter experiments have been carried out.

2.

HELICAL FIELD CONFIGURATION

Two sets of toroidally localized multipole (I = 3) field coils ('local helicalcoils', Fig. l(a)) have been used; they are mounted on two vacuum vessel sectors of
30° at opposite sides of the torus. These coils were installed initially with the aimof achieving m = 3/n = 1 ergodic limiter configurations [3]. When the coil currents
are driven in the anti-parallel direction, an m = 3/n = 2 helical field with 1/1, 2/2
and 4/2 side-band components (A (anti) mode: core ergodization mode) is produced;
when the coil currents are driven in the parallel direction, an m = 3/n = 1 field with
2/1,4/1 and 4/3 components (D (dual) mode: edge ergodization mode) is produced(Fig. l(b».

PlasmaVesse|

100

FIG. 1. (a) Configuration of the local helical coil, (b) Radial field components of the local helical coilfor an A-mode pulse with an m = 3/n = 2 dominant component (Ij = —I2), and for a D-rnode pulse
with an m = 3/n — 1 dominant component (1/ = I2), with Ihe[ = 75 hi.

IAEA-CN-47/A-V-5

311

The JIPP T-IIU machine (R = 91 cm, a = 24 cm) was operated at
I T < Bt s 3 T and 100 kA < I,, =s 300 kA. The applied 3/2 radial perturbation
field strength at the plasma surface, Br3/2(a), is about 30 G at a maximum helicalcoil current Ihe, of 15 kA (Fig. l(b), left). The ratio of this perturbation field to thepoloidal field of the plasma current, Br3/2(a)/B^(a), is about 1.0-1.5%.

3.

DISRUPTION CONTROL EXPERIMENTS:
Pulse application to unstable discharges

The changes in MHD activity during pulse application are measured mainly bysingle, 10- and 20-channel Mirnov coil poloidal arrays, a 10-channel grating polychromator receiving second harmonic electron cyclotron emission (ECE), and a
7-channel soft-X-ray (SXR) pin diode poloidal array.
The typical discharge waveforms with and without pulse application to themildly unstable discharges are shown in Fig. 2. Figure 3 is a plot of the time variations of the electron temperature profile T^ 5 (r), before pulse application, without apulse and with a pulse; third-order spline-fitting curves are used; the current profilej(r) can be roughly estimated from the electron temperature profile (j(r) oc T|-5(r)).
Without pulse application (Fig. 2(a)) the plasma current (q^ « 3.2-3.4 andne « 2 x 10"13 cm-3) is terminated by a few pre-disruptions and by the subsequentmajor disruption. In this case, the discharge proceeds with successive phenomenacharacterized by the stages of initial 'precursor oscillation', 'soft mode locking','pre-disruptions', 'major disruption' and 'current quench' (Fig. 2(a)). The peaked
electron temperature profile at t = 90 ms is changed to the flattened profile near thecentre by the onset of precursor oscillations (Fig. 3(a)). During the precursor oscillation phase, the strong 2/1-mode toroidal 'co'-rotation (rotation in the plasma current
direction) gradually comes to a standstill and then leads to the pre-disruption.
According to the magnetics, SXR and ECE temperature fluctuation measurements,a pure m = 2/n = 1 mode (R « 100 cm) accompanied by 1/1 deformation(R « 88 cm) grows, and an m = 1/n = 1 island quickly develops with double mode
frequency on this 1/1-deformed surface just before the pre-disruption. The ECE temperature decay during the pre-disruption is initiated mainly near the q = 2 surface
as a result of 2/1-1/1 interaction, as predicted by theory [4]. The island width of the
2/1 mode is not reduced by this pre-disruption and finally the plasma current is terminated through the two-step disruption process. Although the T]-5(r) profile does notcorrespond exactly to the the j(r) profile, it is suggested from the flat profile(Fig. 3(b)) that the central qg before the major disruption in the case without a pulse
is larger than 1.3 [5]; this is consistent with the theoretical prediction of 'bubble'
formation [6].
When an A-mode pulse (m = 3/n = 2 dominant) is applied during the unstable
period (Fig. 2(b)), the m = 2 mode toroidal rotation gradually comes to an end and312

YAMAZAKI et al.

> >

}
JD1.

-i-

8
>*

•of

IAEA-CN-47/A-V-5

313

(a)

0.7

0.9
1 i
R(m)

(Te(keV)]' /
1.5
'

116.5

FIG. 3. Time evolution of the one-half power of the electron temperature profile T'e-5: (a) before pulse
application (t < -100 ms), (b) without a pulse (t > -100 ms) and (c) during an A-mode pulse(t > -100 ms).

is finally more tightly locked in phase than it is without pulse application. A minidisruption is excited before the above-mentioned pre-disruption in the case withouta pulse; this mini-disruption is characterized by a mild and global change in theplasma profile. The 2/1 magnetic island disappears in this mini-disruption stage,
whereas it did not disappear in the pre-disruption stage when no pulse was applied.

314

YAMAZAKI et al.

After one or two mini-disruptions the plasma current channel becomes more peaked
than in the case without a helical pulse, which often leads to the occurrence of small
sawtooth oscillations. In contrast to the situation existing in the case without a pulse,
the profile after the mini-disruption induced by the A-mode pulse remains almost
constant, except just during pre-disruption, and does not change to the flat profile(Fig. 3(c)). This quick rearrangement of the magnetic configuration due to thefavourable mini-disruption helps avoid disruptive current terminations.
The D-mode pulse (m = 3/n = 1 dominant), however, is not effective insuppressing major disruptions and sometimes the discharge is destabilized. This may
be due to the fact that the 3/1-dominant field does not directly affect the 2/1-1/1 interaction and that it causes undesirable disturbances on the q = 3 surface or the plasmaedge surface.

8

4

0
240

120

r

Ip(kA)
4
2
f

100
time (ms)

2 00

Te«ntre(au)

a.u.

100

150
time(ms)

200

FIG. 4. Ergodic limiter experiments with application of a D-mode pulse.

IAEA-CN-47/A-V-5

4.

315

ERGODIC LIMITER EXPERIMENTS:
Pulse application to stable discharges

Helical magnetic field perturbations are also applied to stable discharges withq^ ~ 3-4 in order to clarify the effect of ergodization on the transport process.
When a D-mode pulse is applied, a hard-X-ray burst due to loss of runaway electrons,a decrease of the C-IV, C-V, O-V and Fe-IX line intensities, and a decay of edge
ECE temperature are observed, with a time delay At! after pulse application (seeFig. 4). Ergodization of the edge confinement region and a resultant favourable effectof impurity influx suppression in the time interval Atj are suggested. The quick
onset and duration of this ergodization depend on the pulse height and the width of
the helical perturbations, which suggests the existence of an ergodization threshold.
On the other hand, an A-mode pulse does not significantly influence the stable
discharges, even with a maximum coil current of 15 kA.

5.

CONCLUSIONSIn the JIPP T-IIU tokamak experiment it has been demonstrated that them = 3/n = 2 dominant helical field perturbation produced by toroidally localized
multipole field coils induces one or two mini-disruptions in the tokamak plasma andthat the 2/1 magnetic island is ergodized by this mini-disruption. Temperature flattening is interrupted and two-step disruption is avoided. With this scenario, a compact
local helical coil system producing 'dirty' spectral field components is used insteadof complicated purely toroidal helical coils. This may be acceptable in a large tokamak because it reduces the probability of disruptive current terminations throughslow feedback.
Also, preliminary ergodic magnetic limiter experiments have been carried out
using the local helical coil system with an m = 3/n = 1 dominant helical component.
The favourable effect of edge ergodization on the edge temperature drop has been
confirmed and the effectiveness of this coil configuration for transport and stability
control verified.
REFERENCES
[1]
[2]
[3]
[4]
[5]

[6]

KARGER, F., WOBIG, H., CORTI, S., et al., in Plasma Physics and Controlled Nuclear Fusion
Research 1974 (Proc. 5th Int. Conf. Tokyo, 1974), Vol. 1, IAEA, Vienna (1975) 207.
OHYABU, N., et al., Nucl. Fusion 25 (1985) 1684.
KAWAMURA, T., et al., J. Nucl. Mater. 111&112 (1982) 268.
TURNER, M.F., WESSON, J.A., Nucl. Fusion 22 (1982) 1069.
YAMAZAKI, K., et al., Quick Profile Reorganization Driven by Helical Field Perturbation for
Suppressing Tokamak Major Disruptions, Res. Rep. IPPJ-796, Inst. of Plasma Physics, Nagoya
Univ. (1986).
KLEVA, R.G., et al., Phys. Fluids 29 (1986) 457.

316

YAMAZAKI et al.

DISCUSSION
K. McGUIRE: In your A-mode type stabilization of the major disruption, could
you improve the density limit or the low q^ limit?
K. YAMAZAKI: No, we have not yet managed to extend the operating region.
The density limit was somewhat improved, but not substantially. The reason why thepresent system was not effective is that in the case of low qyhigh density the growthof the MHD mode is too fast to be controlled by the present slow pulse coil systemoutside the vacuum vessel. Another reason might be that the disruption mechanism
for low q/high density is different from the 2/1 - 1/1 interaction of mildly unstablemedium q and medium density discharges. However, we are still trying to improve
the operational regime by using the present 'mini-disruption' scheme with other helical field components.
D.C. ROBINSON: I noticed that your A-mode coil configuration (m = 3,
n = 2) had a significant m = 1, n = 1 component. Did you observe any significant
effect of this coil configuration on sawtooth activity?
K. YAMAZAKI: We applied the A- and D-mode helical pulses to stable OH
and ICRF heated discharges with sawteeth, but did not observe any significant
changes in sawtooth behaviour. We think the m = 1/n = 1 component is still too
small to modify the sawtooth activity. We plan, however, to add new coils and to
perform further sawtooth stabilization experiments.

IAEA-CN-47/A-V-6

STUDIES OF MHD MODESAND HIGH FREQUENCY FLUCTUATIONSON THE HT-6B AND HT-6M TOKAMAKSJikang XIE, Jiayu CHEN, Wenkang GUO, Yuping HUO,
HT-6B GROUP, HT-6M GROUP
Institute of Plasma Physics,
Academia Sinica,
Hefei, Anhui, China
Abstract
STUDIES OF MHD MODES AND HIGH FREQUENCY FLUCTUATIONS ON THE HT-6B AND
HT-6M TOKAMAKS.
Detailed studies of MHD modes have been performed on the HT-6B tokamak by means of weak
resonant helical fields (RHFs). The experimental results show that the MHD behaviour cannot be
explained by the current tearing mode theory. A very interesting point is that a weak RHF (about 0.5%of Bp) could significantly influence sawtooth processes and thereby change the global thermal properties of the plasma. A series of voltage pulses were applied to HT-6M plasma. The investigations
included low frequency MHD modes, high frequency magnetic fluctuations, H o line emission and ECE
signals. High frequency fluctuations have been suppressed by the voltage pulses.

1.

INTRODUCTION

Plasma parameters for conditions close to ignition have already been achievedin tokamaks. However, there are still many important processes of the plasma besides
transport which are not yet clear. There are many reports on high frequency fluctuations, but no conclusion has been reached as to their physical origin, their relation
with other plasma processes and the way of control. Low frequency MHD perturbations, which are directly related with plasma disruptions and which have an influenceon the global thermal properties of the plasma, have long been regarded as tearing
modes [1]. Recently, this theory has been questioned because of some experimental
results [2]. It seems necessary to find a new explanation, also for sawtooth processes.
Therefore, a careful study of such processes is certainly of value for the further
development of tokamaks.
To obtain more information and to investigate different aspects of MHD modes
and high frequency fluctuations, we studied the changes of these modes under the
influence of external disturbances [3, 4]. When I = 2 and ( = 3 weak resonant
helical fields (RHFs) were used as active magnetic probes, many new phenomena
could be observed; these seem to give a physical picture for the low frequency modes
which is totally different from the usual one. Also, when skin current pulses wereapplied, some new properties of high frequency fluctuations were found.
317

318

XIE et al.

The HT-6B is an air-core transformer tokamak with i = 2 and i = 3 windings
on its vacuum chamber. The operational parameters are: R = 45 cm, a = 12.5 cm,
3B, = 6-8 kG, Ip < 40 kA, n,. = (0.7-3) x 1013 cm "= 150-250 eV,
Zeff < 2. The diagnostics of HT-6B are nearly all conventional, including
multichannel soft-X-ray fluctuation measurements, multichannel spectroscopy, fivechannel ECE, cos m$ coils and more than twenty magnetic probes.
The HT-6M is a new air-core transformer tokamak with the following
parameters: R = 65 cm; a = 20 cm; B, = 15 kG, Ip < 120 kA, Zeff = 1.5-2.5,
Te = 400-700 eV, ii,, = (0.9-2) x 1013 cm" 3 . It is equipped with I = 3 helicalwindings and a second Ohmic heating winding just outside the vacuum chamber.

2.

MHD MODE STUDIES WITH A RESONANT HELICAL FIELD

The helical windings on HT-6B can produce steady or pulsed RHFs with amagnitude of 0.5-5% of Bp. Without an RHF we observed all the typical MHD
phenomena which appear also on other devices. The main results of experiments with
RHFs are as follows:(a) The fact that different n,m modes always have the same frequency has long
been observed on all tokamaks including the HT-6B. Normally, this is explained by
the rotation of the magnetic islands which can be observed by the phase shifts ofsignals of Mirnov probes located at different poloidal positions. On HT-6B, the rotation of the magnetic island was stopped by a weak RHF, as shown in Fig. 1; the RHF
was turned on at 6 ms and turned off at 12 ms. In some cases, non-rotating Mirnov
signals or signals without any phase relation were observed even without an RHF.
Obviously, this means that the frequency (about 20 kHz on HT-6B) is not entirely
due to magnetic island rotation.

(b) I

T

I

FIG. 1. Phase shifts on adjacent Mimov probes,(a) without RHF and (b) with RHF.

319

IAEA-CN-47/A-V-6

FIG. 2. Suppression of m = 2,3,4,5 modes, and enhancement of sawteeth
on soft-X-ray and ECE signals by an I = 3 RHF.

-I
FIG. 3. Changes ofm—2 and m = 3 modes, soft-X-ray signals
and ECE signals when ant-2
RHF is turned on.

(b) The weak I = 3 RHF of about 1% Bp(a) can suppress all m = 2,3,4
modes effectively when q(a) > 3 (Fig. 2). Oscillations of the same frequency in softX-ray and spectroscopic signals are also suppressed. In the case of a weak
( = 2 RHF with q(a) > 2.5 the same phenomena could be observed (Fig. 3). It
should be emphasized that our RHFs were rather weak so that only the magnetic
structure close to the corresponding resonance surface could be changed. This local

320

XIEetal.

t y^js\_/\y\>^^J=3 RHF on

400us

FIG. 4. Reduction and relaxation of sawteeth on soft-X-ray signals
to 20 kHz Mimov oscillations when the RHF is turned off.

Weak J = 3 RHF on

FIG. 5. Application of a strong 1 = 2 RHF to an t = 3 stabilized plasma,causing a new mode on cos m6 coil signals.

resonant interaction of the RHF with the plasma has been proven by two facts:(1) when the direction of the toroidal magnetic field was reversed, the weak helicalfield could not influence the plasma fluctuations; and (2) in the case of q(a) < 3,
the I = 3 helical current could not suppress the MHD modes. Since the magnetic
islands of different m-modes did not overlap in our experiments, it is difficult to
explain why such localized disturbances could suppress the m = 2,3,4,5 modes,
each of which is concentrated only near the corresponding resonant surface according
to tearing mode theory.
(c) The sawtooth oscillations which appeared on HT-6B in soft-X-ray signals,
ECE signals and, in some cases, even in magnetic signals, were dramaticallymodified by a weak RHF. Mostly, an RHF (I = 3 or I = 2) increased the sawtoothIAEA-CN-47/A-V-6

321

amplitude and prolonged the repetition period, as shown in Fig. 2. The ECE signals
showed an increment of the central electron temperature. In extreme cases, an RHF
moved the inverse surface of a sawtooth (q = 1 surface) outwards, and the sawtoothoscillation appeared even in the soft-X-ray signal of the edge chord. In somedischarge regimes without an RHF there was no sawtooth oscillation, only a 20 kHz
fluctuation in the soft-X-ray signals. After the weak RHF was turned off, the large
amplitude sawteeth disappeared with a 1 ms delay (see Fig. 4). All these facts seemto indicate that the very localized changes of the magnetic structure caused by a weak
RHF could bring about a strong resonant change of the energy process of the plasma,
especially in the inner part.
(d) A relatively strong ( = 2 RHF in some cases increased the 20 kHz oscillation and suppressed sawteeth, but did not disrupt the discharge. When the I = 2 RHF
was larger than 2% Bp(a), the discharge deteriorated immediately. An experiment
was performed in which two RHFs (I = 2 and ( = 3) were combined. The MHD
fluctuations were stabilized by a suitable I = 3 RHF; in this way the discharge could
be prevented from suddenly turning on a strong I = 2 RHF, as shown in Fig. 5.
When the I = 2 RHF was turned on, there were suddenly more m = 2 and m = 4
perturbations, but the t = 3 mode remained stable.
(e) Similar effects on RHFs have also been observed on the HT-6M tokamak.

3.

SKIN CURRENT PULSE EXPERIMENTS ON HT-6M

In an attempt to change the edge conditions, four voltage pulses (V-pulses),
0.2-1 ms wide, with 10-100 V loop voltage, were applied to the plasma during thedischarges via the second Ohmic heating winding. Pulse currents, Ip, of up to 15 kA(20% of Ip) were induced in the co-direction and the counter-direction with regard
to Ip (Fig. 6). Some of the preliminary results are as follows:(a) The amplitudes of the m = 2,3,4,5 MHD modes did not change muchduring the pulse process with Ip up to 15% of Ip (Fig. 7). When Ip was slightly
larger than 15% Ip, the amplitudes of the m = 2 mode were enlarged, but those of
the m = 3 mode were not enlarged. Usually, the discharge was interrupted immediately after a fast rise of the m = 2 mode amplitude, with Ip more than 20% of Ip.
It was found that the instant of the enhancement of the m = 2 mode or the low
frequency magnetic oscillation Br (see below) was at the set-up of a positive electricfield of the V-pulse, i.e. in the direction of Ip.
(b) A broad-band high frequency magnetic fluctuation was observed by a small
Br coil close to the edge of the plasma. The amplitude of this fluctuation was
Br/Bp(a) = 10~5 to 10 " 4 . In many cases there was a large component of about
250 kHz. Suppression of these fluctuations by a V-pulse was observed (Fig. 8). After
the V-pulse was switched on, the high frequency fluctuation Br gradually disappeared, but a low frequency (about 10 kHz) magnetic oscillation began to grow;

XIE et al.

322

FIG. 6. Discharge with V-pulse, plasma current Ip (20.4 kA/div.), loop voltage VL (6.26 V/div.),
position of plasma Dh (10.5 kA-cm/div.) and electron density Ne (3.5 x IO12 cm~3/div.).
The dashed line denotes the position of the plasma without a V-pulse, time 3.84 ms/div.

4b

J1*

m=2
* » • •4-r

r

—•

m=3
•«••

m=4

m=5

4*
FIG. 7. Waveforms of the m = 2,3,4,5 MHD modes
during a V-pulse, time 200 fis/div.

I
m=2
••*

i

f
1

it

HFM

•ill

k

III

l!

M
*<

/7G. 8. Waveforms of the m = 2 MHD mode am* //je WgA frequency
magnetic fluctuation (HFM) during a V-pulse, time 200 jisldiv.

IAEA-CN-47/A-V-6

323

this was not correlated with the MHD modes mentioned before. The amplitude of thisoscillation increased with increasing voltage.
(c) The horizontal displacement of the plasma underwent a shift during thepulse processes and did not recover after the V-pulse had passed by (Fig. 6). The
delay of the shift may be related to the slight change in the profile of the plasma
current density.
It was found that the Ha line emission from the edge as well as the line densityof the plasma did not increase during the V-pulse process. The edge heating effectwas observed by ECE measurements.

4.

DISCUSSION

Low frequency MHD oscillations are very complicated phenomena which
depend sensitively on the discharge conditions, such as electron density, impurities
and wall conditions. Also, the discharge conditions have an influence on theeffectiveness of the RHF. Although the experimental RHF results discussed above
can only be reproduced in some discharge regimes, they show that the observed'MHD oscillations' cannot be properly explained by the tearing mode theory;
however, they give some clues to finding the mode structures.
Our experimental RHF results can be partly explained by assuming that theobserved mode was a single mode which extended through the whole plasma. Froma recent study of profile consistency it became clear that the RHF strongly influences
sawtooth processes through this mode.
Our skin current experiments also showed the complexity of the low frequency
magnetic fluctuations. We still do not know the real nature of the observed 10 kHz
mode, but it is obvious that the high frequency fluctuation is of electromagnetic
origin.
REFERENCES
[1]

[2]
[3]
[4]

WHITE, R.B., MONTICELLO, D.A., ROSENBLUTH, M.N., WADDELL, B.V., in Plasma
Physics and Controlled Nuclear Fusion Research 1976 (Proc. 6th Int. Conf. Berchtesgaden,1976), Vol. 1, IAEA, Vienna (1977) 569.
CAMPBELL, D.J., GILL, R.D., GOWERS, C.W., et al., Sawtooth Activity in OhmicallyHeated JET Plasmas, Rep. JET-P(86)05, JET Joint Undertaking, Abingdon, Oxfordshire (1986).
ZHAO, Quingchu, CHEN, Jiayu, XIE, Jikang, et al., in Plasma Physics and Controlled Nuclear
Fusion Research 1984 (Proc. 10th Int. Conf. London, 1984), Vol. 1, IAEA, Vienna (1985) 345.
HT-6B GROUP, Paper ASIPP/11, presented at 12th Europ. Conf. on Controlled Fusion andPlasma Physics, Budapest, 1985.

IAEA-CN-47/A-VI-l-l

ANOMALOUS TRANSPORT IN JET PLASMAS
D.F. DUCHS, T.E. STRINGER, A. TARONI, M. BRUSATI,
N.A. GOTTARDI, T. HELLSTEN, F. TIBONEJET Joint Undertaking,
Abingdon, Oxfordshire,
United Kingdom
Abstract
ANOMALOUS TRANSPORT IN JET PLASMAS.
Local thermal fluxes have been determined experimentally for ohmic, RF wave and neutral beamheated JET plasmas through direct power balance calculations and through a modified application of'predictive' equilibrium transport codes. A realistic way of describing the RF power deposition profile
has been developed. Free-boundary plasmas, important for magnetic limiter cases in JET, can now be
computed fully self-consistently. The experimental local fluxes have been compared with available
analytically derived fluxes, but no acceptable agreement could be found. Many discharges can,however, be modelled empirically by modifying the electron thermal transport equation so that thesteady state T e profile is constrained to fit the observed canonical shape, with both ohmic andadditional heat sources.

1.

INTRODUCTION

In this paper we attempt to assess the status of theoreticalunderstanding of anomalous transport in JET plasmas.
This isdone by deriving local fluxes from measurements and comparing
them with presently available
theoretical ones which areapplicable under JET conditions.
Sections 2 and 3 present theexperimental values which are obtained firstly by the usual
method of the integrated power balance and secondly by applying
a full predictive transport code in steady state.
The chosencases include JET discharges where the plasma is heatedohmically, by ion cyclotron waves, and by neutral injection.
Those discharges have been selected where a maximum number ofdiagnostics were available with reliable results so that theuncertainties on the fluxes could be minimised. In Section H, anumber of proposed theoretical formulae are reviewed and checkedagainst JET data. Finally, in Section 5, an empirical approach isoutlined by which also the plasma evolution in time is modelled
successfully. The highlights are summarised in Section 6.
2.

FLUXES FROM INTEGRATED POWER BALANCE

If V. denotes the volume inside a chosen magnetic flux
surface S., we define the average primary electron thermal energy
flux density at time point t asfUt) =

i
325

326

DUCHS et al.

consisting of a convective and a conductive term. We prefer thisquantity f e rather than x e (or X p because it is closer to theactual measurements and avoids special assumptions and errors
connected with local gradients.
The flux is calculated as afunction of i^ from the power balancefe(i|>,t) = 1|- i_ /

i n e T e dV

dV

- /

na _ H — I dV

PRF

dv + ;

PNBI

dV}

(2)

An analogous equation holds for f ^ ^ . t ) . Additional possible
terms
(e.g.
ionisation, adiabatic
compression, etc.) were
estimated as negligible. The geometry, i.e. ty, S^ and V.,aretaken from solutions of the Grad-Schltlter-Shafranov equation
which agree with outside magnetic probe measurements.
These
solutions also provide the ohmic heating term.
The bolometer
array signals deliver the radiation loss term after the charge
exchange neutral share is computed and subtracted.
Only the electron part of the ion cyclotron heating in thecentral region can be estimated experimentally. Therefore, thetotal deposition profile is calculated using the global wave code
LION [i].
This code calculates wave propagation for themagnetosonic wave in toroidal geometry, assuming the plasma
species to be Maxwellian.
In the case of on-axis heating ofhydrogen minority at its fundamental cyclotron resonance, thedistortion of the velocity distribution of the heated minority
ions becomes strong [2]. At low density, the heating of theminority species outside the central region is still high enoughto produce strong Doppler broadening of the resonance, whichbroadens the power deposition.
The power transfer to theelectrons and ions from that absorbed by the minority ions is
calculated with the Fokker-Planck code BAFIC [2]. Directelectron heating has been neglected. The power deposition on theelectrons is then checked against measurements, and generally
good agreement is found. At high temperatures, significant ion
heating occurs,due to second harmonic absorption by the
deuterons.
Also the power deposition from neutral beams cannot bemeasured directly and has been computed by Monte-Carlo or pencil
beam codes, with the total power, temperature and density
profiles, and Z g ^as measured input [3]. A major source of
uncertainty is the ion temperature profile which we took from theneutral particle analyser diagnostics [i] when available.
These profiles suffice to determine the remaining terms ineq. (2) if the equilibration (te ) is assumed to be classical.

327

IAEA-CN-47/A-VI-l-l

-1.5

•2

-4

.6

.8

1

r [m]

1.4

o

.2

.4

.6

.8

1

1.4

FIG. 1. Local power balance (integrals over magnetic flux surface volume inside r) for a JET pulse
during (a) ohmic heating and (b) RF heating. Ip = 2 MA, B, = 2.2 T, H minority (nH/nD -2%)
heating in (b).

Figure 1 shows the individual terms of eq. (2) for anohmically (la) and an RF heated (1b) JET plasma.
The averageminor radius r is used here as a flux surface label, i.e. it
specifies S. and V.. The integrated ohmic input power is denotedby P«, the RF input by P^p. the power exchanged between electrons
and ions by P e ^, the radiation loss by Pracj» and the temporalchange in thermal (electron) energy content by £ . The result of
the balance is F g = fg'S..
These cases have been chosen because F g is a significant
term (i.e. not only a difference of large numbers) over the wholecross-section.
In order to set off the difference between ohmic andadditionally heated plasmas, the figures 1 to 3 refer to two
different time slices of the same discharge with a total plasma
current I = 2 MA and a toroidal field (on axis) of 2.2 T.
Figure 2 presents a typical result obtained through theabove procedure for the ohmic case. Figure 2a shows the measured
plasma characteristics, i.e. density, temperature and q-profiles,
while the local mean thermal fluxes are given in Fig. 2b. The
same quantities for a neutral beam heated case are shown in
Fig. 1.
The error margin for these fluxes is assessed by varying theterms in eq. (2) within the experimental errors.

DUCHS et al.

328

2.5

2.0

rum
FIG. 2. (a) Plasma parameter profiles, and (b) measured surface averaged thermal fluxes ffor electrons
and ions, for the ohmically heated JET plasma (Fig. I (a)). Theoretical results for the dissipative trappedelectron mode (DTEM), scaled down by a factor of 50, and for the ion temperature gradient (1TG)
instability, scaled down by a factor of 50, are also shown.

0

.2

.4

.6

.8

1

r[m]

1.4

0

.2

.4

.6

.8

1

FIG. 3. (a) Plasma parameter profiles for an RF heated JETplasma, and (b) surface averaged thermal
fluxes fas derived (Section 3) for electrons and ions. The uncertainty off is enhanced within the sawtooth
region (dashed curves). The electron heat fluxes predicted for the dissipative trapped electron mode
(DTEM) and the microtearing (MT) mode are also shown.

IAEA-CN-47/A-VI-l-l

329

-'.

JfA

1

330

DUCHSetal.

3.

FLUXES FROM EQUILIBRIUM/TRANSPORT CODES IN STEAD? STATE

Our second method to obtain measured fluxes uses a fullequilibrium/transport code (in i|>(R,Z),t) containing more detailed
physics. For example, this code has been restructured to allow
fully consistent free-boundary equilibrium transport calculations.
The mathematical methods have been published recently [5]. The
code interacts with the external coils, which is especially
important for magnetic limiter cases in JET. In the interpretative application it is, of course, used to reproduce themeasured magnetic (probe, loop) signals, thus checking whether
the equilibria are consistent with (transport) pressure andcurrent density profiles.
In addition, it contains detailed
modelling of impurity diffusion and radiation, neutral atom
distributions, the sawtooth region, and boundary conditions.
The present method of application is as follows:The
measuredbasic
plasma parameters(densities, temperatures,fields) are inserted as initial values.
The measured complex
quantities, such as radiation losses, neutron yields, charge
exchange neutrals, are examined immediately through the models
for their consistency with the basic parameters; this leads
either to corrections or to an assessment of error bars. In thespirit of "shooting methods" an ansatz for the desired fluxes isguessed and programmed; for example, the result of the previous(cruder) power balance method could be chosen, or any function of
the basic parameters and their derivatives.
The code is thenstepped forward in time, and, in stationary state, must reproduce
the input initial value profiles, provided that our ansatz forthe fluxes was suitable, otherwise it is to be modified. Thismethod lends itself readily to sensitivity studies by determining
those terms which predominantly cause deviation from steadystate.
The method can, of course, also be applied to non-stationary
conditions, except that then several sets of measurements at
consecutive time slices are required.
Figure 3 presents the fluxes for an RF-heated JET plasma
evaluated by this method.
We have analysed many other cases which are being collected,
with some attempts at "scaling" for a forthcoming JET report [6].
U.

COMPARISON WITH THEORY

Expressionsfortheanomaloustransport,
derivedanalytically for different instabilities, have been evaluated for
JET parameters and compared with the measured fluxes.
The
extensive review in Ref. [11] and more recent literature has beentaken into account. Only those modes which may be unstable in
JET conditions, and for which predicted loss rates are available,
will be discussed. Space does not permit these expressions to be
reproduced here, they may be found in the references quoted. The
notation here is the same as in the references.

IAEA-CN-47/A-VI-l-l

331a) Electron transport
A significant level of agreement has recently been achieved[7,8] between the global energy confinement times measured on
several tokamaks and prediction based on the dissipatlve trappedelectron mode (DTEM). However, when the predicted local thermal
diffusivity [ 8 ] , including Z e f f in the collision frequency v e i ,is compared with the experimental value in JET, the predictedvalue is typically an order of magnitude too high within theregion where the DTEM should be dominant (see Fig.2b). Becauseof the strong dependence of the diffusivity on temperature (x e ~T e 7 / z for the DTEM and x e ~ T e 3 / * f o r t n e collisionless trappedelectron mode) and on the characteristic lengths r n and r^., theglobal confinement time is sensitive to the temperature profileand cannot be adequately expressed in terms of the peak or mean
temperature.
It is questionable whether the temperature in JET is low
enough, even in the edge region, to satisfy the criteria for therippling instability [9]. In any case, the predicted diffusivity
turns out to be too low by about two orders of magnitude.
The microtearing mode (MT) [10] could develop under JET
conditions for mode numbers m such thatin MKS units and eV. The predicted electron thermal diffusivity,
X e - T 7 ^ 2 /n Z r.p2 B 2 , has a parametric dependence similar to the
DTEM, and is larger in magnitude. Close to the plasma edge in
JET its value is comparable to the experimental value, but as one
moves into the plasma it exceeds the experimental value by a
rapidly increasing factor, as illustrated in Fig. 3b.
Collisional drift wave transport [13] could occur in the JET
edge region; however, the magnitude of the flux comes out too
small by 1-2 orders of magnitude.
Stochastic transport due to electrostatic trapping in drift
waves has been proposed in Ref [it]. It could occur within theoperating range of JET. However, as most of the other models, it
reproduces neither the spatial dependence nor the scaling with
global parameters obtained in JET.
b) Ion transport
The neoclassical ion heat diffusivity [15] is a factor 1-8
below the experimental value in JET. For example, in Fig. 3bbest fit was obtained with once neoclassical.
The observed saturation in T E at higher densities has been
reproduced in several tokamaks by including in the transport code
the diffusivity derived for the ion temperature gradient

332

DUCHS et al.

NBI + RF
9MW

4MW

1.10B

2

3.25

3.5

FIG. 5. Comparison between measured and empirically modelled (evolving) Te profiles for combined
ion cyclotron heating (H minority) and neutral beam heating of a JET plasma (lp = 2 MA,
B, =2.2 T).

instability (ITG) [7,8].
However, the predicted thermal
diffusivity for this mode in JET is more than an order ofmagnitude higher than the experimental value (see Fig. 2b).
5.

EMPIRICAL FLUX DESCRIPTION

ctory
the development of empirical prescriptions [16] has been pursued
further.
A widely applicable prescription is based on the experimental finding that, largely independent of additional heating
and radiation losses, two parameters suffice to describe theelectron temperature profile T g (p) in all steady states (forQ b < 7) byTe(p)= Te(pb)

expla[i-(p/pb)2]JCOor an expression with similar profile characteristics.
The flux surface coordinate ("radius") p b lies towards theinside of the radiation dominated layer, for most JET discharges
close to the limiter radius. The parameter a can vary between
1.5 < a < 2.2
with a strong peak around a = 1.9, and no clear dependence onother plasma parameters.
All rigorously derived transport equations are based on theassumption that the parameters of the system tend towards
thermodynamic equilibrium, e.g. flat profiles for vanishingsources. It is possible to restructure the balance equations so

333

IAEA-CN-47/A-VI-l-l

Shot 10237<T e >

1.0t(s)

0.0-4
12

13

/7G. tf. Evolution of on-axis electron temperature Te(0) and volume averaged electron temperature<.Te). Comparison of measured values computed with empirical ('profile consistent') fluxes.

that the parameters, here T e , always approach prescribed equilibrium profiles, e.g. the ones given above in formula ( M ) .
Such apparent neglect of thermodynamic principles might
indicate that not all important parameters are computed selfconsistently, i.e. that we are dealing with an open thermodynamic
system.
A physically consistent model to solve this problem, basedon partial ergodisation o f the magnetic field, is being developedby P.H. Rebut et al. [12,17].
A restructured balance equation for the electron thermalenergy does, of course, allow for temporal deviations from thegiven equilibrium profiles. In comparison with the experiment it
must satisfy two constraints:a)

T h e e q u i l i b r i u m profiles for T e ( p ) , obtained—(2. n g T e ) = 0,0 = p f e (p) +fromi.e. (integrated once over p) from
+ P R F + P N B I - P r a d - p e l ) P dp (5)must coincide with formula (4), andb)f e should reproduce the observed rate of relaxation towards
the given equilibrium profile (or the maximum deviation from
it for given sources).

334

DUCHS et al.

It is not clear whether these requirements lead to a uniquesolution, nor is a sufficiently wide data base available yet forpoint b.
An application of this procedure to particle transport isgiven by the popular anomalous inward particle flux term [18,19].
Several practical approaches to solve the above problem forT e have been reported [20,21], and the connections to widely used
empirical formulae have been pointed out. Our method [21 ] has,in the meantime, been successfully checked against a number of
JET discharges including strong additional combined RF andneutral beam heating and X-point plasmas.
Figure 5 shows anexample of "profile consistency".
Figure 6 demonstrates good
agreement with the measured time evolution of the electron
temperature.
6.

CONCLUSIONS

Measured local thermal fluxes in JET plasmas are presented.
A vital ingredient is the power deposition for the various
heating schemes.
For the RF deposition the theoretical model
which has been developed shows good agreement with measurements.
The JET code now allows transport calculations with fully
consistent free boundary equilibria.
The application ofpredictive transport codes to obtain experimental local fluxes isoutlined.
Previous studies have mostly compared measured and predicted
global energy confinement times. The comparison of local energy
fluxes made here is a more exacting test of theory.
The lowcollisionality
regime accessible to JET also extends theparameter range. Our comparison shows that theoretically
predicted loss rates do not agree with the observed radial
variation in the fluxes.
Empirically the temporal evolution of discharges in JET canbe reproduced well by a prescription which enforces a certain
shape of the T g profile in ohmic and additionally heated steadystate plasmas.
ACKNOWLEDGEMENTS
We are indebted to W.G.F. Core, S. Corti, M. von Hellerman, F.
Romanelli and E. Springmann for their help and useful
discussions.

REFERENCES
[1]
[2]
[3]

APPERT, K., et al., Plasma Phys. Contrail. Fusion 28 (1986) 133.
SUCCI, S., et al., Comput. Phys. Commun. 40 (1986) 137.
WATKINS, M.L., et al., in Controlled Fusion and Plasma Heating (Proc. 13th Europ. Conf.
Schliersee, 1986), Part I, European Physical Society (1986) 156.
[4] CORTI, S., et al., ibid., p. 109.

IAEA-CN-47/A-VI-l-l
[5]
[6]
[7]
[8]
[9]
[10]
[11]
[12]
[13]
[14]
[15]
[16]
[17]
[18]
[19]
[20]
[21]

335

CENACCHI, G., TARONI, A., in Computational Physics (Proc. 8th Europ. Conf. Eibsee,
1986), European Physical Society (1986) 57.
DUCHS, D.F., et al., JET Report, to be published.
DOMINGUEZ, R.R., WALTZ, R.E., Rep. No. GA-A-18184, General Atomic, San Diego, CA(1986).
ROMANELLI, F., TANG, W.M., WHITE, B.B., Rep. PPPL-2310, Plasma Physics Lab.,
Princeton Univ., NJ (1986).
GARCIA, L., DIAMOND, P.H., CARRERAS, B.A., CALLEN, J.D., Phys. Fluids 28 (1985)
2147.
DRAKE, J.F., GLADD, N.T., LIU, C.S., CHANG, C.L., Phys. Rev. Lett. 44 (1980) 994.
LIEWER, P.C., Nucl. Fusion 25 (1985) 543.
REBUT, P.H., BRUSATI, M., Plasma Phys. Contrail. Fusion 28 (1986) 113.
TERRY, P.W., DIAMOND, P.H., Phys. Fluids 28 (1985) 1419.
HORTON, W., CHOI, D.-I., YUSHMANOV, P.N., PARAIL, V.V., to be published.
CHANG, C.S., HINTON, F.L., Phys. Fluids 25 (1982) 1493.
COPPI, B., Comments Plasma Phys. Controll. Fusion 5 (1980) 261.
REBUT, P.H., et al., Paper IAEA-CN-47/E-III-4, these Proceedings, Vol. 2.
COPPI, B., SHARKY, N., Nucl. Fusion 21 (1981) 1363.
BEHRINGER, K., et al., in Divertors and Impurity Control in Tokamaks (Proc. IAEA Tech.
Comm. Mtg Garching, 1981), Max-Planck-Institut fur Plasmaphysik, Garching (1981).
TANG, W.M., Rep. PPPL-2311, Plasma Physics Lab., Princeton Univ., NJ (1986).
TARONI, A., TIBONE, F., in Controlled Fusion and Plasma Heating (Proc. 13th Europ. Conf.
Schliersee, 1986), Part I, European Physical Society (1986) 160.

IAEA-CN-47/A-VI-1-2

MICROINSTABILITY BASED MODELSFOR CONFINEMENT PROPERTIESAND IGNITION CRITERIA IN TOKAMAKS
W.M. TANG, C M . BISHOP1, B. COPPP, S.M. KAYE,
F.W. PERKINS, M.H. REDI, G. REWOLDT
Princeton Plasma Physics Laboratory,
Princeton University,
Princeton, New Jersey,
United States of America

Abstract
MICROINSTABILITY BASED MODELS FOR CONFINEMENT PROPERTIES AND IGNITIONCRITERIA IN TOKAMAKS.
The results of theoretical studies are reported for: (1) the use of microinstability based thermaltransport models to interpret the anomalous confinement properties observed in key tokamak experiments such as TFTR, and (2) the likely consequences of the presence of such instabilities for future
ignition devices. Transport code simulations using profile consistent forms of anomalous thermal diffusivities due to drift type instabilities have yielded good agreement with the confinement times and temperatures observed in TFTR under a large variety of operating conditions including pellet fuelling inboth Ohmic and neutral-beam-heated discharges. With regard to achieving an optimal ignition margin,
the adverse temperature scaling of anomalous losses caused by drift modes leads to the conclusion that
it is best to operate at the maximum allowable density while holding the temperature close to the minimum value required for ignition.

ANALYSIS OF ANOMALOUS THERMAL CONFINEMENT IN TOKAMAKS
Identification of the primary physical processes responsible
for the anomalous thermal transport properties currently observedin tokamaks is vitally important not only for the effective
development of methods to improve performance in existing devices
but also for the proper planning of future ignition experiments.
In recent work [1-4] it has been demonstrated that if realistic
profile constraints are invoked, then most of the significant
confinement trends in lowjbeta tokamaks can be explained byelectrostatic(E * B o )transport
causedbydrift-type
microinstabilities. Anomalous thermal transport models based on
the associated physics have yielded predictions in very favorable
agreement with a substantial experimental data base with respect
to both the scaling and the magnitude of the energy confinement
time ( T E ) and the central electron temperature ( T e o ) .
1
2

Culham Laboratory, Abingdon, Oxfordshire, UK.
Massachusetts Institute of Technology, Cambridge, MA, USA.

337

338

TANG et al.

Inthepresent
studies,theprofile-consistentmicroinstability models derived in Ref. [2")have been used in the
BALDUR (time-dependent, 1D) transport code [5] for the purpose ofinterpreting experimental results from TFTR including ohmic
heating, neutral-beam-injection (NBI) heating, and pellet-fuellingin both ohmic and NBI-heated discharges.
The theoreticalexpressions for the anomalous thermal diffusivities (xe, Xi) arebased on the presence of trapped-electron drift modes and/or
toroidal ion temperature gradient (n^) modes in the bulk region of
optimal thermal insulation (roughly located between the q = 1 andq = 2 surfaces in the plasma).
They incorporate the familiar
mixing length estimate for the magnitude of the fluctuations andalso
includethe assumption
thata temperature"profile
consistency" constraint [6] is satisfied. The latter is required
here because the present state of development of nonlinear
toroidal microinstability theories cannot provide reliable scaling
with non-dimensional parameters such as q and because mechanisms
outside the realm of microinstabilities could be playing aprominent role outside the q = 2 region. If the parameter, n^ =dJlnTj/dilnn^, remains below a critical magnitude, (npc (roughly
about 2 ) , then the dominant instability is the trapped-electronmode, and the associated form of the electron diffusivity can be
expressed as
TP0

X e (cm /sec)

O.H(1+.25a)1oV' 8 F (p)0.170.2R 0.4 0.9
n eQ R
a
Z
BT
qa

(1>

with
/a
[1-exp(-o )]exp(2a r 2 /3a 2 ) Joorr/adxxh(x)[ne(r)/neo] [r2/aZ)\\ dxxh(x)Hwhere a = q a + 0.5, P(MW) is the total equilibrated ("steadystate") value of the input power (minus radiation losses), h(x) isthe energy deposition profile, n eo (10cm ->), R(m), a(m), B T (T),and a n is the usual parabolic density profile parameter.
Inarriving at Eq. (1), the intrinsic temperature dependence has been
eliminated in terms of a power dependence through an assumed
thermal steady state [2]. The influence of these modes on Xi
tends to be masked by the usual ion neoclassical transport [2].
However, if n^ > (ni) c , then the toroidal n^-modes become dominant[7], and the associated profile-consistent form of the anomalous
ion diffusivity becomes
n
i,
*i

,

K

08
0 2 H
(RB_q ) ' (a)T^a

For a sufficiently collisionless plasma, Xe a l s 0 s c a l e s a s EQ(2).
On the other hand, for values of \>*e (the usual banana

IAEA-CN-47/A-VI-1-2

339

regime colllsionality parameter) i 0.1 to 0.2, the electron loss
channel remains sensitive to collisions in the presence of n^modes.
This is due to the fact that even though it does notstrongly affect these instabilities, the non-adiabatic trappedelectron response does determine Xe' Hence, the associated formof x e tends to be very similar to Eq. (1).
Simulations of nearly two dozen TFTR discharges covering awide range of experimental conditions were carried out with the
BALDUR transport code using x e = *e E ° + x£ E . Xi = X i E 0 + Xi* (withX^1 = 0 if Hj i 1,5 at q = 1,5). Other important effects, such as
sawteeth behavior inside the q = 1 region, radiation processes,
etc. have also been included and are described in Ref.[5l. With
regard to particle transport, empirical models were constructed to
produce density profiles approximating the Thomson-scatter ingmeasured density data. In particular, D = D N E 0 + D A N 0 M and v =0 + vAN0M.
For example, in simulating the pellet-fuelledvcases, D AN0M (cm 2 /sec) = n x 10 16 /n e (r) with v A N O f f = 2 D A N 0 M r / a 2
before and V A N O M = 0 after the pellets are injected. The resultsfrom the predictive simulations of numerous TFTR large plasma
discharges are summarized on Fig. 1 and Table 1 . On Fig. 1 thecalculated energy confinement times are plotted as a function of
the familiar neoAlcatorparameter and compared with theexperimental results from a large number of gas-fuelled ohmicallyheated discharges (crosses) as well as from pellet-fuelled ohmic
cases (open diamonds).
For the gas-fuelled cases it is quite
clear that the anomalous enhancement of Xi associated with r^modes (solid circles) can account for the observed saturation of
T E at values below those predicted by calculations using only Xi E °(open circles). The result from the simulation of an ohmic PDX
plasma is also displayed to indicate consistent size-scaling at asmall value of the neoAlcator parameter.
With regard to thepellet-fuelled cases the simulation.results are again in very good
agreement with the experimental data. As shown on Fig. 1, thefirst pair of diamonds correspond to values of t E at two points intime after the injection of a single pellet, and the second pairof diamonds to the analogous situation for a three-pellet
experiment. The higher T E in each pair represents the confinement
at the time closer to the time of injection,. In these cases theglobal confinement times are lower than those for the gas-fuelledexperiments because of larger energy losses associated withconvection and radiation. However, the simulations also indicatethat the central confinement is significantly improved after
pellet injection because of the resultant peaking in the density
profile. This trend, which is in agreement with that found in the
TRANSP data analysis, is likely due to x e scaling as 1/ne(r) andto the fact that the r^-driven anomalous enhancement of x^ c a n beeliminated (e.g. in the 3-pellet case). Results for T E and T efrom simulations of a representative set of NBI-heated TFTR
discharges are given in Table 1 and show good agreement with theexperimental values. These include a pellet-injection NBI case

340

TANG et al.

^

<sj

(Sj

«

d o d o

M

Km »

O>

—* ^-

H

*«

«

o

^

—<

d d d d dod•* — <s
d d d

o
Tt

-*t
N

oo oo
(<i «

oo

oo

©

.1 >

00

.0 (7.0

ri ri rn

en

So

tN

m

d

oo

O

00

oo

r^

oo

r\)

M

Cj

00

l
JJ -a

1.1
t-

I
on

en

ir,

t

8

341

IAEA-CN-47/A-VI-1-2
0.7

1 ' '

0.6 -

o

T -^Neo Alcator

0.5 -

IP

_Q4
V>

M a8

8

t

. siu X e T E ,X i n e o +X i 1 ' i* Gas Puff Experiment

0.2-

.V

« Pellet Experiment

0.1 -/,pof,,

I

,

10

20

neqa

FIG. 1. Results of simulations of neutral beam heated TFTR shots.

(shot 14773) and a typical "supershot" discharge (shot 22014).
All of the results displayed in Fig. 1 and Table 1 were computed
using Eq. (1). With regard to the "supershot" case, preliminary
calculations using a more realistic version of Xe^r^ (which takes
into account only the power into the electron channel and theproper transition into the collisionless regime) has yielded close
agreement with the experimental values given in Table 1; i.e. T E= 0.14s and T g = 6.2 keV.
In addition, all of the computed
temperature
profiles correlated
well withthe data, • andcomparisons of the theoretical \e(r) given by Eq. (1) and thoseobtained from TRANSP data analysis studies of a number of ohmic
and NBI-heated TFTR discharges have also indicated reasonableagreement.
Beforeproceedingwiththefurtherdevelopment
andapplication of the profile-consistent models just described tothe important class of H-mode NBI-heated divertor plasmas, it isnecessary
to first establish
the proper behavior of the
microinstabilities invoked under such conditions.
In previous
studies involving comprehensive kinetic toroidal computations [8],it was demonstrated that the weaker pressure gradients in theinterior region of H-mode plasmas led to correspondingly weakerinstabilities.
Although this reduction in the local anomalous
thermal diffusivity in the inner zone is consistent with theobserved improvement in T E , it leaves open the question of thenature of the transport properties in the very steep gradientregion near the edge.
Specifically, in order to be consistent
with the thermal fluxes calculated in the interior, the local
thermal diffusivity in the edge zone must be greatly reduced.

342

TANG et al.

This issue has been addressed in the present work by interfacing
the kinetic stability code [8] with a model tokamak equilibriumwith a separatrix [9]. The equilibrium here takes into accountthe enhanced local shear effects near the separatrix and isbasically an extension of the familiar "s-a model" [10] whose
finite-beta properties are primarily governed by the magnitude ofthe local pressure gradient. Fully electromagnetic calculationsof the dominant toroidal microinstabilities were carried out using
as input parameters the edge data obtained from carefulmeasurements of H-mode and L-mode PDX tokamak discharges.
The
results generally indicate that the proximity of the separatrix tothe magnetic surface considered has a significant stabilizing
influence. This effect is considerably stronger for the H-mode
cases with their characteristically steeper edge profiles.
In
particular, the finite-beta-modified drift-type electrostatic
modes are found to be stable over a region extending several
centimeters inside the separatrix for the H-mode profiles but overa much smaller range for the L-mode.
These enhanced local
stability properties near the separatrix thus serve to support thenotion that thermal transport properties in H-mode plasmas arereconcilable with microinstability-based models.

CONSEQUENCES OF MICROINSTABILITIES FOR IGNITION IN TOKAMAKSThe favorable correlations between the confinement trends
observed under a large variety of tokamak operating conditions andthe results from the anomalous transport studies just described aswell as from previous work [1-4] serve to emphasize the importanceof addressing the consequences of drift-type instabilities for
future ignition devices.
Under steady-state conditions, thesimplest form of the basic homogeneous (zero-dimensional) thermalenergy balance equation is given by

p - p
+ p
+ p _ p
h
Le
Li
R
ct

(3)
V

where P n is the external heating power density, P L e and P L 1
represent the losses due to electron and and ion conductivity, PR
represents radiation losses, and P a is the a-particle heatingterm.
In earlier calculations [11,12] it was demonstrated thatthe familiar Murakami density limit represents a balance between
radiation losses (PR) and ohmic heating input power (P^)However, since the fusion power (Pa) will offset PR attemperatures characteristic of ignition conditions (T Q > 6 keV),
this limit should not be relevant. Instead, attention needs to befocused on the influence of enhanced forms of P^g and Pj^ causedby the presence of microinstabilities.
First consider thesituation where the dominant terms in Eq. (1) are P a , PR (frombremsstrahlung), and P L e from anomalous losses associated with thedissipative trapped-electron modes. Using the simple local form

IAEA-CN-47/A-VI-1-2

343

for>

X e given in Ref. [1] and applying the boundary conditions ofsawtooth flattening for q £ 1 and T e = 0 for q >_ 2, thecorresponding 1-D radial equation yields eigenvalue solutions
which can be expressed in the form of an ignition criterion,[no(i015Cnf3)BT(i0 T)]2[a(m)]2-5[R(m)]1-5 > U . e ^ / q J 1 * V

COSince P^e is independent of density in the dissipative trappedelectron regime, it can be concluded that for ignition experiments
falling in this regime, the density can be increased to the point
where the criterion given by Eq. (4) can be satisfied, provided
external power (Pn) is supplied to compensate for the heat
capacity of the increased density and to maintain T Q > 6 keV sothat P a > P R .
The power required to balance the thermal
conduction losses is estimated to be P cr it(MW) = 120 (3/qa) '(To/6 keV) -t> a ^ / t Z B ^ U R ) 0 - 5 ) .
These results clearly suggestthat the optimal ignition scenario would involve operating at(i) the lowest allowable temperature (consistent with ignition),(ii) the highest allowable density (consistent with MHD 8-limits
but not the Murakami limit), and (iii)the flattest density
profiles (because of the strong an-dependence in the local model
used for xe^*
However, with regard to (iii), it should bestrongly cautioned that if the density profiles are allowed to be
too flat, then n^ can easily exceed (ni)c«
The consequentenhancement of P L i due to n^-instabilities can then pose a very
restrictive
density
constraint
in the ignition temperature
regime.
In order to simply illustrate this point, it isconvenient to express Eq. (3) in a schematic form highlighting thedensity and temperature dependences in each term, i.e. (taking It= T e = T)P

h " YTE T

9/2

2f"•" t V

1/2

5/2^

" V

>+ n \ T

5/2

,-,

(5)

q/2
with. ^ T E "being the dissipative trapped-electron losses,n^^BN b e i n 8 losses due to brejivastrahlung and ion neoclassicalconduction, Y a T 5 / 2 s P a , and nY n T 5 / 2 being the r^-mode losses. In
the absence of the last ternf, Eq. (5) clearly indicates that,provided P h > Y T E T y / , there would indeed be no density limit for3
T 2. T M I , with T M I = (TTBN/Ya)
minimum ignition temperature.
However, if present, this density-dependent loss term would force
operating below an upper density bound at T < T M I and above alower density bound at T > T M I .
The latter situation simply
corresponds to requiring P Q > ni^/P. Estimates from Eq. (5)indicate that the external poweri needed to reach ignition is
proportional to T^i• Hence, in order to avoid such stringent
constraints, pellet-injection fuelling procedures could be adopted
to produce density profiles sufficiently peaked to suppress the
Hi-modes.
After exceeding the required ignition temperature, a344

TANG et al.

less discriminating fuelling system would be adequate since P a
could then offset any ensuing anomalous losses of this type. In
general, for T > T ra j, the power losses resulting from drift modes
increase at least as rapidly with T as the fusion power.
Consequently, the optimum path to ignition should exploit
favorable scalings with B T , density, size, etc., while holding T
close to T m j .
ACKNOWLEDGMENTS
We are indebted to P.C. Efthimion, G.L. Schmidt, D.R.
Mikkelsen, and the TFTR physicists for providing the data as wellas valuable assistance needed to carry out the BALDUR
simulations. Special thanks are extended to S. Hiroe, B. LeBlanc,
and B. Grek for providing the profile data from PDX used in the Hmode calculations.
This work was supported by United States Department of Energy
Contract No. DE-ACO2-76-CHO-3073 and for F. Perkins also by NASA
Contract No. NAGW-91 and by the National Bureau of Standards.

REFERENCES
[1]

[2]
[3]
[4]
[5]
[6]
[7]
[8]
[9]
[10]
[11]
[12]

PERKINS, F.W., in Heating in Toroidal Plasma (Proc. 4th Int. Symp. Rome, 1984), Vol. 2,
Int. School of Plasma Physics, Varenna (1984) 977.
PERKINS, F.W., SUN, Y.C., Rep. PPPL-2261, Plasma Physics Lab., Princeton Univ.,
Princeton, NJ (1985).
TANG, W.M., Nucl. Fusion 26 (1986) 1605.
DOMINGUEZ, R.R., WALTZ, R.E, Rep. GA-18184, GA Technologies, Inc., San Diego, CA(1986).
ROMANELLI, F., TANG, W.M., WHITE, R.B., Nucl. Fusion 26 (1986) 1515.
SINGER, C.E., et al., Rep. PPPL-2073, Plasma Physics Lab., Princeton Univ., Princeton, NJ(1986).
COPPI, B., Comments Plasma Phys. Controll. Fusion 5 (1980) 261.
TANG, W.M., REWOLDT, G., CHEN, Liu, Phys. Fluids 29 (1986) 3715.
TANG, W.M., et al., Nucl. Fusion 25 (1985) 151.
REWOLDT, G., et al., Phys. Fluids 25 (1982) 480.
BISHOP, P.M., Rep. CLM-P-764, UKAEA Culham Lab., Abingdon (1985).
CONNOR, J.W., et al., Phys. Rev. Lett. 40 (1978) 396.
PERKINS, F.W., HULSE, R.A., Phys. Fluids 28 (1985) 1837.
COPPI, B., TANG, W.M., Rep. PPPL-2343, Plasma Physics Lab., Princeton Univ.,
Princeton, NJ (1986).

IAEA-CN-47/A-VI-1-3

DRIFT WAVE MODELSOF TOKAMAK CONFINEMENT
R.E. WALTZ, R.R. DOMINGUEZ, S.K. WONG1,GA Technologies, Inc.,San Diego, California
P.H. DIAMOND, G.S. LEE2, T.S. HAHM, N. MATTOR
Institute for Fusion Studies,
University of Texas,
Austin, Texas
United States of America
Abstract
DRIFT WAVE MODELS OF TOKAMAK CONFINEMENT.
Drift wave models are used to explain the observed tokamak confinement scaling with density,
power, size and magnetic field. Some additional edge mechanism is required to obtain the observed
scaling with safety factor and to account for H-mode transitions.

1.

INTRODUCTION

Simple scaling models for Ex Bo transport from drift wave turbulence have had considerable recent success in explaining many universally observed features of tokamak confinement. This paper summarizes
those successes and discusses the remaining difficulties of such models.
The theory is broken into drift wave models for the bulk plasma which
control the scaling of confinement time with power (temperature), density, size, and magnetic field, and models for the edge plasma which
control the scaling of confinement time and temperature profiles withsafety factor, q, and possibly transitions to improved divertor (H-mode)
confinement.
We report first on the application of some simple mixing length
models in a standard tokamak transport code [1]. Then,as an example ofmuch recent work on the theory and simulation of drift wave turbulence
itself, we summarize an analytic re-examination of ion temperature gradient mode turbulence [2]. Finally, we discuss several models for theedge plasma, their difficulties and their promises.
1
2

Jaycor, San Diego, CA, USA.
Oak Ridge National Laboratory, Oak Ridge, TN, USA.

345

346

2.

WALTZ et al.

TRANSPORT CODE SIMULATIONSIn the bulk plasma, turbulent transport from two drift mode
branches is important: the dissipative trapped electron (DTE) modefor o>*/i>eff < 1 or the collisionless trapped electron (CTE) mode for^•/^efT > 1> and the ion temperature gradient (rji) mode. At the plasmaedge, the collisional circulating electron (CCE) mode for ue/u)te > 1 orpossibly the universal circulating electron (UCE) mode for ue/<jjte < 1
may be important. In our transport code simulations we have assumed
that these have 7/A^ diffusivities scaling respectively as
DTE = e^(u>./kl) • [(«./!**), 1]™
Di = (t*/*2) • [(2/r) • (LJR) • ( W ^ ) ] 1 / 2
DCE = {"./kl) • (u;,/u>te) • [(!/e/u;te, 1 ] ^

(la)
(lb)
(lc)

where 7 is the linear growth rate and kx = (1/Axk) is the inverse
mixing length. We assumed kx equal to an rms poloidal wavenumber
kg = 0.3p~ l . [The notation is standard: e = r/R, a>» = kg(cTe/eB)/Ln,
VfS =• ue/e, ve is the electron collision frequency, r = Te(Ti, Ln and
LT are density and temperature gradient lengths, u;te = ve/Rq,-with
ve = (2Te/m e ) 1/2 , p. = c,/9,i with c, = (Te/rm)1'2 and Q = eB/rmc]
To specify the model further, we took the turbulent heat diffusivities to be Xe = c • 5/2 • [bTE + DcB) (1 + ceifi • LJR • Ln/LT) and
Xi = c • 5/2 • [Difi + Cie(DTE + -DcJsOLwhere Ce< = 3 represents electrontransport driven by the ion modes and Cie = 1 represents ion transport
driven by electron modes. /< = 1(0) for r/i = Ln/Lri ~> (•C) ?7ith ^ 1>
the ion mode threshold. A small and generally unimportant electron
to ion exchange A = Cje • nT • C(DTE + DCE)/L^twas included, c isthe "fit" coefficient (e.g., c = 1 for Dili, Alcator-A and -C, JET, andc = 0.3 for ISX-A). While the model was designed to fit Dili, subsequent
simulations of ohinic Alcator density scans showed premature density
saturation due to the electron driven CCE-mode scaling. This suggeststhat we assume the electron branch of the scaling to be stabilized byelectron temperature gradients rje except on the extreme edge. Neoclassical ion heat loss, classical exchange, Spitzer's relation between current
and voltage, as well as sawteeth degradation inside q = 1 were included.
Edge losses from convection, and charge exchange were not explicitly
included, and, more importantly, radiation was also not included.

347

IAEA-CN-47/A-VI-1-3

OH, 1 = 473 kA .
B = 24 kG
/

50

/
—

C.

t

1

T

• w<
10
Sff-

-

-

Vi
4.0
1.8

3.0
1.4

2.0
0.8

1.5
0.6

1.0
0.5

1

1

1

n(x1013cirr3)
FIG. 1. Dill density scan with Ohmic heating, showing total energy confinement lime TE with i)i mode"on" and "off" and electron energy confinement time TEe in comparison with experiment;n = n(0) (1 ^ 2

In low density ohmically heated plasmas, the DTE mode provides
the confinement time scaling (rB oc a?/x) r ^ 7 1 « na 1 - 04 ^ 1 - 79 ? 1 - 165-°- 33 Z- 0 - 17 (L n /a) 0 - 83 , which is very close to the established neoAlcator scaling. At higher densities,many tokamaks (e.g., Dili, AlcatorC gas) have progressively flatter density profiles (Ln large) with r)i increasing past the threshold. This causes the ion channel to be degraded
well below the neoclassical confinement limit, as often observed (e.g.,
ISX-A, Alcator-A, Alcator-C pellet), and TE<|H"n to saturate. This is
illustrated in Fig. 1 for DHL Radiation effects are needed for final saturation of Tjtj as the Murakami ohmic density limit is approached.

348

WALTZ et al.

OH "OFF", 50%-50% WEDGE HEATINGn = 6x10 1 3 c m " 3 , Pg = 3 MW, B = 24 kG, a,, = 0.6,50

1.75-2.25;
2.75-3.25
2-2.5; 3-3.5

10

6.0
, A

q(a)-

100

4.19
A I

3.0
A •

2.45
.A

500
KkA)

FIG. 2. Dili confinement time TE versus current at fixed power; 0% with no tearing mode short circuit;
50% partial tearing mode short circuit; 100% complete short circuit q > 2 (see Ref. [3]); dashed lines
indicate data for short circuit only at q = 2 and 3.

Dili (and several other tokamaks) typically beam heat from the77; saturated ohmic regime. The power scaling from this mode whenacting alone is rEB ex p-°6n0-6I°-8a2-2(a/R)-lsq0-6.
With sufficient
power or steepening density profile, the electron confinement may saturate with nearly identical CTE mode scaling [3]. Accounting for the
beam penetration effect modifies the density scaling from n0<6 to nO-O,asobserved. Simulations at low density show an improvement in T& withbeam power as the (classical) hot ion channel opens up in accord withexperiment (e.g., ASDEX). Thus the scaling with density (n), power(P), and size (o or R) are consistent with experiment. The scaling withcurrent (/) and field (B) are consistent,apart from the g0<8 factor which
gives a favorable B rather than the observed favorable / scaling. Thisis the principal failure of the bulk drift wave models.

IAEA-CN-47/A-VI-1-3

349

Some edge mechanism must be found to degrade the edge confinement at high q, making the effective size smaller than a and cancelingthe q0-8 factor [3]. The CCE mode with x «q2R2B~2L-2nZT-^2
has these qualitative properties. However, when acting alone with thebulk drift waves, it produces only a modest current scaling (no morethan r oc I°A). Some additional high-g mechanism is needed. If we assume that low-m tearing modes short-circuit the heat flow near integer
g-surfaces (q = m to m + VB, m = 2,3,4..., and VB is the fractionalamount between adjacent rational surfaces that is short-circuited), thenthe observed scaling T OC IB0 can be obtained for VB = 0.50. The current scaling of these models is illustrated by Dili simulations shownin Fig. 2.
In addition to controlling the g-scaling, the edge mechanism mustcool the edge sufficiently to force current to the center, thereby producing sawteeth and maintaining q(0) ~ 1. [The drift wave models alone are not sufficient and the tearing mode short circuits areneeded to force sawteeth at high q.] We have shown [4] that g(0) = 1
and Spitzer's law are sufficient to bound the temperature profile toq(a) > T(0)/{T) > g(a) 2 / 3 and provide the so-called principle of profile consistency. In fact, both ohmic and beam heated discharges from
Dili and TFTR have T(0)/(T) ~ g(a) 2 / 3 , the minimum requirement
for sawteeth, and recent high-g Alcator discharges without sawteeth fall
well below the peakedness bounds.
3.

THEORY OF ION TEMPERATURE GRADIENTMODE TURBULENCE

Tji-mode turbulence or ion pressure gradient driven turbulence(IPGDT) is of interest, not only because of its crucial role in tokamak confinement scaling described above, but also as a paradigm forsaturated multifield turbulence. It has a long history [5], with contributions from many workers. To facilitate analytic progress, the more
tractable sheared slab model is considered here rather than the toroidal
model [6]. The latter includes ion curvature drifts (oc 1/R) in addition
to ion parallel motion. Special emphasis is placed on the validity of themixing length estimates and the relation of ion thermal transport toelectron heat and particle confinement.
IPGDT develops from unstable Tft-modes, which result from thecoupling of sound waves to radial ion temperature gradients. For

350

WALTZ et al.

Vi ^ Wicrit > 1> IPGDT can be described by coupled equations for vorticity (Vniji), parallel flow velocity (v||), and ion pressure (p). In a fieldwith shear length La = Rq/s (s = d£nq/d£nr) and with L^/LaLj- <C 1,
the characteristic growth rates of the longwave hydrodynamic modes
are 7fc ~ [(1 + 7?«)/T]U>» • Ln/La. In general, the characteristic lengthof the modes is Ax*. ~ [(•yk/v*)/(Ln/La)]1/2pa.
Shorter wavelength
modes develop frequencies u>r ~ 7 as k$pa —» 1. Ultimately, ion Landau damping, represented in the fluid model as a parallel ion viscosity//|| ~ (<^/T)/|7*|, truncates the spectrum at kgpa ~ 1.
A natural description of the nonlinear fluctuation dynamics of
IPGDT utilizes coupled equations for the energies Ew ~ (<£2 -J- (V<£)2),
EJC ~ (^M)> Ep ~ (p2), thus ensuring consistency with the conservation laws and constraints of the system. The energy evolution is determined by physical processes characterized by three time-scales, whicharercfc = (Z?k/Ax2)"1, the nonlinear interaction time; Teq.jt = {[{1+Vi)fr]1/27fc}~1, the sound wave energy equipartition time; and Tdk= [nwkgAxl/L*]'1, the dissipation time-scale. For IPGDT, the timescales follow the ordering req.fc < rck < T^. Hence, exploiting thefact that all nonlinearities are convective, the energy equations may be
added so that the total energy £ is determined by™

(2)Here T1.2 and S\ 2 refer schematically to the Ex Bo nonlinearity triplet
and the pressure gradient source terms, respectively. Thus, saturated
IPGDT results from the nonlinear coupling of robust, hydrodynamic(7 > <jjr) long wavelength modes with marginal or damped short wavelength modes with k$ps ~ 1.
Standard methods of renormalized turbulence theory have beenused to solve Eq. (2). A "Reynolds number" Re =|| Ti.2 || / || M||Vjj ||~ Tdk/Tck — (1 + Vi) c a n ^>e defined. As Re > 1, the turbulence saturates by coupling to smaller scales,with typical scales fc^ ~ (Re)1/2fc0,
where kopa ~ 0.4, the rms wave number with weak parametric dependence. The predicted ion thermal diffusivity Xi = \C(R&)\2xYL>
where x ^ 1 = ifcoC^^fco 1S *^e ^ ^ ^ " ^ g t h prediction, and C(Re)= (TT/2) £n(Re). Note that Xi depends weakly on the dissipation /xy.
This is in contrast to mixing-length estimates, which effectively balance
2i.2[£] ~ ^1.2' However, dependence on ^i|| is to be expected sinceIAEA-CN-47/A-VI-1-3

351

even if "inertial range" properties are insensitive to /i||, macroscopic observables are determined by spectrum sums whose limits of integration
depend on dissipation. Indeed, C(Re) —» oo as ^|| —> 0!
Several transport coefficients of practical interest have been derived using the theory of IPGDT. The ion thermal diffusivity is givenby
Xi = 0.4|C(Re)|2 (j>2.c./L.) [(1 + m)/rf(3)

Using the dissipative trapped electron response (for i/*e < 1), the accompanying electron thermal Xe diffusivity and particle diffusion Dn are
obtained from quasilinear theory [xe, Dn oc |e<^/T|2],with n/n ~~ |C(Re)| 2 (7fc/w.) • (Axk/Ln):

Xe = 3.2 v ^ ( R e ) | 4 {CIPI/V&L]) [(1 + m)/r]Z(4a)

Dn = v^|C(Re)|4 (gAlualfi(4b)[(1 + m)/r]a • (1 + 3ifc/2)

For collisional electrons, the IPGDT-driven particle flux is given byr n =n 0 |C(Re)| 4 [x. + (1 + a T ) 2 ] [1 - (W»J.crlt)]
x (me/mi) (0.51/fc) (y/Ln) p\ [(1 + ^)/r] 2(5)where Xe = 1-61, « T = 0.71, and T]ecrit = 1-77. For T)e > f?ecrit» an
inward flow results. A normally outward heat conduction flow withsimilar scaling is induced by IPGDT at the edge.
There are several correlations with experiments. IPGDT drives a
toroidal flow viscosity fx± ~ Xt>which may explain the Dili observationthat toroidalflowdamping times r^ are comparable to Tg in co-injection
heated plasmas. In addition, impurity gradients can have a significant
effect on IPGDT. In particular, for an impurity species with density noj
and scale length L n j, IPGDT transport is modified according to the substitution (1+17,-) -»' A(l+7 7i ), where A = [l + Z(noI/noi){L^/L^)]-1.
Thus, impurity distributions peaked at the edge enhance IPGDT, while
distributions peaked on axis quench it. In particular, the inward par-352

WALTZ et al.

tide flux at the edge (u»e > 1) is strongly enhanced,r n —» A 3 r n ,foredge-peaked impurity profiles. This enhancement may underlie thedensity profile steepening and improved particle confinement following
impurity puffing observed during the Z-mode of the ISX-B tokamak.
From the previous intermediate formulas, alternative scalings canbe obtained in other physical limits. For example, in the extremely
flat density profile limit,L2n/LaLT » 1. j k - ^*[{Ln/L,){l + Vi)M1/2,
giving the mixing length estimate Xi ~ [(1 + ^i)/T]/:>4c»/in>aPart fromlogarithmic factors. In the extreme toroidal limit, jy. ~ u3m\2(Ln/R)(1 + m)/r]1/2, with Xi ~ [2 i,L,/R) (1 + Vi)H • /^c,/L n ,using the shear
mixing length [6]. [The latter formula differs from the toroidal scaling
Eq. (lb) [7,8] where a shearless mixing length Ax = 3p, was used.]
These formulas differ most significantly from Eq. (3) in their dependenceon La. However, they are difficult to distinguish experimentally becauseq{r)/s(r) is a weak function of q(a) in the bulk plasma where rji mode
turbulence is dominant.

4.

THE EDGE PLASMAOur understanding of the edge plasma and the mechanism forcontrolling g-dependence remains unsatisfactory. Other models havebeen considered in place of the CCE-mode scaling with only partial
success. For example, limiting the edge pressure gradient to the critical
ideal value (P' < P^t <* B2q~2) [9] is a high-g mechanism with similar
behavior when acting in concert with the low-m tearing mode shortcircuit model. Unfortunately we have not been able to theoreticallyjustify the latter. High-g modes (q — 4,5,6...) in particular are difficult
to destabilize even with thermal effects from radiative cooling at theedge.
Other models for dissipative fluid turbulence (DFT) at the edge
have been only partially explored. However, they have some noteworthy
common features which may be relevant to understanding the H-mode.
For example, a heuristic derivation for magnetic transport from microtearing modes has, in the collisional limit, Xe — ^avtPel^T' This is
identical (apart from the replacement of Ln by LT) to the CCE-mode
scaling Eq. (lc). Perhaps more promising is a recent result for B x Bo
transport from resistivity gradient driven turbulence (rippling modes)[10]:

IAEA-CN-47/A-VI-1-3

o c i - V i ? 2 (£?|,/5) 4 / 3 T-^n^L^a2'3

353

(6)

where E\\ = V\OOP/2TTR — T}\\J\\. (This scaling is at fixed mode number( m ) ^ suggested by numerical calculations. It also includes impurity
effects.) These diffusivities from DFT all have favorable T-scaling (incontrast to transport from bulk drift waves) and x « L ] oc s~2q2. We
speculate that both features may be used to explain the improved edge
confinement layer just inside the separatrix of the ASDEX H-mode.
Diverted plasmas naturally have a hotter edge on closed field lines exploiting the favorable T-dependence. Perhaps more importantly, it canbe shown [11] that while q diverges logarithmically near a separatrix,s~2q2 vanishes algebraically as (1 — r 2 /r^p) 2 , allowing a steep gradientin T as well as n to be supported as r —> r^p. Also, the steepening
density gradient may allow ideal ballooning modes to be diamagnetically stabilized. Indeed, using a> = a>»;/2 ± (w2^ — 47^ IHD ) 1 / 2 /2, modes
with kjp2 > (2Ll/LpR) [1 - (Lp/P)((3/Lp)crit], corresponding to thefull range of "high-n" modes for ASDEX parameters, are stable.

ACKNOWLEDGMENT
This is a report of work sponsored by the U.S. Department ofEnergy under Contract Nos. DE-AC03-84ER53158 and DE-FG0580ET53088.

REFERENCES
[1]
[2]
[3]

DOMINGUEZ, R.R., WALTZ, R.E., Nucl. Fusion 27 (1987) 65.
LEE, G.S., DIAMOND, P.H., Phys. Fluids 29 (1986) 3291.
PERKINS, F.W., in Heating in Toroidal Plasmas (Proc. 4th Int. Symp. Rome, 1984), Int. Schoolof Plasma Physics, Varenna (1984).
TANG, W.M., CHENG, C.Z., KROMMES, J.A., et al., in Plasma Physics and Controlled
Nuclear Fusion Research 1984 (Proc. 10th Int. Conf. London, 1984), Vol. 2, IAEA, Vienna(1985) 213.
[4] WALTZ, R.E., WONG, S.K., GREENE, J.M., DOMINGUEZ, R.R., Nucl. Fusion 26 (1986)
1729.
[5] POGUTSE, O.P., Sov. Phys. — JETP 25 (1967) 498.
COPPI, B., ROSENBLUTH, M.N., SAGDEEV, R.Z., Phys. Fluids 10 (1967) 582.
[6] HORTON, W., CHOI, Duk-In, TANG, T.M., Phys. Fluids 24 (1981) 1077.

354
[7]
[8]
[9]
[10]

WALTZ et al.

ANTONSEN, T., COPPI, B., ENGLADE, R., Nucl. Fusion 19 (1979) 641.
WALTZ, R.E., Phys. Fluids 29 (1986) 3684.
CONNOR, J.W., TAYLOR, J.B., TURNER, M.F., Nucl. Fusion 25 (1985) 65.
HAHM, T.S., TERRY, P.W., DIAMOND, P.H., Rep. IFSR-180, Inst. for Fusion Studies,
Texas Univ., Austin (1985), submitted to Phys. Fluids.
[11] HAHM, T.S., DIAMOND, P.H., Rep. IFSR-230, Inst. for Fusion Studies, Texas Univ., Austin(1986); Phys. Fluids (in press).

355
DISCUSSION
ON PAPERS IAEA-CN-47/A-VI-l-l TO A-VI-1-3
B.P. LEHNERT: Did you take into account interaction of plasma and neutral
gas in the edge region? The reason for my question is that even small amounts ofrecycling neutral gas could have an effect on the plasma profiles in a quasi-steadystate. As an example, an increasing average plasma density will lead to a decreasing
penetration length of the neutral gas, and this in turn will lead to a growing plasmapressure gradient in the edge region.
D.F. DUCHS: Neutral gas is self-consistently included in both the JET modeland the Princeton model. The model developed by GA Technologies explicitly
excludes neutral gas. We do not believe that it affects our results very much.
G. BECKER: One problem with electron heat diffusivities derived from drift
wave turbulence is that their scalings differ from the scalings of the coefficients determined from the experiments by local transport analysis or transport modelling. Discrepancies occur with the dependence on the ion mass number and the main magnetic
field, and in neutral beam heated plasmas also with the current or B . I therefore
conclude that the present drift wave models do not explain the experimental
behaviour observed.
D.F. DUCHS: As I have shown, our comparisons would lead to the same conclusion. However, if global profile consistency is considered, for example including
constraints, the dependent variables, such as Te in the expression for xe» c a n be
expressed in terms of BT. Such scaling shows good agreement with the experimental
results. In this sense, one should not rule out unstable drift waves as possible turbulence mechanisms.
K. ITOH: Did you observe that the discrepancy between the experimentallyderived heat flux and the theory, for example in the dissipative trapped electronmode, scaled with the plasma size? In other words, when comparing the data withthe experiment, did you find that an 'intermittence of the scale length' appeared inV?
eD.F. DUCHS: In our JET calculations we did not vary the plasma rise verymuch. However, we know from experience that, for example, 'Venetian blinds'
effectively reduce the plasma radius.
B. COPPI: The nature of the TJ; modes changes considerably over the rji range
1 - 4. In addition, the spatial mode localization is an important factor to be consideredin assessing the relevant transport. How is all this accounted for in the expression
for Xi presented by the group from GA Technologies?
D.F. DUCHS: The range of validity of the •ql theory presented is:•fi.crit. < *?; < W L j - Clearly, the theory fails for ^ ~ ij i c r i t , where detailed kinetic
analysis is needed, and for ^ > La/Lj., with a weaker dependence on rj;. The latter
regime may be relevant for some of the JET cases. For ij, cril < T\{ < L6/LT, Xj iswell approximated by the given expression. In the codes, a threshold turn-on function, varying rapidly with rj-^ is used as an approximation for reasons of practicality.

IAEA-CN-47/A-VI-2

MHD STABILITY AND TRANSPORT
OF BEAM HEATED ASDEX DISCHARGESIN THE VICINITY OF THE BETA LIMIT
O. GRUBER, G. BECKER, G. von GIERKE, K. GRASSIE, O. KLUBER,
M. KORNHERR, K. LACKNER, J.K. LEE 1 , H.P. ZEHRFELD, H.S. BOSCH,
H. BROCKEN,
A. EBERHAGEN,
G. FUSSMANN,
O. GEHRE,
J. GERNHARDT, E. GLOCK, G. HAAS, J. HOFMANN, A. IZVOZCHIKOV2,
G. JANESCHITZ,
F. KARGER,
M. KAUFMANN,
M. KEILHACKER3,
M. LENOCI, G. LISITANO, F. MAST, H.M. MAYER, K. McCORMICK,
D. MEISEL,
V. MERTENS,
E.R. MULLER 3 ,H. MURMANN3,
H. NIEDERMEYER,
W. POSCHENRIEDER,
H. RAPP,
H. RIEDLER,
H. ROHR, J. ROTH, F. RYTER4, F. SCHNEIDER, C. SETZENSACK,
G. SILLER, P. SMEULDERS3, F.X. SOLDNER, E. SPETH, K.-H. STEUER,
O. VOLLMER, F. WAGNER, D. ZASCHE

Euratom-IPP Association,
Max-Planck-Institut fur Plasmaphysik,
Garching, Federal Republic of Germany

Abstract
MHD STABILITY AND TRANSPORT OF BEAM HEATED ASDEX DISCHARGES IN THE
VICINITY OF THE BETA LIMIT.
Neutral beam heated H mode discharges in ASDEX exhibit a beta limit at /3C = 2.8 Ip/(aBt)[%, MA, m, T] in agreement with theoretical predictions. A 'hard /3 saturation' is always observed
nearly without prior confinement deterioration up to |8 max . Enhanced losses set in after the attainmentof finm, causing a subsequent decrease of the energy content by typically 30%. Analyses with the
TRANSP and BALDUR codes using measured plasma parameters reveal this transport enhancement tobe mainly due to electron heat conduction and convection losses. Impurity accumulation is correlatedwith the time evolution of j3. It can be ruled out as directly responsible energy loss channel, but may
cause enhanced current density flattening and a consequent reduction in the maximum stable /3 values.
Using the pressure and current density profiles from the transport analyses, the authors try to identify
the responsible instabilities theoretically. Low m ( < 6) pressure driven tearing modes are predicted by
the 3-D CART code and are seen in Mirnov coil signals, but show no clear experimental correlation
with the observed 0 limitation. The solution of the full ideal MHD ballooning equation predicts theseplasmas to be stable, leaving resistive ballooning modes in an intermediate range of wavenumbers(between 30 and 50) as a possible cause of the observed behaviour.

1

GA Technologies, San Diego, CA, USA.
Academy of Sciences, Leningrad, USSR.
3
JET Joint Undertaking, Culham, Oxfordshire, UK.
4
Centre de recherches en physique des plasmas, Ecole polytechnique f&Je'rale de Lausanne,
Lausanne, Switzerland.
2

357

358

GRUBER et al.

1. Introduction
In the divertor tokamak ASDEX neutral beam-heated H mode discharges allow,due totheir superior confinement^ test of the B limits over a large range of magnetic fields(1.2 < B{ < 2.7 T ) and plasma currents (0.2 < L < 0.48 MA) with the available
modest beam powers of up to 4.5 MW. A clear "operational" beta limit at
6 C = 2.8 lp/(a Bt) [%, MA, m, T] is observed IV (2a = 0.8 being the horizontal
plasma diameter). In the L mode even combined neutral beam and ion cyclotron
resonance heating with powers up to 5 MW allowed - possibly restricted by thelimited heating time IV - only values of B m a x <, 0.85Bc. H discharges with amaximum beta value B m a x below 0.75Gc display a nearly constant time behaviour ofbeta after reaching 3 m a x . Their normalized beta, B/Gc, is found to be proportional
to the normalized heating power P N = 3P/(bR0Bt) [MW, m, T ] , where P is the total
heating power (absorbed beam plus ohmic heating power), 2b is the vertical plasma
diameter (b/a=1) and R o is the toroidal major radius (=1.65 m). The resulting
energy replacement time T E * depends only on the plasma current. Various
combinations of P, L, and Bj produce no gradual reduction of confinement when B is
increased, but a "hard B saturation limit" when B c is approached. Such a hardsaturation is also seen in the time evolution of discharges with B m a x > 0.8Bc, which
also exhibit only a small confinement deterioration before B m a x (see Fig.1 and Ref.
/1,21). After attainment of B m a x the beta values of these discharges decay to about
0.7Bc and often reach a new stationary state at this level(see Section 2). Thissaturation near the B c limit without disruption is a peculiarity of the ASDEX
discharges. Disruptions, however, may occur both during the rise and the fall of beta
expecially at values of the cylindrical q a below three.
A Bc limit similar to the one observed in ASDEX and other experiments 13,4/ hasbeen predicted by MHD stability calculations for low m surface kinks /5/ andballooning modes /6/. For parameters and profiles of an actual ASDEX discharge
series with B m a x = 0.9 B c stability was found both with respect to ideal kink modes(using the ERATO code 121) and to ideal ballooning modes (by applying thecriterion in the approximation for circular, concentric magnetic surfaces:

2u
M

R
Bt2

q*

2
V

dp

v

dr

'M

359

IAEA-CN-47/A-VI-2/ * , 17005

j

1.0-

oh

8 468v/ +P/Pc.

16 981 ^ /

0.8-

*

D
y o
- 1 8 304
-18056^18041

a

+' /¥
>' /•

0.6-

1*-!— 16 996

0.4,'^£=0.151 [MA]0.2- //

0-Iro»-o* l = 270 kA ap// H-mode<

310 kA410 kA A

I

Ip = 310 kA •380 kA o

3|P-dE/dt)
bR0 B

FIG. 1. Maximum normalized /3 values (S/0C (from diamagnetic &pX) as a function of normalized heating power PN = 3 (P — dE/dt)/(b RQB,) for NI heated H mode discharges. Points connected by lines
indicate 0 decrease, after ^max, of the corresponding discharge.

where f(s) is a known function of the shear s =( r/q*) dq*/dr 171, p is the totalpressure, q*= 5B l r 2 /(2jtR 0 0 J r j{ r dr)[T, m, MA], and j t is the toroidal plasma
current density).
After these initial results, stability and transport studies of ASDEX divertor
discharges at the 3 limit have been intensified applying refined diagnostics (e.g.
detailed measurements of electron density and temperature profiles and of Mirnov
and soft XR signals) and more detailed theoretical analysis. Based on measured
profiles, plasma transport properties have been investigated by means of the PPPL
analysis code TRANSP /8/ and of a modified version of the predictive transport code
BALDUR121 (see Sections 2 and 3). The pressure and current profiles obtained
from transport computations were then used to calculate consistent MHD equilibria(see Section 4), which were in turn subjected to stability analysis by the 3-D CART
code /10/ (see Section 5) and by an exact solution of the ballooning mode equation(see Section 6).

GRUBER et al.

360
1.5-

6.0
/Prdl
[10^]
-4.0

PrH06$

1.1

1.3

1.4 Us]1.5

FIG. 2. Time evolution of ft±, central non-inverted chord intensity \prd( of bolometer array and central radiation power density pr obtained by Abel inversion for a DP — D+ discharge (No. 17005) andtwo H° — D+ discharges (No. 18040, No. 18041) with the same discharge parameter.

2. Current density profile and B limit

The p decrease after reaching transiently p c happens on a time-scale of about 30 to
50 ms (see Fig.2). An explanation for this behaviour, advanced in /11/, consisted ina reduction of the ballooning mode limited beta due to broadening of the toroidal
current density profile with time. In fact, integration of Eq. (1) over the wholecross-section yields a maximum attainable value of I3qa7e increasing with q a */q 0/12/ or the internal inductance Ij /11/ but depending only weakly on the finer
details of the current density distribution (here e = a/R 0 , q a *= q*(a), q o » q(r= 0)).
lndeed,during the high-beta phase, an increase of q 0 is indicated by the

IAEA-CN-47/A-VI-2

361

disappearance of sawteeth, and an Ij decrease can be deduced from magnetic
measurements, although the characteristic time for current diffusion due toneoclassical resistivity significantly exceeds the time-scale of the R changes.

There are two possibilities for reducing this apparent discrepancy of time-scales.
First, the instability which causes the B limit might also give rise to an anomalously
fast current flattening. Measurements suggest an additional, more classicalexplanation. In H mode discharges with Improved energy and particle confinement a
central accumulation of metal impurities takes place / 1 / yielding strongly peakedradiation power densities p r . Spectroscopy and soft XR diagnostics confirm that theevolution of the central peak of the bolometric profiles given in Fig. 2 is mainly due
to increasing line radiation of highly ionized iron and does not result from densitychanges. In 8 limit shots using H° into D + injection the radiated central power
density is smaller (see Figs. 2b and 2c), due to particle confinement deterioration
with respect to the D°-> D + discharges of Fig. 2a. Although the central p r values
become comparable to the local beam heating of the electrons, the globally radiatedpower rises at most by about 25 % and remains always less than one third of thetotal heating power. This implies that radiation itself is not responsible for the beta
limitation. But central Z ef f values above 4 are easily deduced if iron is assumed tocause the central radiation losses and if the temperature-dependent radiative power
loss function including CX recombination with beam neutrals is applied/13/. This
increase in Z e f f at r < 15 cm can explain an enhanced and accelerated broadening of
j t : TRANSP calculations in this case predict during the 6 decay of the D°-» D +discharge an increase of q 0 from 1.1 to 1.45 (rather than to only 1.2 as would result
for a fixed Z e f f (r, t) = 1.5) and an Ij decrease from 1.2 to 1.04.

This postulated influence of impurity accumulation and the consequent j t broadening
is supported further by the clear correlation of the central radiation intensity withthe time evolution of beta close to B m a x . For the same values of L, Bj, R o and n edischarges in which the central radiation increases earlier - this may happen due toan earlier L to H transition or a longer ELM(edge-localized mode)-free H-phase reach only smaller values of G/6C. Subsequently however betas decay to similarplateau values for both cases(see Figs. 2b and 2c). Overall, this time behaviour isconsistent with the conjecture that the plasma stays at a marginal stability limit
governed by current density profiles.

362

GRUBER et al.

3. Transport in H mode discharges close to the B limit

Detailed transport analyses have been carried out for all the discharges indicated in
Fig.1. Input data used were:the electron density profiles n e (r, t) measured by a 4-channel HCN
laser interferometer and a 16-spatial channel multi-pulse Thomson scatteringsystem which provides profiles every 17 ms,
electron temperature profiles T e (r, t) from 4-channel ECE diagnostic
and multi-pulse Thomson scattering,the ion temperature Tj measured by passive and active CX diagnostic,
Doppler broadening of CX recombination lines and neutron production rates,
the bolometrically measured profiles of the radiation losses and global
parameters like the loop voltage V L , l p , B p ± = B p t h + B p ± b e a m from diamagnetic
flux measurements and Bp(( + I/2 with Bp)| = 8 p t h + . 5 { B p l b e a m + B p || beam )deduced from the poloidal fields and fluxes. The calculated kinetic pressures include
the contributions due to the anisotropic fast beam ions ( p i | b e a m = 2 p 1 b e a m ) and arein good agreement with the magnetically measured ones.
Beam-heated H mode discharges staying below I3C show energy replacement times
comparable in magnitude to the ohmic ones (see Fig. 3) with (for q a > 3) a scalingT

E = 0-1 'pAj [s, MA, Aj = ion mass number] independent of the heating power (seeFig.1). For these high confinement times and consequently high plasma energies theconvective losses P c o n v /P =• 5/2rk(T e + Tj)/( G -EO) (with r the particle fluxand <E> the medium energy of the injected neutrals) and the ion conduction losses
become comparable to the electron conduction losses. So, over a large part of theplasma cross-section in D° -» D + discharges P conc j e amount to only between 0.2(Ip = 310 kA, q a *= 4) and 0.35 ( l p = 410 kA, q a * = 2.7) of the total heating powercompared to PQQ^ e = (0-5 — 0.65) • P in ohmic and L mode discharges /14/. Ion
heat conduction x\ is taken throughout as 3 times the neoclassical value as calculated
by C. Chang and F. Hinton for all discharge phases (OH, L, H), which gives the best fit
to the Tj measurements.
In discharges coming close to p c transport analyses confirm that additional energy
losses set in just after B m a x has been reached. This confinement degradation can beseen by the time evolution of the confinement times in the D° ->• D + discharge

IAEA-CN-47/A-VI-2

363

0.15

disruption

1.2

13 t[s] 1.4

t.u.- (b)y3.02.0at

1.0-

0-

0.5

r/Q

1

FIG. 3. (a) Time behaviour of energy replacement times TE* = E/(P — dEldt) with (TE*) and without(TE lh*) fast ions, and of thermal energy confinement time rE = E/(Pcond + Pconv) for a EP — D+
discharge (No. 17005).
(b) Evolution of electron thermal diffusivity \e (r, t) for the same discharge.

shown in Fig. 3a. For these discharges the ion heating by the beams, P bi , exceeds
the electron heating, P b 6 | (P b j < 3P be ) and the beam contributes about 25 % to thetotal 6 during the rise phase. The latter fact results in the difference of the x* E
values with and without beam contribution. The confinement time x E is defined
including only the conductive and convective losses and illustrates, when compared
to T * E , the influence of the radiation and CX losses.

These additional losses have to be attributed to an enhanced electron heat conduction
and partly to the convection associated with the density decrease. This is shown by
the time variation of the radial profiles of the electron heat diffusivity x e given in
Fig. 3b which are found to be nearly time-independent in the H mode during the B
rise phase, but are increased over the whole plasma during the G decrease. As n e andT e decrease during the latter phase, this translates only into an approximate
doubling of the electron heat conduction losses, which, at r = 0.75a increase from450 kW at 1.26 s to 900 kW at 1.38 s. An assumed additional increase of the ionheat conduction losses by 30 % brings the computed kinetic variation of B p ^ and
Bpll + lj/2 in agreement with magnetic measurements.

364

GRUBER et al.

A transport simulation of the H° -> D + discharge (see Fig.2b) by the predictive
BALDUR code reproduces the measured evolution of the n e , T e and Tj profiles and of
Rp up to the 3 limit if transport scalings derived from H-regime discharges well
below Bg/15/ are applied: Xe H ( r )= 1-3 a 2 / (R o rn l p ) [m 2 / s, m, MA], particle
diffusivity D H (r)=0.4.x e ^(r) and anomalous pinch velocity
v jn H (r)=-0.5iD H (r)r Te " 1 [m/s, m] (with rn = -n/(3n/ 3r) and r T e =-Tg/( 3Tg/ 3r)). After the attainment of G m a x also here the electron heatdiffusivity in the region r < 0.7a has to be enhanced by a factor of 4 to simulate themeasurements.

4. MHD equilibria for stability analyses
The above transport analyses produce profiles of the total plasma pressure p(r, t) =p t n + 1/2 ( p i | b e a m +PjL beam ) and - by solving the poloidal field diffusion equation- of the current density profile j t (r, t). These profiles are used to calculate MHD
equilibria whose stability is then analyzed. For given currents in the external
conductors a static plasma equilibrium configuration is uniquely determined, if
profiles of pressure p(V) and of the total toroidal current inside a given flux surfacel(V) are prescribed as functions of the volume V enclosed by it. The toroidal current
density can then be described in an unambiguous form as
2

1=2TH?{ 1 / Rt*•< 1 / R2 T 1 + B (1 - H R2< 1 / R2 r 1 ) ] —
1

P

J

dV

(2)

where B p is a local beta poloidal value given by Gp=dp/d4' • dV/dl, ¥ is the poloidal
magnetic flux and < > denotes the usual flux surface average. The corresponding
equilibria are then calculated by using the Garching free-boundary flow equilibrium
code NIVA. In order to make the radial data p(r) and l(r) suitable for a use in thetwo-dimensional equilibrum calculations we relate the normalized small radius r/ato the volume V enclosed by a magnetic surface by r/a = (V/V a ) 1 / 2 , where V a is thetotal plasma volume.

5. Low m mode stability
Our investigations of low m MHD activity are based on results from Mirnov probe
measurements /15/ and two-dimensional soft XR tomography /16/ and on their

365

IAEA-CN-47/A-VI-2#18041

1.2

1.3

1.4 Us]1.5

1.5' = 7.5 cm

r = 30cm

1.1

/Wv

1.2

1.3

1.4

1.2

1.3

14

tCsl

1.5

FIG. 4. Time behaviour of diamagnetic poloidal beta, line integrated soft XR radiation at various
tangential radii and Mirnov signals at torus outside.

comparison with stability calculations using the 3-D code CART developed at GA,
which is capable of treating the linear and nonlinear evolution of ideal and resistive
MHD instabilities at finite Gp^RQ/a.

The measurements show that the observed modes always have a toroidal mode
number n = 1. During the L phase of beam-heated ASDEX discharges a "continuous"mode develops, whose amplitude decreases after the L to H transition to a low level.
Both an m = 1 (detected by soft XR) and an m = 3 componenent (from Mirnov
signals) are present at the same frequency in the range of 10 to 20 kHz (see Fig. 4).
During the H phase ELM's occur IV and additionally MHD bursts are observed witha "fishbone-like" character. The fishbone amplitudes are much smaller than those of
the continuous mode and exhibit an m = 3 to 4 structure at lower I3p values

366

GRUBER et al.

(13/Rc<0.6) and an m =6 structure near the B limits (from Mirnov signals). Thispoloidal mode number is above the value of q at the boundary of an equivalentcircular discharge at finite P q a = q a * (1 + e 2 (1 + .5 (I3p + I/2) 2 )), so thatthe resonant surfaces m = q should lie close to, but inside the separatrix. The
frequency of the mode is observed to increase with heating power and Bp up to=25kHz. Both the continuous and the fishbone-like mode show an amplitude ratio up
to 20 between torus outside and inside and propagate with the central toroidal
rotation velocity. Sometimes, during the B decay, a strong m = 2 to 3 mode with
f=5-10 kHz develops (seen in soft XR and Mirnov coils, see Fig. 4), triggered by
an ELM, which frequently leads to a disruption especially at low q values /16/.
The low-m MHD activity observed seems to be mainly due to tearing modes. Moving a
conducting wall to the plasma surface in the CART stability analyses of thesedischarges does not significantly change the perturbed linearized n = 1 magnetic flux
function ¥ or the linear growth rates y. An indication for the driving mechanism of
this instability was obtained from the dependence of the computed 7-values on theresistivity r\. In simulations of our low B p discharges y scales like TI 3 / 5 , as is
expected for current-driven tearing modes. From simulation of discharges near the
B limit a scaling y°< TI 1 / 3 is derived which suggests that pressure-driven modes are
involved.
For a direct comparison of numerical results with the Mirnov signals we havecalculated the poloidal component of the perturbed magnetic field B e as a function of
the poloidal (see Fig. 5) and toroidal position. The nonlinear treatment, allowing forsuperposition of all toroidal Fourier components n, confirms the measured
dominance of the n = 1 modes. Within the poloidal region accessible to measurements
not only the m numbers but also the amplitudes are reproduced by the code, whereas
differences in the poloidal mode pattern exist. The code also predicts the shift in themode pattern to higher m numbers with increasing plasma pressure actually shownby the measurements.
However no clear correlation between the observed n = 1 mode activity and the B climit or the B decrease can be detected in the experiments. The strong m = 2 to 3
mode sometimes observed during the p-decay seems to be no necessary prerequisite
for it, but may be produced by the increase of q 0 during the high-B phase describedin Sect. 2, which shifts the q = 2 surface further to the plasma boundary. The

367

IAEA-CN-47/A-VI-2

1.2

1.4

1.6

1.8

2.0 R[m]FIG. 5. (a) Perturbed magnetic flux contours for unstable n = 1 mode of discharge with /3//3c = 0.6.
Full line shows magnetic separatrix and dotted line circle where Mirnov coils are seated,(b) Comparison of calculated magnetic fluctuation amplitude (full line inside, dotted line outside computational box of (a), with measurements (m — 4 mode: • , m = 3 mode: o ) . Hatched region shows
poloidal range covered by probes.

contribution of this mode - if present - to the 6 decrease is hard to state. Finally,
the energy losses due to the ELM's are already present in medium beta H mode
discharges showing no beta decrease and are therefore not responsible for the
additional losses at the beta limit.

6. Ballooning mode stability

A ballooning mode stability analysis of true ASDEX equilibria has been done bysolving the corresponding Sturm-Liouville problem for magnetic surfaces withvolume values V e (0 , V a ). Written in the form
d

l

0

(3)

the ballooning equation can be conceived as having an explicit dependence on the local
Bp, and an implicit one via the coefficients S, C-|, and C 2 depending on the total
equilibrium. In order to obtain information on the presumably attainable marginalvalue we have, based on a given equilibrium configuration, modified the explicitly
appearing local B p until a marginal solution of equation (3) with a Gp value B^ was
obtained. Fig. 6 shows the calculated beta ratio Bp/B M = p 7p ' M (with ' = d/d¥) asfunction of r/a for the already described D° -> D + and H° -> D + discharges with
B=6C at the time of 3 m a x .

368

GRUBER et al.

0.2

0.4 0.6 r / Q 1.0

0.2

0.4

0.6

r/tt

1.0

FIG. 6. Radial dependence of pressure gradient p' normalized to marginally stable one p'M, due toideal ballooning modes (a(S) criterion: 1, large aspect ratio expansion: 2, exact solution: 3), together
with pressure and q profiles used (p = pth + 1/2 (p1*™1 + p\eam) for a D° ~ D+ (No. 17005) anda H° - D+ (No. 18041) discharge at |3 ;.

Besides the general evaluation of equation (3),we have also performed a large
aspect ratio expansion of the coefficients S.C-) and C 2 . Although assuming circular
flux surfaces we have taken here into account the correct differential displacement of
the magnetic surfaces due to toroidal curvature r/R0, B_ = 2 \io( p- p)/B_ 2 andinductivity coefficient k
A'(r)=dA—
drr=- - (p
+1
/2)
p
R
*
o

(4)

with the average being defined by p= 0 J v pdV 7V = 2 o f pr 'dr 7r 2 . This treatment
is different from that in /7/ (see eq.(1)), where the quantities a and s are the only
free parameters of the problem, as well as from that in ref718/, where a special
functional form of the differential displacement was assumed. Except for theneighbourhood of the magnetic axis we find the results of our large aspect ratio
expansion in good agreement with those of the complete treatment; both these refined
approaches yield however p'/p'^ values about 20 - 30 % lower than given by thesimple the a = f(s) criterion when p' is in the vicinity of the marginal limit (seeFig.6). Broader current density profiles having higher q o values are more

IAEA-CN-47/A-VI-2

1.20

1.23

1.26

1.29

369

1.32

FIG. 7. Contours of constant pressure gradient p' normalized to marginally stable onep'M, due to ideal
ballooning modes (large aspect ratio expansion) as a Junction of time and normalized radius r/afor a
D° - D+ discharge (No. 17005).

susceptible to ballooning instabilities,due to their lower shear and vice versa. This
manifests itself for the given p profiles more by a p'/p'^ profile variation than by
an increase of its maximum value, which stays always below 0.9 for 0.8 < q 0 < 1.2.
(The differences in the p(r) profiles of Figs. 6a and 6b are due to the different
penetration of 40 keV D° and H° neutrals).
The time development of the ideal ballooning stability according to our large aspectratio expansion in the D° -> D + discharge is illustrated in Fig. 7. At the time of
maximum beta (about 1.26s),the discharge reaches a normalized pressure gradient
P'/P'M = 0.8 in the region 0.5 < r/a < 0.9 This is also the region where the enhanced
plasma losses start according to the TRANSP analysis as demonstrated in Fig. 3b.
During the 3 decrease between 1.28 to 1.38 s, the p'/p'^ = 0.8 region shrinks only
slightly, but the maximum value of p'/p' M stays below 0.9. Using the generalevaluation of equation (3) yields for 0.15 < r/a < 0.85 a region with p'/p'jyj values
between 0.7 and 0.85 during the beta decrease. The pressure gradient decreases infact, but so does p'^j due to the broadening of the current density profile (the latterbeing calculated using classical resistivity and the increase in the central Z e f f
described in Sect. 2). For discharges with flmax < 0.8 3 C the p'/p'^ values remain
significantly below unity with maximum values being smaller than thecorresponding global 6/6 c ratio.
In the ASDEX discharges close to the B limit the resistivity is very small but finite
so that, strictly speaking, the modes to be studied are resistive ballooning
instabilities. The influence of these modes and resistive interchange modes on the 8limit was investigated near the plasma axis (r <0.1 a) /19/. There the ratio

370

GRUBER et al.

S = tpj/ T ^ amounts to 10^ .where TR is the resistive time and x^ is the poloidal
AhVen time. For r = 0.1 a and mode numbers 20 < n < 40 that are not yet stabilized
by the finite ion Larmor radius effects the critical beta due to resistive ballooning
modes is 25 to 40 % below the ideal MHD limit. It is concluded that the discharges
under investigation are ideal ballooning stable but may be marginal with respect toresistive ballooning modes with high wave numbers. This would also explain thenear-invariance of the p /p ^ profiles during the B decrease.

7. Summary
The good confinement of H mode discharges has allowed the investigation of the
Bc=2.8lp/(aBt) limit derived on the basis of ideal MHD stability calculations. The
maximum betas are only reached transiently - normally without any disruption and beta decreases thereafter,saturating at a level of 0.7 to 0.8Bc. Nearly no
degradation of confinement is observed when approaching the limiting p: additionalenergy losses not following the usual H mode transport scaling set in near p cleading to a "hard" saturation. These losses persist, leading to a subsequent decreasein B,surmised to be a consequence of the reduction in the marginally stable p due to aflattening of the current density profile. Global MHD modes with m = 6 and ofpressure-driven tearing-mode type are observed, but no clear correlation between
their appearance and the B development has been identified. Though the B limit
discharges seem to be stable against ideal ballooning modes, resistive ones with highwave numbers m > 20 might be unstable.

REFERENCES[1]

KEILHACKER, M., et al., in Plasma Physics and Controlled Nuclear Fusion Research 1984
(Proc. 10th Int. Conf. London, 1984), Vol. 1, IAEA, Vienna (1985) 71; KEILHACKER, M.,
et al., Plasma Phys. Controll. Fusion 28 (1986) 29.
[2] BECKER, G., et al., Nucl. Fusion 25 (1985) 705.
[3] STRAIT, E.J., et al., in Controlled Fusion and Plasma Heating (Proc. 13th Europ. Conf.
Schliersee, 1986), Vol. 10C, Part I, European Physical Society (1986) 204.
[4] McGUIRE, K., et al., in Plasma Physics and Controlled Nuclear Fusion Research 1984 (Proc.
10th Int. Conf. London, 1984), Vol. 1, IAEA, Vienna (1985) 117.
[5] TROYON, F., et al., Plasma Phys. Controll. Fusion 26 (1984) 209.
[6] SYKES, A., et al., in Controlled Fusion and Plasma Physics (Proc. 11th Europ. Conf. Aachen,1983), Vol. 7D, Part II, European Physical Society (1984) 363.

IAEA-CN-47/A-V1-2

[7]
[8]
[9]
[10]
[11]
[12]
[13]
[14]
[15]
[16]
[17]
[18]
[19]

371

CONNOR, J.W., et al., Phys. Rev. Lett. 40 (1978) 396;
LORTZ, D., NUHRENBERG, J., Phys. Lett. A68 (1978) 49.
HAWRYLUK, R., in Physics of Plasmas Close to Thermonuclear Conditions (Proc. Course
Varenna, 1979), Vol. 1, Rep. EUR-FU-BRU/XII/476/80, CEC, Brussels (1980).
POST, D.E., SINGER, C.E., McKENNEY, A.M., BALDUR: A One-dimensional Plasma
Transport Code, Rep. 33, PPPL Transport Group, TFTR Physics Group, Princeton, NJ (1981).
LEE, J.K., Nucl. Fusion 26 (1986) 955.
VON GIERKE, G., et al., in Controlled Fusion and Plasma Heating (Proc. 13th Europ. Conf.
Schliersee, 1986), Vol. IOC, Part I, European Physical Society (1986) 33.
LACKNER, K., SYKES, A., private communication.
HULSE, R.A., et al., J. Phys., B (London). At. Mol. Phys. 13 (1980) 3895.
GRUBER, O., et al., in Controlled Fusion and Plasma Heating (Proc. 13th Europ. Conf.
Schliersee, 1986), Vol. IOC, Part I, European Physical Society (1986) 248.
BECKER, G., Nucl. Fusion 24 (1984) 1364.
KLUBER, O., et al., in Controlled Fusion and Plasma Heating (Proc. 13th Europ. Conf. Schliersee, 1986), Vol. IOC, Part I, European Physical Society (1986) 136.
KORNHERR, M., et al., ibid., p. 140.
CHOE, W.H., FREIDBERG, J.P., Phys. Fluids 29 (1986) 1766.
CORREA-RESTREPO, D., Plasma Phys. Controll. Fusion 27 (1985) 565.

DISCUSSION
M. OKABAYASHI: You reported that fishbone instabilities were excitedduring beta collapse. What is the actual power input if you include the fast ion losses
due to these fishbone instabilities?
O. GRUBER: We do not observe fast ion losses with the passive charge
exchange energy measurements. So I don't believe that the observed fishbone modes
contribute to the energy loss. Moreover, these modes are also present in discharges
with /3 < 0.7 |3C, showing a dependence of constant /3 on time and, therefore, noincreased losses.
K. McGUIRE: The edge relaxation phenomena can cause a large energy loss,
as was pointed out by Dr. Okabayashi1. What is the effect of the edge relaxation
phenomena and of core radiation on the beta saturation in ASDEX?
O. GRUBER: The total radiated power has no influence on the global balance,
despite the increased central radiation power density. Edge localized modes might bealmost absent before /3max, or are present after the H-transition during the whole beta
increase. This does not affect the beta limit observed in ASDEX discharges. We findedge localized modes also in the case of power beta values. Their contribution to thepower losses is included in our normal H-mode scaling.
D. OVERSKEI: Can you say what the thermal plasma contribution to thepressure gradient is? And what is the ratio of pt'herma, to Pcritical as a function of radius?
In addition, how does the fast ion contribution to p ~ {(5) vary in time and space
during this relaxation?
Paper IAEA-CN-47/A-V-2,

these Proceedings, Vol. 1.

372

GRUBER et al.

O. GRUBER: For the D° — D+ injection cases, ^beam is up to 25% of the total
beta before j8max is reached. Afterwards the beam contribution increases owing to
the decay of the thermal energy content until a new saturation level is reached. The
contribution of p^erma, to p^, is about 70% in the D° — D + cases, having a radial
profile similar to the total p'/pj^. For the H° — D + cases, fast particles aredeposited further inwards and the thermal contribution to p' near the centre is
reduced.

IAEA-CN-47/A-VI-3

BETA LIMIT DISRUPTIONS*
Mingsheng CHU, E.J. STRAIT, J.K. LEE, L. LAO,
R.D. STAMBAUGH, K.H. BURRELL, R.J. GROEBNER,
J.M. GREENE, F.J. HELTON, J.Y. HSU, J.S. KIM,
H. St. JOHN, R.T. SNIDER, T.S. TAYLORGA Technologies, Inc.,San Diego, California,
United States of America

Abstract
BETA LIMIT DISRUPTIONS.
Temporal behavior of disruptions in the maximal |3, discharges in Doublet III is analyzed
systematically by using ideal and resistive MHD models. These discharges disrupt at beta values
predicted by the Troyon-Sykes-Wesson limit. The basic disruption mechanism is identified as the overlapping of the central sawtooth mixing zone with the edge stochastic zone. This overlap is predicted
theoretically to be greater for more square current profiles and also to be enhanced at larger values ofbeta. This suggests that in order to find a disruption-free path to high beta, the current profile must be
made broader as beta increases. Eventually, even the broadened profile will become unstable to tearing
modes, at an average beta value lower than that of a profile optimized for stability against ballooning
modes alone, resulting in disruptions. This scenario is consistent with the experimental results.
Discharges which successfully approach the beta limit have a lower internal inductance (broader
profiles) than similar discharges at lower beta. For long periods preceding the disruptions, a significant
portion of the profile is calculated to be at or slightly beyond marginal stability for ballooning modes.
This long stable period argues against the ballooning modes as the cause of the disruption. As beta
increases, the window for avoiding the overlap of the sawtooth mixing zone and the edge stochastic zone
becomes narrower, until finally no way can be found to avoid the disruption; this defines the'operational' beta limit.

1.

INTRODUCTION

The maximum achievable fit value is the most important parameter determining the suitability of the tokamak as an energy producing
fusion device. A number of groups have published first stable regime (3tscaling laws of the form (3t (%) = C/(MA)/a(m) B (T). The constant* This is a report of work sponsored by the United States Department of Energy, underContract No. DE-AC03-84ER53158.

373

374

CHU et al.

C varies from 2.8 for ideal external kink modes with no wall stabilization [1] to 4.4 for ideal ballooning modes with optimized profiles [2].
The physical basis for the parametric dependence of this limit has been
elucidated by Wesson and Sykes [3] who have pointed out that if tearing mode stability is taken into account, C is reduced to 2.2 without anexternal conducting wall.
Experimentally, a value of C = 3.5 ± 0.5 was obtained in Doublet
III [4] covering a range of elongation 1.0 < « < 1.6 and triangularity—0.1 < 8 < 0.9. Concurrently, the parametric dependence of the experimentally achievable beta value has been analyzed by Bernard, Helton,
Moore and Todd [5]. They used the ideal MHD theory and assumed
that the plasma should be stable to both the localized ballooning modeand the external kink mode with an external conducting wall located at
1.5 plasma radius. A scaling law was deduced which explained the limitof the achievable beta in Doublet III. Experimentally, it is invariably
found in the Doublet III experiments that the plasma ends in disruption. Because of the narrow range of the theoretical predictions about
the experimental value for C, it is not possible to identify the modecausing the flt limit disruptions on the basis of the value of C alone.
We must therefore examine the details of the /3t limiting process. We
have carried out such a program of detailed examination of the observed
phenomena leading to the disruption, using both linear and nonlinear,and ideal and resistive MHD stability calculations for the particularequilibria obtained in the experiment.
A general pattern emerged in which the probable cause of the fit
limit is the unchecked growth of the m=2,n=l tearing mode shortly
after the last sawtooth. This 2/1 mode growth is caused by the finite (3 effect of shifting the current centroid and steepening the edge
current density gradient, and is triggered by the drop of the last sawtooth. Finite (3 also enhanced the field line stochasticity in the sawtoothmixing region and around the 2/1 island. In low qt (<3.1) discharges,
the ideal ballooning mode j3 limit of the configuration has not beenreached, whereas at high qt > 3.5, the ballooning mode stability limit
was violated over a large portion of the inner region of the plasma. The
destabilization of the ballooning modes in the central region is consistent with the experimental observation from the soft x-ray signal thatthe sawtooth mixing region was greatly extended in the last sawtooth
before disruption. It is proposed here that the overlapping of this central stochastic region with the edge stochastic region centered aroundthe q=2 surface is the direct cause of the (3 limit disruptions.

IAEA-CN-47/A-VI-3

2.

375

0 LIMIT DISRUPTIONS IN DOUBLET III

The /? limits in Doublet III were investigated by stepwise increasing the neutral beam heating power in consecutive discharges or in thesame discharge until the /3t limit disruption was produced [6]. No specific operational techniques were successful in trying to prevent these
disruptions except copious gas puffing which degraded TE and so limited the energetically achievable /3. A predisruption interval of 10 to
20 ms preceded the disruption. This interval is initiated by the drop of
the last sawtooth. During the interval, the global energy confinement
becomes poorer, the measured high frequency magnetic perturbation
signals on the Mirnov probe are replaced by the signal of a low frequency 2/1 mode, and intense interaction with the limiter ensues withlarge signals in the Ha limiter light. Prior to the predisruption interval,
the discharges have increasingly large sawteeth and constant heating
efficiency (AW/AP) as the NBI power and (3t rise. The current profile
steadily broadens, as evidenced by a decreasing £{.
2.1. Current Profile and Beta Optimization
In this section we try to elucidate the importance of plasma current
profile and finite f3 effects in determining the maximum achievable ft
value.
We explore first the (3 limit for a specific class of current profiles
for a large aspect ratio circular tokamak. The effect of /3 on the current
profile is temporarily ignored. This problem was first studied by Wesson
and Sykes [3]. They emphasized the optimization with respect to all
profiles with fixed q0 and qt. In this paper, we aim at understanding
the generic relationship among profiles which are not yet optimized withrespect to ballooning modes. A convenient parameterization is given by
3/2

A is a current peakedness parameter, which is higher for a peaked current
profile. At fixed qo and qt, TQ increases slightly with increasing A. P<jrepresents an amount of current pedestal, Ae is used to make the current
at the plasma edge vanish. Shown in the inset of Fig. 1 (a) are the typicalcurrent profile variations for qo = 1, qt = 3.05, Pj, — 0, and A = 2 and10. Higher A corresponds to a more square current profile. The (3t limit

CHU et al.

376

0.10

-

6.0

F/G. 7. (a) Maximum |3/e veri«5 ltfor a class of current profiles, low \ for flat current profile, high\for square current profile. Shown in the inset are the typical current profiles for X = 2 and X = 10.
(b) Location of the q = I surface and the 2/1 island size with q0 as a parameter for different current
peakedness. The enlargement of the 111 and 2/1 stochastic regions as a function of @/0c is shown
schematically in the inset.

for a particular current profile for infinite n ideal ballooning mode isgiven by [3]
-0.30

1\

, ,

IAEA-CN-47/A-VI-3

377

Shown in Fig. 1 (a) is the /3m versus li, with A as a p a r a m e t e r .
Ae has been fixed a t 30, with q0 = 1 and P j = 0. It is observed t h a t
a peaked current profile will admit of a larger /3m value, whereas aflatter current profile will accommodate a smaller (3m value. W i t h P&
nonzero, t h e ranges of /3m a n d li are somewhat reduced. If we could
achieve optimization only with respect to the ballooning m o d e without
an effect on tearing modes, we could aim at peaked current profiles withlarge li values.
Shown in Fig. 1 (b) is t h e location of the q=l surface a n d t h equasilinear m=2 island size for a qi — 3.05 plasma with P<j = 0. Before
the sawtooth drop, the plasma central q normally has 1 > q ~ 0.95,
and t h e sawtooth mixing region has a radius ~-\/2 larger t h a n t h e q=l
surface. After t h e sawtooth, the central q rises above 1 t o a value of~1.0 t o 1.2. Starting with a current profile with A = 4, which is moresquare t h a n the experimental profiles, and with central q ~ 0.95, after
the sawtooth there will b e a n overlapping of the 2 / 1 island a n d thesawtooth mixing region. For a less square current profile (A <^ 3), t h e
2 / 1 island and t h e sawtooth mixing region could be well separated. For
a very broad current profile (A < 2), t h e plasma becomes unstable t oexternal kink modes at low qi ( < 3 ) . Therefore, in order to optimize/3t for ideal ballooning modes while avoiding the overlapping of t h e 1/1
mixing region a n d t h e 2 / 1 island, we should aim at achieving t h e highest
li value which does not cause a mixing of t h e two zones. Experimental discharges in D i l i were almost always optimized t o obtain sawtoothoscillations because of their favorable property of expelling impurities
from t h e central region of t h e plasma. T h e resultant experimental profiles therefore h a d a relatively high li value corresponding t o A <* 3.
For a finite aspect ratio and finite (3t tokamak, increasing /3 t while
keeping t h e total current and qQ constant reduces £;. This reductionis brought a b o u t by t h e increased toroidal shift of t h e current centroid
and thus t h e reduction of the central current density. At t h e sametime, t h e edge current density gradient on t h e outboard side of t h etorus is enhanced a n d t h e 2 / 1 tearing mode destabilized. In contrast
t o the zero-beta cylindrical theory, t h e drop in li associated with finite
eflp is a destabilizing effect. Further, t h e finite e(3p effect enhances t h ecoupling to the side-band harmonics a n d increases t h e magnetic field line
stochasticity b o t h at t h e edge of the sawtooth mixing zone a n d aroundthe 2 / 1 islands. This could be visualized in Fig. 1 (b) by enlarging
b o t h t h e central mixing zone a n d t h e 2 / 1 stochastic zone. For a fixedcurrent profile this increasing stochasticity is expected t o close t h e stable

378

CHU et al.
1.2

o 0aB/K2.O• 0aB/l>3.O
1.0

--

0.8

-

O

°o

3

a
z

i-

z

0.6
2.0

w
I *>•
•
*

LIMITER DISCHARGES
40<a<42cm

i

1

3.0

4.0

5.0

q (SAFETY FACTOR)

FIG. 2. Experimentally obtained l: as a Junction of safety factor. Higher (-, values have lower
normalized /3.

operating window, as shown schematically in the inset of Fig. 1 (b). Thisis confirmed by numerical computations with increasingly large plasmapressure. It is observed that as the pressure increases, the 2/1 instability
generates a stochastic region with increasingly large size until a critical(3 value is reached, at which the whole plasma region becomes stochastic.
For this reason, in order to optimize /?t, we need to start with profiles
with even lower £{ as determined from the cylindrical theory, to avoid
the overlapping of 2/1 and 1/1 in the finite (3t plasma.
Experimentally, the statistical behavior of ^ as a function of qt
is given in Fig. 2 for a large number of discharges during the high /?
studies. It is seen that for discharges in which disruption is postponedby broadening the profile, a higher /3 is achieved. As beta increases, thewindow for disruption-free operation becomes narrower, until finally no
way can be found to avoid the disruption. This defines the "operational"
beta limit.
Another route which could be adopted is to eliminate the 1/1
region by raising the central q or stabilizing the 2/1 region by using acurrent pedestal and to aim at higher £{ with higher (3t values [7,8].

IAEA-CN-47/A-VI-3

379

2.2. Detailed Comparison Between Theory and Experiment
To further ascertain the fit limit disruption process, a detailed
comparison between theory and experiment is necessary. Stability
against ballooning modes [9] was investigated using the stability code
MBC [10]. The current and pressure profiles were obtained from the
MHD equilibrium analysis using EFIT [11]. An analysis using equilibria incorporating kinetic data does not result in substantially different
results. EFIT is an ideal MHD equilibrium code for determining arbitrary noncircular plasma cross-section and current profiles separating
fip and £; values. In discharges near the fit limit at high qi > 3.5, theinner portion (r/a < 0.7 to 0.8) was found to be unstable to high n localized ballooning modes. Crossing this theoretical threshold was foundto permit the achievement of stationary l{ and fit values before the disruption. The edge region of the plasma [where q(r) > 2] remains stableto ballooning modes throughout the discharge. For low qi < 3, marginal
stability is reached only just before the disruption, with steadily rising
fit and decreasing l{. Modes that were consistent with localization at aresonant surface (m/n=2/l, 4/2, 6/3) were observed near the fit limit.
This difference in the approach to the maximum fi is shown forthe high qt (~5) discharge 36288 in Fig. 3 (a) and for the low qt (~2.6)
discharge 43481 in Fig. 3 (b). It is seen that at low qt the trend of theti, fit and fi* development does not change until the disruption time,
whereas at high qt all these quantities show that a stationary value
was reached. The achievement of the stationary values of the plasma
equilibria quantities correlates with the ballooning stability criterion
development shown in Fig. 4 for high and low qt. At high qt the stability
criterion is violated over a large region for a time much longer than theenergy confinement time, whereas at low qt the stability criterion is only
marginally violated.
A last sawtooth drop initiated the predisruption interval. This last
sawtooth is characterized by an increased size of the sawtooth inversion
region as determined from soft x-ray emission signals. A particularly
large inversion radius enhancement is observed to occur at high qt shots
which violated the ballooning stability criterion. Thus an extended
central mixing zone is produced by this last sawtooth. This predisruption interval was shorter than the normal sawtooth. The central soft
x-ray signal also fails to rise to a comparable level. These findings support the hypothesis that in this central mixing zone, flux surfaces did
not completely reform.

CHU et al.

380

1.0

-

500.0

600.0

700.0

800.0

TIME (msec)

Fig. 3a. Behaviour of l{ (internal inductance), /3m (maximum theoretically stable 0J, 0, (toroidal fl),and /3* [= @(aB/I)] for a high q discharge, showing saturation behavior of £;. The vertical line
indicates the time at which the localized ballooning mode threshold is first crossed.

In the predisruption interval, high frequency and/or high mode
number MHD activity gives way to a low frequency mode with m = 2,
n = 1 in the magnetic probe data and m = 1 in the soft x-ray data.
The details of the experimental data traces are shown in Ref. [6]. Afterthe last sawtooth, MHD equilibrium code fits indicate that the toroidal
current density gradient and the pressure gradient have increased nearq—2 surface. Calculations using the ordinary A' analysis or using the reduced MHD stability code CART [12] predict that a tearing mode withdominant m=2,n=l character in ip should be unstable. Shown in Fig.
5 are a typical perturbed magnetic flux function and perturbed pressure patterns. The amplitude of the magnetic perturbation measured
at the Mirnov probes indicates that sizable (~20% of minor radius) 2/1
islands were excited. In resistive MHD simulations, toroidal and 0p coupling (e/?p ~ 0.4) with continued mode growth induces large field line
stochasticity outside of the q=2 surface which would result in a degradation in confinement over that region. Poorer edge confinement may
be the cause of the observed enhanced limiter interaction, as indicated381

IAEA-CN-47/A-VI-3
0.900

2.0

3

1.0

-

0.0
600.0

700.0

800.0

-

0.800

-

0.725

900.0

TIME (msec)

FIG. 3b. Same as Fig. 3a, showing the corresponding behavior for a low q discharge; no apparent
saturation of li: 13, or 0*.

by the increased level of the i^iim light. This is also correlated withthe fact that the frequency of magnetic perturbations is drastically reduced, indicating loss of charge by the plasma and the resultant poor
angular momentum confinement of the predisruption plasma. Material
recycled from the limiter may cause the edge and the 2/1 island centerto cool. Code simulations with enhanced resistivity in the island center
give enlarged island sizes and enhanced higher order island overlappingin the edge region [13]. It is observed that this enlarged edge stochastic
region can overlap the central sawtooth mixing region [14].
We conclude that there is agreement between experiment andtheory, confirming that an overlapping of the edge stochastic region
and the central sawtooth mixing region is the crucial event that leads
to disruption.

382

CHU et al.
1.50

(a)
750
725.^^775
1.00

0.50

t

_ a > 575^^"
650 "790
625M
600-

-

1

1

0.25

0.50

0.00
0.00

0.75

1
1.00

1.50
(b)

1.00

0.50

j

0.00
0.00

0.25

i
0.50

i
0.75

3
1.00

FIG. 4. Temporal development of P'IP'm as a function of the normalized flux function. P' is thepressure gradient and P^ is the maximum pressure gradient stable to the infinite n ballooning modes.
Times in milliseconds are indicated (a) for a high q, (~ 5) discharge, and (b) for a low qt (= 2.6)
discharge.

3.

CONCLUSION AND SUMMARY

We have performed systematically an analysis of the sequence of
events leading to the high /? disruptions in the neutral beam heated plasmas in Doublet III. It is found that the path through which the high f3
values were reached was consistent with avoiding disruption throughout
the experiment, except at the very end.
It is found that the disruption interrupts the process of pressure
and profile broadening through which the plasma reaches the ballooning
mode limit. At high q (^>3.5), the pressure gradient was able to develop

383

IAEA-CN-47/A-VI-3-0.044-0.046 t
0 0.2 0.4 0.6 0.8
—m- X

1.0

0 0.2 0.4 O.S 0.8 1.0—~-XFIG. 5. Perturbed magnetic flux function and pressure distribution from the resistive MHD code CART
computed for the low q discharge 43481 for equilibria during the predisruption period.

to the maximum allowed by the ideal ballooning modes over a large
portion of the plasma interior, whereas at low q (<^3.1), the ballooning
mode limit was barely reached. The general trend of increasing /3 in asawtoothing plasma is to destabilize the 2/1 region and to enlarge
the sawtooth mixing and 2/1 stochastic zones. The immediate causeof the disruption is triggered by a sawtooth instability which causes
the plasma current gradient to further steepen around the q=2 surface,
thus exciting a large 2/1 island. The accompanying loss of charge slows
down the island rotation and enhances the interaction of the island withthe limiter. Then, either a thermal island growth follows or the profilechange induces other edge resonance modes. The further growth of
the islands causes the enlarged edge stochastic region and the sawtoothmixing regions to overlap; this leads to disruption.
As beta increases, the profile window for disruption-free operation becomes narrower, until finally no way can be found to avoid thedisruption. This defines the experimental "operational" limit.

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CHU et al.

REFERENCES[1] TROYON, F., GRUBER, R., SAURENMANN, H., SEMENZATO, S., SUCCI,S., Plasma Phys. and C!ontr. Fusion 26 (1984) 209.
[2] SYKES, A., TURNER, M.F., PATEL, S., in Controlled Fusion and Plasma
Physics (11th Euro. Conf. Aachen, European Physical Society), Vol. 7D, Part
II (1983) 363.
[3] WESSON, J.A., SYKES, A., Nucl. Fusion 25 (1985) 85.
[4] STAMBAUGH, R.D., MOORE, R.W., BERNARD, L.C., KELLMAN, A.G.,
STRAIT, E.J., et aJ., in Plasma Physics and Controlled Nuclear Fusion Research(Proc. 10th Int. Conf. London, 1984) Vol. 1, IAEA, Vienna (1985) 217.
[5] BERNARD, L.C., HELTON, F.J., MOORE, R.W., TODD, T.N., Nucl. Fusion
23 (1983) 1475.
[6] STRAIT, E.J., CHU, M.S., JAHNS, G.L., KIM, J.S., KELLMAN, A.G., etai., in Controlled Fusion and Plasma Heating (Proc. 13th European Conf.,
Schliersee, 1986), Europhysics Conference Abstracts IOC. Part I, 204.
[7] HINTON, F.L., CHU, M.S., DOMINGUEZ, R.R., HARVEY, R.W., LAO, L.,et al., in Plasma Physics and Controlled Nuclear Fusion Research (Proc. 10th
Int. Conf. London, 1984) Vol. 2, IAEA, Vienna (1985) 3.
[8] GIERKE, G.V., BECKER, G., BOSCH, H.S., BROCKEN, H., BUCHL, K.,
et a]., in Controlled Fusion and Plasma Heating (Proc. 13th European Conf.,
Schliersee, 1986), Europhysics Conference Abstracts IOC, Part I, 33.
[9] CHU, M.S., LAO, L.L., MOORE, R.W., ST. JOHN, H., STAMBAUGH, R.D.,
STRAIT, E.J., TAYLOR, T.S., GA Technologies Rep. GA-A18546, submitted
to Nucl. Fusion.
[10] MOORE, R.W., MBC, A Ballooning Stability Code for Finite Toroidal Mode
Number, GA Technologies Rep. GA-A16234 (1981).
[11] LAO, L.L., ST. JOHN, H., STAMBAUGH, R.D., KELLMAN, A.G., PFEIFFER, W., Nucl. Fusion 25 (1985) 1611.
[12] LEE, J.K., Phys. Fluids 29 (1986) 1629; Nucl. Fusion 26 (1986) 955.
[13] KIM, J.S., CHU, M.S., GREENE, J.M., GA Technologies Rep. GA-A18586,,
submitted to Nucl. Fusion.
[14] CHU, M.S., GREENE, J.M., KIM, J.S., LAO, L., SNIDER, R.T., STAMBAUGH, R.D., STRAIT, E.J., TAYLOR, T.S., GA Technologies Rep. GAA18579, submitted to Nucl. Fusion (1986).

DISCUSSION
O. GRUBER: I would like to mention that we do not have sawteeth during the
H-mode phase of ASDEX, so overlapping of m = 1/n = 1 islands and m = 2/n = 1
islands is not possible in our beta limit shots. My question is as follows. In the paper
you presented at the 13th European Conference in 1986 it was stated that the ideal
ballooning mode limit is violated only near the plasma centre, whereas it is strongly
stable over the major part of the plasma cross-section. Now you have violation of theideal ballooning limit in the outer part of the plasma. How are the different pressure
profiles determined?

IAEA-CN-47/A-VI-3

385

Mingsheng CHU: If we use the kinetic hot ion pressure determined directlyfrom the hot ion deposition profile, then the central pressure gradient becomes verylarge and seriously violates the ballooning criterion at the centre. This is not very
reasonable. Therefore we use the magnetically determined pressure profile. The
ballooning criterion is only moderately violated. This is more satisfactory. A more
self-consistent hot ion pressure deposition profile is definitely called for.
M.A. DUBOIS: For your conclusions to be valid, you need to measure theq = 2 island size, for instance by a full Fourier analysis of the m = 2, n = 1 5B
components measured by the Mirnov coils. Are you sure that such a direct measurement of the q = 2 island width using Mirnov coils has been made on Doublet III?
Mingsheng CHU: In Doublet III, the magnitude of the magnetic perturbationis routinely measured by Mirnov probes located close to the plasma boundary. The
signal measured is compared with the computed magnetic perturbation amplitude atthat particular location in order to determine the m = 2, n = 1 island size. For
instance, shot 46167 is computed to have an island size greater than 0.31a at the timeof disruption. The q = 2 surface is at 0.83a, and the sawtooth mixing region withradius = 0.52a remains flat in terms of temperature throughout the predisruption
period, as determined from the soft X-ray emissions.

IAEA-CN-47/A-VI-4

SAWTEETH AND TEMPERATURE PROFILESIN TOKAMAKS
D.A. BOYD, R.E. DENTON, J.Q; DONG, J.F. DRAKE,
P.N. GUZDAR, A.B. HASSAM, R.G. KLEVA, Y.C. LEE,
C.S. LIU, F.J. STAUFFER
Laboratory for Plasma and Fusion Energy Studies,
University of Maryland,College Park, Maryland,
United States of America
Abstract
SAWTEETH AND TEMPERATURE PROFILES IN TOKAMAKS.
A detailed theoretical/experimental study of sawteeth, electron confinement, and temperature
profiles in tokamaks has been completed.

1.

INTRODUCTION

A theoretical/experimental study of sawteeth, electron confinement, andtemperature profiles in tokamaks has been completed. Temperature profiles in TFTR
discharges are distinctly trapezoidal (flat in the centre with a constant slope in theouter regions) at low values of the safety factor at the limiter, qL. For a given qL,
the profiles are insensitive to auxiliary injected power as well as other plasma
parameters. Such profiles result from sawtoothing at the plasma centre withanomalous cross-field thermal conduction in the outer plasma which depends strongly
on the electron temperature gradient, VT e . Such trapezoidal profiles cause skincurrents to form during the sawtooth cycle. The resulting current profiles produce
compound as well as normal sawteeth and enable us to simply explain the many
discrepancies between experimental observations of sawteeth and the standard
Kadomtsev model. A theoretical model of anomalous transport resulting from a shortwavelength VT e driven mode produces the observed scaling of central electron
temperature in TFTR and produces profiles in agreement with observations.

2.

ELECTRON TEMPERATURE PROFILES IN TFTR

An experimental study of electron temperature profiles in TFTR discharges has
been completed. Measured temperature profiles were normalized by the following
prescription: all minor radial co-ordinates were divided by the limiter radius and all
387

388
Q)

BOYD et a).

1.0
08

/T—
/T(0)
04

-1.0

00

LO

r

/o

FIG. 1. (a) T/T(0) versus r/a for six low qL Ohmic discharges. Here qL = 2.65, 2.65, 2.70, 2.79,
3.39, 3.41; \ndl = 4.8, 6.6, 4.0, 2.1, 2.0 x lO19 m'2; B = 4.66, 4.66, 4.80, 4.70, 5.81, 3.05 T;
R = 2.59, 2.59, 2.56, 2.57, 2.07, 2.79 m; a = 0.80, 0.80, 0.82, 0.82, 0.41, 0.58 m.
(b) T/T(0) versus r/a for four low qL, neutral beam heated discharges. Here qL = 2.65, 2.65, 2.67,
2.65; \ndt = 6.9, 8.7, 9.9, 9.3 x 1019 m~2; B = 4.66, 4.66, 4.83, 4.79 T; R = 2.59, 2.59, 2.49,
2.49 m; a = 0.8, 0.8, 0.83, 0.83 m; the injected power PB = 4.20, 4.26, 10.50, 9.56 MW.

temperatures by the temperature at r = 0, so that graphs of T/T(0) versus r/a could
be produced. Thus the shapes of profiles with very different temperatures and limiterradii could be compared. The profiles were measured with a fast scanning Michelsoninterferometer utilizing the second harmonic of the electron cyclotron emission
spectrum [1],Two basic profile shapes have been identified. The parameter that seems to havethe strongest influence in determining the shape is the limiter safety factor, qL.
The plot of T/T(0) versus r/a for six low qL, ohmically heated discharges isshown in Fig. l(a). These discharges range over a factor of 2.2 in density, 1.9 intoroidal field strength, 1.35 in major radius, and a factor of 2 in minor radius. The
cylindrical q L 's vary from 2.65 to 3.41. The average T/T(0) values are plotted withthe error bars showing the full range of T/T(0) at each r/a. The plot reveals a basically trapezoidal shape: flat within the sawtooth region and then a linear decrease intemperature until the limiter is approached. The solid line is a trapezoid fitted to thedata.
Figure l(b) shows a similar plot for low qL discharges that are heated byneutral beam injection. In this case a four shot average is shown. The density varies
by over a factor of 1.4 and the injected power varies from 4.2 to 10.5 MW. Although
the density increases by a factor up to 1.4 and the temperature up to 1.36 under injection, the profile shape barely changes (the trapezoid from Fig. l(a) is shown forcomparison).
High and mid qL discharges have a different shape. A plot for three high qL,
qL > 7, ohmically heated discharges is shown in Fig. 2(a). The shape shows strong
central peaking and from 0.1 < r/a < 0.6 the shape can be fitted with an exponen-

389

IAEA-CN-47/A-VI-4

a)

b)

i.o

i.o

0.8

0.8

T/T(0)

T/T(O)

0.6

0.6

0.4

0.4

0.2

0.2

-1.0
r/o

+1.0

-1.0

-1.0
r/a

FIG. 2. (a) T/T(O) versus r/a for three high qu Ohmic discharges. Here qL = 7.47, 7.87, 7.88;\ndt = 1.0, 1.3, 1.4 x 10'9m-2; B = 3.90, 4.72, 4.72 T; R = 2.55, 2.56, 2.56m; a = 0.82, 0.82,
0.82 m.
(b) T/T(0) versus r/a for six mid high qL, neutral beam healed discharges. Here qL = 7.91, 7.88,
4.70, 4.73, 4.72, 4.72; \ndt = 2.2, 3.5, 4.8, 4.9, 4.7, 4.9 x 1019 m'2; B = 4.72, 4.73, 4.8, 4.8,
4.7, 4.8 T; R = 2.56, 2.55, 2.48, 2.48, 2.48, 2.48 m; a = 0.82, 0.82, 0.83, 0.83, 0.83, 0.83 m;
PB = 4.23, 4.91, 11.0, 10.49, 11.31, 10.42 MW.

tial [2] with a characteristic normalized length of Ar/a = 0.36. Mid qL (~4.7)
discharges are often sawtoothing and have a flattened central region, but outside thisregion the shape is again exponential with the same characteristic length.
When subjected to neutral beam injection, high and mid qL discharges develop
broadened electron temperature profiles. The sawtoothing ceased during injection forthe mid qL discharges and so their data could be combined with the high qL data. A
six shot average has been plotted in Fig. 2(b). For these discharges the density varies
by a factor 2.2, the central electron temperature by 1.8, and the injected power from
4.2 to 11.3 MW. With high power injection the density increases by a factor of 2.3
and the temperature by 1.25. The profile can still be fitted with an exponential, but
the characteristic length is now Ar/a = 0.45.
Over a very broad range of operating conditions TFTR electron temperature
profile shapes have two distinct forms: trapezoidal at low limiter q and exponential
at mid and high qL. Strong neutral beam heating does not modify the profile shape.
Interestingly, at mid qL with high power neutral beam heating, the electron temperature drops when the beams turn on; the profile broadens and becomes trapezoidal.
When the electrons reheat and the temperature saturates the profile is once again of
the exponential form.
3.

SAWTEETH

The sawtooth cycle [3] results from the thermal runaway of the plasma column
which causes the central current density J to rise. In the Kadomtsev mode [4] the390

BOYD et al.

crash occurs when q = rB0/RBe < 1: the m = n = 1 tearing mode grows from theq = 1 radius until it encompasses the entire central plasma and expels the hot plasma
to the outside. Significant discrepancies between the Kadomtsev model andexperimental observations of sawteeth have been noted. The precursor oscillation
measured prior to the crash could be quite small, indicating that the m/n = 1
magnetic island was very small [5], Moreover, a successor oscillation is often
seen after the crash [5]. Very short crash times, apparently independent ofresistivity [6, 7], and compound sawteeth [8] also did not fit the Kadomtsev model.
We show that in discharges with good central confinement, the current profile
develops a distinct skin current which causes q to fall below one at a finite radius rs.
Compound and normal sawteeth are produced which are consistent with theexperimental observations.
We integrate the 2-D reduced resistive MHD equations for helical perturbations
with a self-consistent evolution equation for the electron temperature:dV2±<£/dt - nV\<f> = B-VJ
V2± \P = J - 2n/m;

dT/dt -

V±-K±V±Td^/dt-= B- V<t> + ijJ
2

KJ (B-V) T= fiijJ2

where d/dt = d/dt + v*- V, V = z X V 0 , B = z X V\p,<j> and i/< being the stream
and helical flux functions, 0 = a2Bo/4irnR2f with f an arbitrary normalization for
T, 7] = TJ/T3/2, and other notation standard. Times are normalized to the Alfven
time, TA = R/cA, and perpendicular scales to the minor radius, a. The perpendicular conduction K± = K I 0 (1 + a I VT I2), with the parameters a, K I 0 , rj, /i and
KI all constants. This form for K± allows us to reduce the central perpendicular
conduction ( K I 0 ) compared with that at the edge and produces the trapezoidal
temperature profiles measured at low edge q (Fig. l(a)). A theoretical model which
justifies such a form is presented in Section 3 of this paper.
We first demonstrate that the structure of the sawteeth in our simulations iscontrolled by the resistivity (temperature) profile and therefore the profile of K± [9].
In Fig. 3(a) we show the time dependence of T (r = 0) for a typical run with constant
K± (a = 0). Other parameters are rj = 2 x 10" 5 , 0 = 1, and <cg = 60. The sawteeth in this simulation are in excellent agreement with the Kadomtsev model (q first
falls below one at r = 0) but the discrepancies with experiment remain. In Fig. 3(b)there is a similar trace of T(t) for a typical simulation with K± reduced around thecentre of the plasma column (a = 1). Figure 4 shows profiles of T, J, and q at
three times during the sawtooth rise from this run. The rate of rise of T is now nearly
constant until the crash, while the central temperature profile remains rather flat, inagreement with the TFTR observations. More significantly, the current profile
develops a skin current causing q to first fall below one at a finite radius rs. For

IAEA-CN-47/A-VI-4

391

(a)
1.7
T(0)
1.5

0

4

8

12
t(IO3)

16

20

(b)
T(0)

1.7

1.5
1.6
T(0.5)
1.4
0

3

6

9
12
t(IO3)

15

18

,.7
T(0)
T(Q3)
8
12
t(IO3)

16

FIG. 3. Temperature versus time for (a) constant K±, (b) reduced K^ft) and (c) reduced KX(0) withlower resistivity.

such a q profile, all the previously mentioned experimental discrepancies with theKadomtsev model can be understood easily.
We first focus on the sawtooth crash. For a q profile of the form shown in
Fig. 4(c) with Aq0 = 0, the helical flux is reversed only in a narrow band Ar
around rs (Fig. 5(a)). Resistivity allows the reversed flux to reconnect as shown in
Fig. 5(b). At this point the tension in the helical field drives the plasma toward thecircular bubble configuration (Fig. 5(c)). Since no topological change in the magnetic
configuration is required for this transition to occur, the formation time of the bubble
is the helical Alfve'n time, T = rA/Aqs. Finally, on the same helical MHD time-scale
the bubble moves to the centre of the plasma column, causing the crash of the central
electron temperature observed experimentally. Detailed MHD simulations of thecrash have been completed for several values of Aqs over a range of resistivity from10" 4 to 6 x 10~7. The crash velocity Vc from these simulations (points) arecompared with the theoretical value, Vc = Aqsrs/7A (solid curve), in Fig. 6. The
error bars represent the scatter of points over the range of resistivities studied. This
figure confirms that the crash time is given by the helical Alfve'n time, independentof resistivity. Values of Aqs as small as 0.004 can cause the 50 /xs crash time
recently documented on TFTR [7].
When Aq0 7s 0 but Aq0 -4 Aqs, the formation of the bubble is essentially
unaffected by the central helical flux. The stronger magnetic field in the bubble
simply pushes aside the weak central flux (Fig. 5(d)), producing a large residual
magnetic island following the crash. Over a longer resistive time-scale the flux in392

BOYD et al.

00

IAEA-CN-47/A-VI-4

393

a)

b)

c)

FIG. 5. Schematic of evolution of helical flux for current profile with skin currents and q(0) — 1 [(a),(b), (c)J and q(0) £ 1 (d).

the residual island reconnects with that in the bubble and the column returns to asymmetric state. This residual magnetic island produces the successor oscillations
which are often seen experimentally.
When Aq0 > AqsAr/rs, the helical flux in the central plasma of Fig. 5(d)exceeds that in the bubble and the central temperature does not fall during the crash.
Such a 'subordinate' crash occurs during compound sawteeth. An example of thisphenomenon is shown in Fig. 3(c) from a simulation with rj = 10~5, a = 1,
fi = 2, and «i = 100. The third and fifth sawteeth are compound with subordinate
crashes occurring near t = 7500 and 15 000. Compound sawteeth occur at low edge
q when K±0 and 77 are small.
We emphasize that the magnetic island present prior to the crash can be verysmall as shown in Fig. 3(b). Since the area within the closed helical flux is conserved
during the formation of the bubble from Fig. 3(b) to (c), an expression relating the
bubble radius rb to the island width w of the precursor is easily computed:
w = Tb/2rs. In recent observations of sawtooth crashes on JET, the formation of abubble much like that shown in Fig. 5 was observed with no obvious precursor [6].

394

BOYD et a).
3.0 /
2.5
2.0

/

'

1.5
1.0
0.5
/

0

i

1 2

i

i

i

i

3 4
Aq(IO"2)

5

6

i

,

FIG. 6. Crash velocity versus Aqs for Aq0 = 0.

However, in their experiment rb ~ rs/3 and therefore w ~ rs/18 so that the precursor would be extremely small. It has been suggested that the pressure driven internalkink is responsible for the sawtooth crashes in the JET experiment [10]. The competition between reconnection and the pressure driving forces in causing the sawtoothcrash is currently under investigation.

4.

A THEORETICAL MODEL FOR ENERGY TRANSPORTThe formation of trapezoidal temperature profiles similar to those observed
on TFTR in our sawtooth simulations with a model cross-field conduction
K± oc I v T I2 provides strong evidence for the role of the electron temperature
gradient in causing anomalous transport. The electron temperature gradient is known
to be an important source of free energy for instabilities causing such transport. The
electron temperature gradient destabilizes the collisional trapped electron modes.
However, the transport coefficient derived for such modes decreases very rapidly
with increasing r contrary to experimentally inferred xe profiles. Recently we havefound [11] that the electron temperature gradient modes can also destabilize modes
with wavelengths between the electron and the ion gyroradii. A systematic study ofthe mode in a sheared slab magnetic field geometry together with full gyrokinetic andfinite beta effects yields modes with a threshold rje = d In Te/d In n = 1.0. Here,
Te is the electron temperature and n the density. The mode frequency ranges from0.05 to 0.4 times the electron diamagnetic frequency u ^ = kycTe/eB0Ln where Ln isthe density scale length, and ky is the azimuthal wavenumber.
Following a quasilinear theory, the anomalous electron energy coefficient isestimated to be

IAEA-CN-47/A-VI-4

395

5r

.4

.2

.1

.2

F/G. 7. 7 / ^ verms r ; /a: O analytic model; © numerical model; and ©. experimental points for
TFTR [12].

Xe = 0.1 C2sve7je(l + 1?e)/<o2KqR
where c is the velocity of light, ve is the electron thermal velocity, u^. the plasma
frequency, and s = d In q/d In r. One interesting feature that emerges from theabove expression is that the temperature gradient dependence is similar to that usedin the sawteeth studies.
We introduce this xe m t 0 t n e steady state equations for electron energy transport in an ohmically heated tokamak. We assume that the density gradient is given
by n = no exp ( —o;r2/a2); here, a. — 2. From our numerical studies it is found thatthe central electron temperature T(0) scales as
T(0) oc B 0 2 / 3 R- 1 / 3 a 2 VV / 3 (Z,'eff)
where 7(Zeff) = Zeff[0.29 + 0.46/(1.08 + Zeff)]. This is in reasonable agreementwith the TFTR scaling [12].
T(0) « B^- 6 R- ol a a7 q£ 2 Z e V
Another interesting consequence of this thermal conduction coefficient is the relationship between r,, the inversion radius for sawtoothing, and the limiter safety factor
as shown in Fig. 7. Curve 1 corresponds to a simple analytical model, curve 2 to thefull numerical model and the points correspond to data from TFTR [12]. Furthermore, we find that r,/a = 0 for qL = 5.6. Thus for qL > 5.6 the discharges should

BOYD et al.

396
1.0
0.8

/

0.6

\

/

T/T(O)
0.4

\

/

\*

0.2
•

/

\

/, . . .

0
I

.8

.6 .4

i ...
.2

0 .2
r/a

.4

A
.6 .8

I

FIG. 8. T/T(0) versus r/a. Solid line is experimental data from TFTR ECE diagnostic for qL = 3.3,
B = 3.04 T, R = 2.79 m, a = 0.55 m.

T/T(O)

-1.0

-0.5

0.0

0.5

1.0

FIG. 9. T/T(O) versus r/a for qL = 3.0, a = 2.5: © Ohmic profile; © profile with auxiliary heating
deposition profile given by ©; ® profile with auxiliary heating deposition profile given by © .

stop sawtoothing. Experimentally, for TFTR and Alcator it is found that the transition to non-sawtoothing discharges occurs around qL = 7.0. Figure 8 shows aplot of the normalized temperature T/T(0) against r/a for TFTR from the ECE
diagnostics. The dots correspond to the theoretical profile.
Next, we investigate the influence of auxiliary power on the temperature
profiles. The theoretical expression for xe has been incorporated into the sawtoothcode. The profiles halfway up the sawtooth rise are shown in Fig. 9. Curve 1

IAEA-CN-47/A-VI-4

397

corresponds to the Ohmic temperature profile. Curve 2 is the auxiliary heated plasma
temperature profile for the same qL as the Ohmic case and an auxiliary power
profile given by curve 2a. The amount of auxiliary power is assumed to be four times
the initial Ohmic power. Curve 3 is the auxiliary heated temperature profile for theauxiliary power deposition profile given by curve 3a. As can be seen, the only
significant changes in the profile occur inside the inversion radius. These changes
become even more significant as one goes to higher limiter safety factors. Thus for
low qL the present model shows that the electron temperature profiles are nearlyinvariant with different heat deposition profiles. However, as one goes to higher
values of the limiter safety factor, the profiles can be modified significantly.
In conclusion, we may summarize that we have derived a transport coefficientfor short wavelength temperature gradient driven instabilities which explain many
facets of ohmically and auxiliary heated discharges.

ACKNOWLEDGEMENTS
This work was supported in part by the US Department of Energy. It is apleasure for us to acknowledge the support of PPPL and, particularly, the TFTR
experimental group.

REFERENCES[1] STAUFFER, F.J., BOYD, D.A., CUTLER, R.C., MCCARTHY, M.P., Rev. Sci. Instrum. 56(1985) 925.
[2] FREDRICKSON, E., et al., in Controlled Fusion and Plasma Physics (Proc. 13th Europ. Conf.
Schliersee, 1986), Vol. 10, Part I, European Physical Society (1986) 41.
[3] VON GOELER, S., STODIEK, W., SAUTHOFF, N.R., Phys. Rev. Lett. 33 (1974) 1201.
[4] KADOMTSEV, B.B., Fiz. Plazmy 1 (1975) 710; Sov. J. Plasma Phys. 1 (1975) 389.
[5] DUBOIS, M.A., MARTY, D.A., POCHELON, A., Nucl. Fusion 20 (1980) 1355.
[6] EDWARDS, A.W., et al., Phys. Rev. Lett. 57 (1986) 210.
[7] McGUIRE, K., private communication.
[8] PFEIFFER, W., Nucl. Fusion 25 (1985) 673; CAMPBELL, D.J., et al., Nucl. Fusion 26 (1986)
1085; YAMADA, H., et al., Rep. PPPL-2213, Princeton Plasma Physics Laboratory (1986);
KIM, S.B., private communication.
[9] DENTON, R.E., DRAKE, J.F., KLEVA, R.G., BOYD, D.A., Phys. Rev. Lett. 56(1986) 2477.
[10] WESSON, J.A., Plasma Phys. Controll. Fusion 28 (1986) 243.
[11] LEE, Y.C., DONG, J.Q., GUZDAR, P.N., LIU, C.S., Rep. No. 86-023, Maryland Univ.,
Baltimore (1986).
[12] TAYLOR, G., et al., Nucl. Fusion 20 (1986) 339.

398

BOYD et al.

DISCUSSION
L.E. ZAKHAROV: Have you any comment to make regarding the obvious
contradiction between your theory of internal disruption and the TEXTOR q(0)measurements?
J.F. DRAKE: We are in the process of including toroidal effects in our code.
There are two important points about sawteeth which will need to be carefully
studied. The first point is the competition between the ideal and resistive modes. Bothof them will produce similar signatures as far as the experiments are concerned.
Which of them is used in the actual machine remains to be seen. The second pointis the TEXTOR results, which require an appropriate explanation.
O. GRUBER: In Ohmic discharges the time behaviour of the electron temperature rise when the sawtooth drops is linear and in beam heated L-mode discharges
it shows a saturation point. The latter behaviour is not present in your calculated Te
traces. Can you comment on this?
J.F. DRAKE: We do not always have a linear temperature rise. If the confinement is degraded in the centre of the plasma, the sawteeth distinctly rise. In our firstsimulations, all the sawteeth rounded out and virtually stopped rising during the risephase. At least for the Ohmic case we found we had to reduce the condition in thecentre in order to get the linear rise. And if your confinement is being degraded inbeam heating, you also get saturation of the sawteeth.
R.E. WALTZ: The advantages of the ije mode model for transport whendescribing profiles have been emphasized. I think, however, I should point out thedisadvantages of this model in explaining the scaling of the confinement time. Firstly,
the model has an Ohmic confinement time scaling roughly equal to TOH « na2Rq, in
contrast to experiments (and dissipative trapped electron drift mode scaling), where
it is TOH a naR2q. Secondly, and more seriously, the model has almost no relation
to the scaling of the auxiliary heated confinement time raux. Thirdly, the model has
weak power degradation, p " 0 3 , as compared with p ~ 0 6 for the ^ drift mode
scaling. Most importantly, the model has no favourable scaling with field strength
B or current I oc B/q, as distinct from the rj; drift mode scaling, which fails to
explain Taux only by a dimensionless factor.

IAEA-CN-47/A-Vn-l

FIRST RESULTS FROM THE HL-1 TOKAMAK
HL-l TOKAMAK GROUP(Presented by Shangjie Qian)Southwestern Institute of Physics,
Leshan, Sichuan,
China
Abstract
FIRST RESULTS FROM THE HL-1 TOKAMAK.
The medium sized HL-1 tokamak (major radius R = 102 cm, plasma minor radius a = 20 cm),
operating up to a toroidal field of 2.6 T and plasma currents of 150 kA, has produced sawtoothingplasmas of a line averaged electron density of 3.3 x 1013 cm" 3 , an average electron temperature of
700 eV and a global energy confinement time of 13 ms, without gas puffing. Its electromagnetic system,
including a particularly heavy copper stabilizing shell (thickness d = 5 cm, d/a = 1/4, d/R = 1/20)
and closely packed, wedge shaped toroidal field coils, with a comparatively uniform toroidal field andwell compensated stray fields, ensures maintaining stable discharge currents in equilibrium for aduration of over 1 s.

1.

INTRODUCTION

The HL-1 tokamak has been in exploratory plasma operation for two
experimental run seasons, April to August 1985 and May to August 1986. Preliminary experimental results obtained are reported in this paper.
HL-1 is the largest of the nine small and medium sized tokamaks currently being
operated in China (Fig. 1). Its principal parameters are:Parameter

Design value [1]Used to date

Major radius, R (cm)
Plasma minor radius, a (cm)Toroidal field, B, (T)Plasma current, Ip (kA)
Discharge duration (s)Toroidal field flat-top (s)Stray fields in plasma (T)Toroidal field power supply (MW)

102
20
5
400

102
up to 20
0.3-2.8
up to 150
up to 1.04
up to 2.8
~10-4B,
up to 66

2.5 x 10~3
160

The HL-1 tokamak is an axisymmetric device of circular plasma minor crosssection, with inductive current drive and Ohmic heating using a 1.75 volt-second flux
swing iron core transformer, a 0.7 MJ capacitor bank, a DC flywheel generator and
399

400

HL-I TOKAMAK GROUP

FIG. 1. HL1 tokamak.

a 4.5 MJ energy storage inductor pulsed for a flat-top of several hundred milliseconds [2]. The inner vacuum chamber consists of four sections of bellows
mechanically jointed and made of 0.5 mm thick Inconel-type alloy GH39 [3].
HL-1 features: (1) a copper stabilizing shell of record thickness d = 5 cm
(d/a = 1/4, d/R = 1/20), used also as the outer vacuum chamber, which, in combination with the wedge shaped and closely packed B, coils (toroidally 82% to 90%
copper covered, with total copper thickness of 13.6 cm), gives a skin eddy field
diffusion time of well over 0.5 s; (2) a comparatively uniform B, with 0.6% ripple
at the magnetic axis; (3) stray fields (with compensation) of ~ 1 0 " 4 B t ;(4) capability of long pulse operation.

2.

OPERATING RANGE AND PLASMA PARAMETERS

The present exploratory phase of plasma experiments (1985-1986) aims atbringing out the basic capability and operating regimes of HL-1. To date, HL-1 hasoperated its toroidal field in the range of 0.3 to 2.8 T with plasma currents of up to
150 kA.

IAEA-CN-47/A-VII-l401

In summary, for stable discharges in hydrogen, we have obtained:
line averaged electron density,
average electron temperature, Te
average ion temperature, T;
global energy confinement time, TE
effective ionic charge, Zeff
safety factor at limiter, q(a)ft,.

0.5-3.3 x 1013 cm" 3
700 eV~ 300 eV
13 ms
2.5-3
2.5-8

The diagnostics in use are magnetic and electric probes, bolometers, a 4 mm
microwave interferometer, optical and VUV spectrometers, hard X-ray detectors andgold-silicon barrier soft X-ray detectors, and data acquisition with a computer having
a total storage of about one billion bytes.
The average ion temperatures have been estimated by several methods,
including SIMS analysis of the depth profile of hydrogen atoms implanted in movablesilicon probes; the results, up to about 300 eV, are roughly in accord with theArtsimovich formula.
The global energy confinement time is given byrE = W/(Pn - dW/dt),

W = 37r2Ra2 n, (Te + Tj)

The experimental values, increasing with q(a), ranged from 3 to 13 ms. For medium
Zeff, medium radiation power and sawtoothing discharges with which we have beenworking, adopting Goldston's values [4] for the exponentials of R and a, we find thefollowing scaling:xE(ms) = 7.9 x 10~6 n^8 (1013 cm" 3 ) R204(cm) a io4 (cm) qO8(a)The average electron temperatures are found to scale with Ip/nea2.

3.

DISCHARGE CHARACTERISTICS: TRANSIENT ELECTROMAGNETICS

After a great deal of breakdown experiments, field adjustments and vacuum
conditioning [5, 6], some 2400 tokamak discharges have been documented
preliminarily. They fall into two categories:(1) regular discharges, more than 95% of all shots;(2) disrupted and irregular discharges.
The regular discharges were the well adjusted and properly driven discharges.
They are stable, well centred and reproducible for each fixed set of operating
conditions. Figure 2 gives a historical sketch of the improvement in plasma currents.

402

HL-1 TOKAMAK GROUP
150 kA (1 Aug. 1986)

0

20

40

60

80

100 120t (ms)

140

160

180

FIG. 2. Improvements in HL-1 plasma current.

No. 754

100

200
t (ms)

100, transformer primary current
U, loop voltage

t(ms)

-2

FIG. 3. Typical discharge in HL-1 (No. 754, hydrogen filling pressure = 8.5 x 10
B, = 2.0 T).

ton,

A typical discharge is shown in Fig. 3. Each of these discharges consists of afast current rise of 4 to 24 ms, usually with a slight overshoot, a current plateau of
150 to 200 ms, then turning into a stable, long decay of a duration that varies between
300 and 800 ms, finally terminated by loss of equilibrium. The whole discharge
lasted for about 500 ms to 1040 ms.
The plasma current rise and plateau of these discharges were similar in pulse
shape and duration, varying somewhat with the individual operating condition. The
reproducibility and the similarity between the discharge currents during the 200 ms
rise and plateau indicate that they were determined inherently by the electromagnetic
system of the device.
The HL-1 plasma equilibrium is maintained by a set of quasi-DC (25 s) externally located vertical field coils, a set of fast pulsing internally located vertical fieldcoils, and the copper shell and surrounding conductors. The toroidally directed

403

IAEA-CN-47/A-VH-l

2.5

0.4

3
0.3

q
4
5

0.2

0.1

0.5
ne R/B T (1O13 T - 1 -cm" 2 )1.0

FIG. 4. Experimentally determined stable operation regime of HL-1. (A) runaway boundary, notdetermined; (B) low q operation boundary; (C) high density operation boundary; (D) low q operation
boundary with gas puffing.

discharge current and all its surrounding conductors form a poloidally close coupledsystem, the secondary of the Ohmic transformer. The transformer primary power is
fed into and drives the plasma current Ip. During the fast current rise, becauseof dlp/dt, the main eddy currents in the copper shell, etc., are establishedsimultaneously. By the end of the plateau, the iron core has exhausted its volt-seconds
and begins working as an air core. The transformer primary circuit keeps on
inputting, but much less effectively, some power into the discharge current up to overa second.
After the plateau, the discharge current, essentially, decays freely, its detailed
shape being modified by the specific electron velocity distribution and current
profile. Experimental indications are that there is a strong component of runaway
electrons in many of the regular discharges.

4.

STABILITY

The MHD behaviour in HL-1 has been surveyed by using multi-location
magnetic probes with Fourier analysers, and hard and soft X-ray detectors.
Figure 4 shows the experimentally determined stable operation regime
for HL-1.
Figure 5(a) shows a sawtooth oscillation with a period of 0.9 ms. The relations
rs(cm) = 0.5 a(cm)/q'/i(a)andrs(ms) = n^ (1013 cm" 3 ) r^(cm) R(cm)404

HL-1 TOKAMAK GROUP

No. 1569

r=8.2 cmr=-2.7 cm

U, loop voltage
S x , X-ray (arb. units)140t (ms)

160

FIG. 5. Sawtooth oscillations from soft X-ray data: (a) steady sawtooth oscillation (No. 1569,
hydrogen filling pressure = 8 x 10~4 ton, B, = 2.0 T); (b) changeover of sawtooth oscillations
with internal disruption, accompanied by m — 1 disturbance (No. 1871, hydrogen fillingpressure = 6 x 10~4 torr, B, = 2.5 T). Negative value of minor radius refers to inboard position.

give 0.74 ms. Figure 5(b) shows the changeover of a sawtooth oscillation intoanother one, with larger amplitude and about twice the period. The event was marked
by the occurrence of an internal disruption, accompanied by a growing m = 1
disturbance.
Weak instabilities developed in a number of discharges. These usually occurred
near the end of or after the plasma current plateau. Some of them were observed to
be associated with m = 2 or m = 2 and m = 3 precursors, appearing during thecurrent rise. There were only rarely m = 3 precursors alone.
In general, we may state that the MHD activities were observed to be impurity
related. However, for operating densities of up to 3.3 X 10 1 3 cm~\ very few
discharges terminated in major (hard) disruptions. These occurred when a new
movable limiter was put in without careful precleaning or when the horizontal fields
were accidentally or intentionally off adjustment.

IAEA-CN-47/A-VH-l
5.

405

CONCLUSIONS
We arrive at the following conclusions concerning HL-1:(1) Engineering performance, fabrication and assembly of the device prove tobe gratifying, meeting and, in some respects, even surpassing the design
requirements.
(2) Exploratory plasma operation has demonstrated the basic capability of
HL-1 as designed and has provided a range of good plasmas for subsequent physics
investigation.
(3) HL-1 could be a suitable device for studies of current drive by combining
its inductive drive and long pulse capability with non-inductive drives, e.g. to study
the production and evolution of suprathermal electron populations and profile effects.
(4) Heavy conducting surroundings over a tokamak plasma, in conjunction withfine plasma profile control and adjustment, might offer some interesting possibilities
for confinement experiments.

ACKNOWLEDGEMENTS
The authors gratefully acknowledge the design and construction of HL-1 in theformer HL-1 Division headed by He Chengxun; the diagnostics from the former
Experiments Division led by Sun Xiang; early conceptual framework as establishedby Gu Yongnian's group. Numerous workers both inside and outside SWIP havecontributed to this project.

REFERENCES
[1]
[2]
[3]
[4]
[5]
[6]

LI, Zhengwu, in Controlled Fusion and Plasma Physics (Proc. 9th Europ. Conf. Oxford, 1979),
Inv. Papers, UKAEA, Culham Laboratory, Abingdon (1979) 485.
ZE, M.R., CHENG, J.S., in Fusion Engineering (Proc. 11th Symp. Austin, TX, 1985), IEEE,
New York (1986) 707.
ZHU, Y.K., et al., J. Nucl. Mater. 128&129 (1984) 784.
GOLDSTON, R.J., Plasma Phys. Controll. Fusion 26 (1984) 87.
LI, G.D., et al., Nucl. Fusion Plasma Phys. 6 (1986) 23 (in Chinese).
YANG, H.R., et al., in Plasma-Surface Interactions in Controlled Fusion Devices (Proc. 7th Int.
Conf. Princeton, 1986), Paper 8.6.

IAEA-CN-47/A-VII-2

TOKAMAK CURRENT DRIVE
VIA AC MAGNETIC HELICITY INJECTION
P.M. BELLAN, P.C. LIEWER
California Institute of Technology,
Pasadena, California
J.M. FINN, T.M. ANTONSEN, Jr.
Laboratory for Plasma and Fusion Energy Studies,
University of Maryland,College Park, Maryland
United States of America
Abstract
TOKAMAK CURRENT DRIVE VIA AC MAGNETIC HELICITY INJECTION.
With AC helicity injection, low frequency modulations of toroidal and/or poloidal magnetic fields
produce an effective steady DC toroidal electromotive force which can sustain a DC toroidal current.
Both toroidal-poloidal and poloidal-only modulation schemes are discussed. For the toroidal-poloidal
scheme, analysis (in both one-dimensional and axisymmetric toroidal geometry) of the effect of the
motion of the plasma-vacuum boundary shows that this motion may largely cancel the DC electromotive
force. Thus, if toroidal-poloidal helicity injection is to work, there must be (i) no moving boundary or(ii) a less symmetric geometry. A low power toroidal-poloidal experiment (Caltech Encore tokamak)shows a change AI in plasma current that is dependent on toroidal-poloidal phasing, but, because of
an offset, AI is always negative. The poloidal-only scheme is based upon the experimentally observed
anomalous skin penetration during the startup phase of tokamaks. Analytic and numerical results indicate feasibility, and scaling of steady state current and power dissipated with frequency and plasma
resistivity is presented. This scheme is potentially more efficient in terms of current generated per unit
power dissipated than other current drive schemes. It is shown that the resistive MHD activity responsible for anomalous current penetration (and therefore required for the success of this current drive
scheme) must be fully three-dimensional. A comparison of the effectiveness of this scheme with thatof F-G pumping in reversed field pinches is presented.

1. INTRODUCTION
AC magnetic helicity injection has been recently proposed [l]-[3] as amethod for obtaining steady state toroidal current drive. This method utilizes
AC modulations of the poloidal and/or toroidal magnetic fields at very low frequencies (G)«a>c;). Modulations of a few percent at frequencies of 101 -102 Hz
could drive the current in a fusion reactor provided there is anomalous current
penetration. In this paper we discuss both toroidal-poloidal modulation
schemes (Caltech authors, Sections 2-3) and poloidal-only schemes (U. Maryland authors, Section 4).
407

408

BELLAN et al.

2. EFFECT OF A MOVING BOUNDARY ON HELICITY INJECTION
Bellan [4] interpreted toroidal-poloidal AC helicity injection as resulting from the beating of two normal modes of the tokamak plasma: (1) modulation of the plasma minor radius driven by toroidal field modulation and (2)classical resistive diffusion (having skin depth penetration) driven by poloidal
field modulation. This beating would produce a DC toroidal emf which could
drive a DC toroidal current. The process was strictly one-dimensional (radial
dependence only) and the resulting DC toroidal current was predicted to be askin current; reconnection (a three-dimensional process) would be necessary
to make the skin current penetrate.
References [l]-[3] implicitly assumed that plasma continuously filled
the vacuum chamber right up to the wall. Yet, Ref. [4] showed that toroidal
modulation should compress the plasma minor radius, so that plasma should
detach from the wall, leaving behind a vacuum region. We have studied [5] theeffect of the motion of the plasma-vacuum boundary using a one-dimensionalslab model and found that this motion makes the current drive significantly
less than predicted by Refs [l]-[4].
To obtain the one-dimensional slab model, zero-pressure resistive
MHD equations in rectangular coordinates were used with x (radial) dependence only. The radial velocity Ux of the plasma was the £*xB* velocity associated with the time-varying, but spatially uniform, toroidal field (longwavelength limit of the compressional Alfven wave). The evolution of thetotal (DC plus AC) poloidal magnetic field By(x,t) was described byThis equation is valid only in the plasma, but not in the adjoining vacuum
region. By including oscillation of the plasma boundary, this model conserves
total plasma throughout the cycle. To complete the model, equations for By in
the vacuum region between the plasma and the wall and in a region
corresponding to the central hole of a tokamak are also included. By cannot be
specified at the wall because it contains an unknown DC component. Instead,the current in the external poloidal field coil is specified, which then gives ajump in By(x,i) at the coil.
The solution of this system of equations and boundary conditions
shows that the driven DC poloidal magnetic field scales as ByBj Bo and thus,
the driven DC poloidal field is always much smaller than By. This is in contrast to the expectations of Ref. [4] which suggested that the driven DC
poloidal field could be comparable to or larger than By. When Eq. (1) was

IAEA-CN-47/A-Vn-2

409

solved with the effect of the moving boundary neglected (e.g. assuming
plasma filled the region up to the wall throughout the oscillation cycle), then asignificantly larger DC poloidal field was obtained. Note that plasma is notconserved when the moving boundary is neglected. In a real plasma experiment there is never a true vacuum region between the plasma and the wall
because of ionization and cross-field diffusion. However, this plasma is likely
to be highly resistive and thus the vacuum treatment may be more realistic.
The importance of a moving boundary is also manifested [6] by anapparent discrepancy between the helicity conservation equations of Jensen
and Chu [3]— + U$'(0 + E?xX) = - 2\t\T'•&dir

(2)

at

andofMoffatt[7]— + \dS •£*(<(> - X-&) = r\\d 3r (if- V2^ + X- V2B*)

(3)

dt

which turn out to differ by a Lagrangian (i.e. moving boundary) term. Using
Moffatt's equation (which incorporates a moving boundary) it is possible toshow [6] that no DC current can be driven in an axisymmetric plasma having
circular flux surfaces. It is still possible to inject net helicity in this situation,
but the injected helicity does not drive DC currents because it is exactly consumed by the increase in helicity dissipation associated with the oscillating
fields producing the injected helicity.
It is crucial to find whether (i) non-conservation of plasma makes it
possible for the plasma to be compressed without producing a moving
plasma-vacuum boundary, or (ii) non-axisymmetric mode structures would
allow helicity injection in excess of the increased helicity consumption associated with the modes, because then the optimistic predictions of Refs [l]-[4]would hold.
3. EXPERIMENTS
Experimental studies of helicity injection are underway on the Caltech
Encore high repetition rate research tokamak. Typical parameters are: H
plasma, n-IO12 cm"3 , Tj-l-lO eV, r e -5-15 eV, B-100-300G, 7-1-3 kA, shot
duration 1-3 msec, repetition rate 10-30 shots/sec, R =38 cm, a=10 cm,
poloidal and toroidal modulations 1-10% at 24 kHz, gap long way around in
one torus quadrant.
Preliminary results, obtained at low power using signal averaging techniques, indicate a change in plasma current A/ dependent on the relative phas-

410

BELLAN et al.

ing of the toroidal and poloidal modulations. However, A/ always has a negative offset so that only phase-dependent reductions in plasma current are
observed. In agreement with theoretical prediction, the phase dependence flips
sign when the DC toroidal field is reversed. However, for certain values ofplasma current or toroidal field, the flipping of sign fails to occur (this is associated with a resonance in A/).

4. CURRENT DRIVE USING POLOIDAL FIELD MODULATION ONLY
Motivated by the phenomenon of anomalous current penetration during
tokamak startup, we investigate theoretically whether steady state current canbe driven by oscillating the poloidal (Ohmic heating) coils only. If double
tearing modes indeed cause anomalous penetration, the induced skin current
parallel to the bulk current may penetrate anomalously, whereas the skincurrent induced in the other half of the cycle should be stable to tearing modes,
and so penetrate only classically. This scheme is related to helicity injection
by toroidal-poloidal modulation [l]-[3] (also called F-9 pumping), but has theadvantage that only one field (poloidal) is modulated. In the present scenario,
the injected helicity is provided by the relaxations that are assumed to occur
preferentially in one part of the cycle. This scheme is inherently low frequency because full nonlinear resistive MHD activity must occur within one
cycle.
In our model the plasma is treated classically throughout the quiescentpart of the cycle, in ID (cylindrical) geometry.
The double tearing
activity is idealized by flattening the j , profile (q profile) while conserving themagnetic helicity and the poloidal flux at the plasma edge, i.e. the flux linking
the torus. The toroidal flux is assumed to remain constant throughout the process. Using this model it can be shown, by analytic and numerical means, thatsteady state current can indeed be driven, i.e. that no oscillation of the toroidal
field is required. From the point of view of helicity injection, current can be
driven against resistive decay, even though no helicity is injected by external
circuits, forthe following reason: when the skin current induced by the oscillating OH is negative, i.e. opposite to the main current, the rate of change ofhelicity dKldt = -l\i?'&dV is positive. On the other half of the cycle, the skincurrent penetrates, conserving helicity, so that the net change due to resistive
decay of skin currents is positive.
The scaling of this scheme with frequency and resistivity can be
estimated as follows. Balancing the resistive decay of helicity with the injection of helicity due to the decay of skin current, we find

IAEA-CN-47/A-Vn-2

411

_

p

Thus, for a given flux swing y the mean current Tp increases as the square rootof frequency. The mean current is also enhanced if the edge resistivity r\cgreatly exceeds the average resistivity <TI>. Here, a is the minor radius. If theflux swing v is the limiting factor, the optimum frequency is the highest frequency for which the double tearing modes can completely reconnect in one
cycle. The ratio of the mean current to the magnitude of the oscillating skincurrent can also be estimated to be Tpl l/~ I ~T|e/ <TI>, which is independent of
frequency and can be made large if the edge resistivity is considerably higher
than the average. The scaling for Tpl P, the ratio of steady state current driven
to the power deposited, can be obtained by similar arguments. The power
deposited in the plasma is P =\i?>d?-ridS ~®fp\fL, where L is the plasma length
L=2idt. Therefore, we findP<r\>%L

Tp

where A, =V-qe/ w is the skin depth. Recalling that PII for Ohmic heating is
IpRI Ip~<x\>IpLl na1, we see that the power dissipated in oscillating current
drive, compared to pure inductive (OH) current drive producing the same
current, scales as O<TI>/ (A,ne). This dissipation is not much larger than pure
Ohmic dissipation.
We have also investigated whether anomalous skin penetration and any
oscillating current drive scheme based upon it can be due to single helicity mln
modes, behaving as exp(/m 8- in ij»), where 6,4> are poloidal and toroidal angles,
respectively, or whether multiple helicity (3D) effects are required. We conclude that single helicity activity with mS2 cannot explain the results and suchactivity with m = 1 modes is unlikely to do so. This conclusion is based upon
the observation that the £7xB* term for modes with m>2 cannot change the flux
at the magnetic axis (the Cowling theorem). Modes with m=\ can experience achange in flux at the magnetic axis, by Kadomtsev reconnections, but double
Kadomtsev reconnections reduce the flux at the axis. That is, if anomalous
current penetration is due to resistive MHD modes, fully three-dimensional
behavior is required. Thus, magnetic surfaces may not exist in some areas in
the plasma. This may lead to a natural flattening of the current profile, but may
also allow transport of heat along ergodic magnetic field lines. It is an open
question whether this latter phenomenon may lead to very poor energy
confinement or whether the region of destroyed magnetic surfaces may be
confined to a part of the plasma volume.

412

BELLANetal.

It is interesting to compare this oscillating current drive proposal fortokamaks to F-0 pumping for RFP's. As we have discussed, the former
scheme depends upon preferential MHD relaxation activity in one part of thecycle, whereas the latter scheme depends upon external injection ofhelicity. The helicity absorbed by the plasma in a cycle in F-9 pumping is&Kdp =47t2z.ev*, where \j/ and o are the amplitudes of the oscillating poloidal
and toroidal fluxes, respectively. Also, e is an efficiency factor that is typicallysomewhat less than unity and depends critically upon the edge resistivity andthe number of relaxations per cycle [8]. This efficiency is related to the loss ofhelicity in F-0 pumping due to decay of skin currents.
On the other hand,for this tokamak scheme, the helicity injected is of orderA^tok =4n2LflJa\jWCn(a)/ <o), from the scaling arguments above. From these
relations it is clear that the current drive scaling is quite favorable compared tothat for F-8 pumping. Essentially, M^ is first order in oscillating fluxes andA£rfp is second order. It appears, therefore, that there is little advantage tooscillating the toroidal field in a tokamak for this scheme.
In summary, we have shown that if relaxations flattening the skinprofiles occur preferentially in that part of the cycle where the skin current isparallel to the bulk current, current drive can be achieved. Estimates based
upon nominal parameters give 1 MA of steady state current driven per voltsec(amplitude) of oscillated flux, and dissipation comparable to Ohmic.
ACKNOWLEDGEMENT
This work was supported by the United States Department of Energy.

REFERENCES[1] BEVIR, M.K., GREY, J.W., in Proc. Reverse Field Pinch Theory Workshop, Vol. 3, Los
Alamos Scientific Laboratory (1981) paper A-3.
[2] SCHOENBERG, K.F., et al., Phys. Fluids 27 (1984) 548.
[3] JENSEN, T.H., CHU, M.S., Phys. Fluids 27(1984) 2881.
[4] BELLAN, P.M., Phys. Rev. Lett. 54 (1985) 1381.
[5] LIEWER, P.C., GOULD, R.W., BELLAN, P.M., One-dimensional modeling of DC current
drive by AC helicity injection, submitted to Phys. Fluids.
[6] BELLAN, P.M., The adverse consequences of a moving vacuum-plasma boundary on axisymmetric helicity injection, Phys. Rev. Lett. 57 (1986) 2383.
[7] MOFFATT, H.K., Magnetic Field Generation in Electrically Conducting Fluids, Cambridge
University Press, Cambridge (1978) 43-45.
[8] FINN, J.M., Phys. Fluids 29 (1986) 2630.

IAEA-CN-47/A-VII-3

STUDY OF ENHANCED CONFINEMENT
AND q = 2 DISRUPTIVE MODESWITH FAST RISING CURRENT
IN THE STOR-1M TOKAMAK
H. KUWAHARA, 0 . MITARAI1, E.J.M. VAN HEESCH,
M. EMAAMI, A.H. SARKISSIAN, S.W. WOLFE,
A. HIROSE, H.M. SKARSGARD
Plasma Physics Laboratory,
University of Saskatchewan,
Saskatoon, Saskatchewan,
CanadaAbstract
STUDY OF ENHANCED CONFINEMENT AND q = 2 DISRUPTIVE MODES WITH FAST
RISING CURRENT IN THE STOR-1M TOKAMAK.
The STOR-1M tokamak can operate in three different modes — as a normal tokamak, as atokamak with a fast current pulse superimposed on the normal plasma current, and as an AC tokamak.
Magnetohydrodynamic activity is observed during the fast pulse which gives access to q(a) as low as
1.5. The fast pulse also produces delayed increases in density and temperature, possibly as a result of
improved containment properties due to current profile modification. Recent results on AC operation
with gas puffing are also presented.

I.

INTRODUCTION

STOR-1M is a small tokamak in which a fast current pulse(~20 ysec duration) can be superimposed on the normal plasma
current. The current pulse provides turbulent heating of theplasma. In addition it can provide a rapid variation of q(a)over a wide range, passing quickly through q-values that would
normally lead to termination of the discharge. MHD activity
can thus be observed on this fast time-scale, particularly in
the important q=2 range. In addition, the fast pulse can provide, on a longer time-scale governed by the magnetic diffusion
time, a modification of the current density profile which can
influence the plasma properties. Finally, the ST0R-1M tokamak
has also been operated in an AC mode; recent results on thebeneficial effects of gas puffing on the AC operation are
included in this report.
1

Department of Electrical Engineering, Kumamoto Institute of Technology,
4-22-1, Ikeda, Kumamoto 860, Japan.

413

414

KUWAHARA et al.

«™m
VISIBLELOOP S O F T X . R A Y
SPECTROMETER\ DETECTORSVIEWING
W
DUMP

/ / <

o

HARD X-RAY
0 E T E C T 0 R (|)
0 E T K T O R

GAUGE

LIMITERMAGNETICCOILS

HARO X-RAY
DETECTOR (2)PHOTO TUBE
and 4mm
MICROWAVE
INTERFEROMETERPOSITION GAS SUPPLY"SENSING COILS^ v n i

TMP

FIG. 1. Schematic diagram of the STOR-1M tokamak including diagnostics.

ST0R-1M has major and minor radii of 22 cm and 3.5 cm
respectively. The duration of the fast pulse is typically
20 ysec and its maximum ramp rate exceeds 1 0 9 A/sec. The
radial plasma position is controlled by a position feedback
system,except during the turbulent heating pulse when thevacuum chamber acts as a conducting shell (field penetration
time - 130 ysec). The duration of the ohmic discharge is about
5 msec. Typical plasma parameters are: initial filling pres9 x 10"" Torr of hydrogen, I = 4 to 7 kA, ne < 3 x 1 0 1 3
sure
3
cm" , T < 100 eV, Z e f f = 2 to 3, qXa) > 3.5 and B t = 0.5 L O T . "Diagnostic instruments include a 4 mm microwave interferometer, hard and soft x-ray detectors, a charge exchange
analyzer and an optical spectrometer in addition to various
standard magnetic coils as shown in Fig. 1.
2.

MHD ACTIVITY DURING THE FAST PULSE

Previous studies of tokamak operation at low q-values showthat some tokamaks can operate [1,2] below q=2 while others
exhibit major disruption [3,4] for q ~ 2 to 2.5. The range of
unstable q-values and the growth rate of the disruptive modeare still not well understood.
The turbulent heating pulse in STOR-1M is superimposed on
the flat top portion of the ohmic discharge current where q(a)= 5.6 at B t = 1.0 T, or q(a) = 4.5 at B t = 0.8 T are chosen for415

IAEA-CN-47/A-V1I-3-With
Without
Fast Pulse

(kA)

0

1

2
Time [m sec]

3

4

FIG. 2. Typical discharges with and without the fast turbulent heating pulse.

P(kA)

t (10 >jsec/div)
FIG. 3. Current and voltage traces during the fast pulse; large voltage spikes are observed for1.85 £ q{a) •& 2.1.

416

KUWAHARA et al.

the initial conditions. Typical discharges with and without
turbulent heating are shown in Fig. 2. By varying the charging
voltage of the turbulent heating capacitor bank, the current
ramp rate can be changed up to a maximum of 1.8 x 1 0 9 A/sec.
As shown in Fig. 3, when q(a) is above 2.1, there are nolarge voltage spikes and the discharge is stable. Voltage
spikes appear below q(a) = 2.1 and disruptions begin to be
noticeable on the plasma current trace. When the plasma remains
below q(a) = 2.1 for a longer time, higher voltage spikes areobserved and complete termination of the plasma current occasionally takes place. The plasma becomes stable, with no voltage spikes, below q(a) = 1.85 if the ramp rate exceeds 4 x 1 0 8
A/sec at q(a) = 2.1; stable discharges at q(a) as low as 1.5
have been achieved. After the peak current, voltage spikes areapparent as the plasma current decrease takes q through the unstable region (1.85 < q(a) < 2.1).
In the unstable region, MHD activity is monitored with magnetic coils. Plasma disruption is preceded by rapid growth of amagnetic perturbation whose growth time (~500 nsec) is still 100
times longer than the Alfve'n time. The largest magnetic perturbation observed is B Q / B Q ~ 0.03% at the coil (r = 5 cm). The
magnetic island width (m=2/n=l) estimated from this amplitude[5] is < 5% of the minor radius, insufficient to cause disruption.
In the LT-4 tokamak an n=l mode was detected, whose amplitude was also not large enough to cause the observed disruption[4]. (Although its m-number was not determined, the central
portion of the hot plasma core showed an asymmetric profile
before a major disruption.) Thus the growth of magnetic perturbations prior to disruption may not play an important role incausing the disruption.

3.

PLASMA HEATING

So far, noise problems have prevented us from measuring theelectron and ion temperature increase during the heating pulse
itself. An indication of effective heating is found from measurements of the resistivity at peak current (Fig. 4 ) . The
resistivity increase with applied voltage demonstrates theenhanced power input (P. a A l n ; n > 2) due to current-driven
turbulence.
Electron temperature measurements are available after 0.15
msec from the heating pulse. For moderate applied voltage thetemperature reproducibly decays within ~ 0.3 msec, from ~ 200 eVdown to the value (~ 100 eV) measured without the fast pulse.

417

IAEA-CN-47/A-VII-3
\

'

I
60

L

I

^

I

'

I

in

J
0

I
20

I

I
40

1

i 1 ,
80
100
Ep(V/m)120

140

FIG. 4. Plasma resistivity during the fast pulse as a function of the applied voltage( O — in the ohmic discharge).

When the applied voltage is increased to > 60 V/m the electron
temperature, instead of decaying with time, shows a double-hump
behaviour, the second peak coinciding approximately with theobserved peak in n e (t) (see Fig. 2 ) . It is as though the containment is improving, leading to an increase in the energy density nT. The implied improvement in containment is generally
greater, the larger the applied voltage, although reproducibility
deteriorates at the highest voltages. (q(a) > 2.1 for the rangeof voltages employed.)
A possible explanation for enhanced global confinement is
the change in the current density profile produced by the fastcurrent pulse. Until the magnetic diffusion (T ~ 0.15 - 0.6msec) can redistribute the current perturbation, the current
density profile is flattened somewhat.
A similar observation of delayed "heating" following the
turbulent heating pulse has been observed in the TORTUR tokamak[6]. In that case it has been attributed to a conversion ofmagnetic energy into thermal energy, supposing that the current
channel expands again from a contracted state which is assumed
to be formed just after the application of the fast pulse.
Accurate measurements of the time variation of the current density profile may be required in order to choose between the two
proposed explanations.
The case of a turbulent heating current pulse applied in
the opposite direction to the ohmic heating current was alsostudied in ST0R-1M. Although there is a small density increase

418

KUWAHARA et al.

1

2

3
4
5
TIME (msec)

FIG. 5. AC tokamak discharge with gas puffing.
I. (2.64 kA/div), V, (10 V/div), ne (2 x 10n cm'''/step), B, = 1 T.

after the pulse, no plasma heating is observed with thisarrangement. Apparently the current density profile producedby the negative fast pulse is not favourable either for heatingor confinement.

4.

AC OPERATION WITH GAS PUFFING

Since 1984, AC operation has been studied in STOR-1M as apossible approach to quasi-continuous tokamak operation. Stable reproducible AC discharges have been obtained without disruptive behaviour or plasma extinction during current reversal[7], even with reversal times ZlpoCdlp/dt)"1 an order of magnitude longer than the energy containment time. Synchronizationof the zero crossing time for the vertical magnetic field andthe plasma current is the most important single factor inachieving AC operation.
The original experiments, with no gas puffing, exhibited adensity decay to ~ 2 x 1012 cm-3 at reversal,followed by acm ) in the reverse half-cycle.
slight increase (to ~ 4 x
Gas puffing has now been employed in an attempt to improve thedensity behaviour. Typical results are shown in Fig. 5. With
appropriate activation of the piezoelectric valve the electron
density now reaches approximately the same value in the firstand second half-cycles of AC operation. It is also possible toIAEA-CN-47/A-VII-3

419

achieve current reversal at higher plasma currents ( I p 0 < 8 kA
instead of 6 kA). The density at current reversal, however isstill about the same as without gas puffing.
This work was supported by the Natural Sciences and Engineering Research Council of Canada.

REFERENCES[1] MIRNOV, S.V., SEMENOV, I.B., Sov. Phys. — JETP 33 (1971) 1134.
[2] DIVA Group, Nucl. Fusion 20 (1980) 271.
[3] TOI, K., et al., Nucl. Fusion 22 (1982) 465.
[4] KUWAHARA, H., et al., private communication, 1986.
[5] TFR Group, Rep. EUR-CEA-FC-1151, Groupe de recherches sur la fusion controls, Association Euratom-CEA, Centre d'6tudes nucteaires de Fontenay-aux-Roses (1982).
[6] DE KLUIVER, H., et al., "Plasma heating in the TORTUR tokamak by low-frequency toroidal
current pulses", Controlled Fusion and Plasma Physics (Proc. 12th Europ. Conf. Budapest,
1985), Vol. 2, European Physical Society (1985) 272.
[7] MITARAI, O., et al., Bull. Am. Phys. Soc. 29 8 (1984) 1337; Rep. PPL-75, Plasma Physics
Laboratory, University of Saskatchewan (1984).

IAEA-CN-47/A-VH-4

COHERENT AND TURBULENT FLUCTUATIONSIN TFTRK. McGUIRE,
V. ARUNASALAM,
M.G. BELL,
M. BITTER,
W.R. BLANCHARD, N.L. BRETZ, R. BUDNY, C.E. BUSH1, J.D. CALLEN 2 ,M. CHANCE, S.A. COHEN, S.K. COMBS1, S.L. DAVIS, D.L. DIMOCK,
H.F. DYLLA,
P.C. EFTHIMION,
L.C. EMERSON1,
A.C. ENGLAND1,
H.P. EUBANK,
R.J. FONCK,E. FREDRICKSON,
H.P. FURTH,
G. GAMMEL, R.J. GOLDSTON, B. GREK, L.R. GRISHAM, G. HAMMETT,
R.J. HAWRYLUK,W.W. HEIDBRINK3,
H.W. HENDEL 4 ,K.W. HILL,
1
E. HINNOV,
S. HIROE ,H. HSUAN,
R.A. HULSE,
K.P. JAEHNIG,
D. JASSBY, F.C. JOBES, D.W. JOHNSON, L.C. JOHNSON, R. KAITA,
R. KAMPERSCHROER,
S.M. KAYE,
S.J. KILPATRICK,
R.J. KNIZE,
H. KUGEL, P.H. LaMARCHE, B. LeBLANC, R. LITTLE, C.H. MA 1 ,J. MANICKAM,
D.M. MANOS,
D.K. MANSFIELD,
R.T. McCANN,
M.P. MCCARTHY,
D.C. McCUNE,
D.H. MCNEILL,
D.M. MEADE,
S.S. MEDLEY,
D.R. MIKKELSEN,
S.L. MILORA1,
D. MONTICELLO,
A.W. MORRIS5,
D. MUELLER,
E.B. NIESCHMIDT6,
J. O'ROURKE7,
D.K. OWENS, H. PARK, W. PARK, N. POMPHREY, B. PRICHARD,
A.T. RAMSEY, M.H. REDI, A.L. ROQUEMORE, P.H. RUTHERFORD,
N.R. SAUTHOFF,
G. SCHILLING,
J. SCHIVELL,
G.L. SCHMIDT,
S.D. SCOTT, S. SESNIC, J.C. SINNIS, F.J. STAUFFER8, B.C. STRATTON,
G.D. TAIT,
G. TAYLOR,
J.R. TIMBERLAKE,
H.H. TOWNER,
M. ULRICKSON,
S. VON GOELER,
F. WAGNER9,
R. WIELAND,
J.B. WILGEN1,
M. WILLIAMS,
K.L. WONG,
S. YOSHIKAWA,
K.M. YOUNG, M.C. ZARNSTORFF, S.J. ZWEBEN

Princeton Plasma Physics Laboratory,
Princeton University,
Princeton, New Jersey,
United States of America

1

Oak Ridge National Laboratory, Oak Ridge, TN, USA.
University of Wisconsin, Madison, WI, USA.
3
GA Technologies Inc., San Diego, CA, USA.
4
RCA David Sarnoff Research Center, Princeton, NJ, USA.
5
Balliol College, University of Oxford, Oxford, UK.
6
EG&G Idaho, Inc., Idaho Falls, ID, USA.
7
JET Joint Undertaking, Abingdon, Oxfordshire, UK.
8
University of Maryland, College Park, MD, USA.
9
Max-Planck-Institut fur Plasmaphysik, Garching, Fed. Rep. Germany.
2

421

422

McGUIRE et al.
Abstract

COHERENT AND TURBULENT FLUCTUATIONS IN TFTR.
Classification of the sawteeth observed in the TFTR tokamak has been carried out to highlightthe differences between the many types observed. Three types of sawteeth are discussed: "simple","small", and "compound". During the enhanced confinement discharges on TFTR, sawteeth related
to q = 1 are usually not present, but a sawtooth-like event is sometimes observed. Beta approaches the
Troyon limit only at low q^, with a clear reduction of achievable /3n at high q^,. This suggests thata /3 limit, rather than the Troyon-Gruber limit, applies at high qc , in the enhanced confinement discharges. These discharges also reach the stability boundary for n — oo ideal MHD ballooning modes.
Turbulence measurements in the scrape-off region with Langmuir and magnetic probes show strongedge density turbulence, fi/n = 0.3 - 0.5, with weak magnetic turbulence, B s /B s > 5 x 10~ 6 ,measured at the wall, but these measurements are very sensitive to local edge conditions.

1.

Introduction

Near-term achievement of ignition in tokamaks depends on
overcoming anomalous heat conduction
and/or control of thecentral energy transport due to sawteeth. Sawteeth in TFTR havebeen studied (Section 2) in an attempt to understand the mechanismthat causes the sawtooth crash and gives rise to enhanced heatpulse propagation.
The MHD activity observed in enhanced
confinement discharges (locally known as supershots) and theconstraints imposed by the Troyon and the B limit are discussed
in Sections 3 and 4.
In Section 5, a summary of turbulencemeasurements carried out on TFTR is presented, and the correlation
between edge turbulence and gross confinement is examined.
Finally in Section 6, we summarize the results of the coherent andturbulent fluctuations measured in TFTR.
2.

Sawtooth Activity

Three types of sawtooth activity have been identified in
TFTR: (1) simple or normal sawteeth, (2) small sawteeth, and (3)compound sawteeth. Simple sawtooth oscillations, as observed inmost tokamaks, are a series of m/n = 0/0 relaxations, each with an
m/n = 1/1 precursor mode, where m and n are poloidal and toroidal
mode numbers, respectively.
Figure 1a shows a simple sawtooth
which is observed in TFTR plasmas with a small q = 1 radius andhigh q . Small sawteeth (Fig. 1b) are similar to simple sawteeth
in period, but the m/n = 1/1 mode is different.
While theprecursor oscillations before the m/n = 0/0 crash are usually notpresent or have very small amplitude, the successor oscillations
have a large amplitude.
Compound (sometimes called "double")sawtooth activity (Fig. 1c) has been observed recently in many
tokamaks [1-3], Each compound sawtooth consists of a subordinate
relaxation followed by a large (up to 20% in central T&) main
relaxation with a smaller inversion radius.
Compound sawteeth
appear frequently in ohmically heated plasmas with high plasma

IAEA-CN-47/A-VII-4

423(a) Simple Sawtooth(b) Small Sawtooth
!J
r»0cm

3400

,

3420,

3440

3460
3480
TIME (msec)

3400

3420

rmOcm

20

30
3410

34I5

3425

3420

3435

3430

t TIME(msec)
Complete

I
Complete

Relaxation

Relaxation

3440

3445
TIME (msec)

2565

2570

(c) Compound Sawtooth

2500

2525

2530

2535
Subordinate
Relaxation

2520

2540

2540

2560

2555

2580

2560
Main
Relaxation

FIG. 1. Soft x-ray emissivity signals, indicating the three main types of sawteeth observed in TFTR.
(a) Simple or normal sawtooth, (b) small sawtooth, and (c) compound sawtooth.

424

McGUIRE et al.

1

1

1

1

1

1

1

6-

1

51

o

•

•/^

•

-

O

3-

~ 2
E
I 1
I
1
2
3
4
( r c r ) DECAY TIME (IO"4sec)

1

F/G. 2. Comparison between the growth time of the m = I mode at the crash of the sawtoothand the crash time of the sawtooth, as measured by the central soft x-ray signal.

current (I > 1.0 MA) and/or high density (iie > 2.0 x 10m" ).
Sawteeth may switch from compound to small during the steady phaseof a discharge, although global parameters such as averaged
temperature, density, Z e - f , and even the inversion radius of thesawtooth do not change [2j. This suggests that small or localizedchanges of plasma parameters affect the characteristics ofsawteeth significantly. The subordinate relaxation and the small
amplitude of the m = 1 precursor indicate that compound sawteeth
and small sawteeth cannot be explained by the Kadomtsev model[4],
It has been found that the duration of the crash (T cr ) is
proportional to the final growth time (T ) of an m = 1 mode during
the crash phase itself (Fig. 2). Note that Tc r appears to becompletely determined by T . Possible explanations include fasttransport or reconnection due to; a) small perturbations at the q= 1 surface, b) two g = 1 surfaces [5] or c) a flat q profile with
q Q close to 1 [1,6],
3.

MHD Activity in Enhanced Confinement Discharges

Recent experiments on TFTR at low I and high-power balanced
neutral beam injection (NBI) have achieved enhanced confinement
relative to the predictions of L-mode scaling [7,8].
Thesedischarges are remarkable from the MHD viewpoint, the most obvious
feature being the lack of sawteeth during NBI for q c y l > 5. Thismay be due to the removal of the g=1 surface from the plasma (somedata show r*+ 0 ) , but qcv-i < 5 enhanced confinement discharges
with sawteeth have r^ nv unchanged.
During enhanced confinement
plasmas with q -^ > 5, there are occasional sawtooth-like events
which are similar to the subordinate relaxations described in

425

IAEA-CN-47/A-VU-4-30-0.5 -0.4 -0.3

-0.2 -0.1 0
0.1
RADIUS (m)

0.2

0.3

0.4

0.5

FIG. 3. Plot ofAA/A of the sawtooth crash versus radius, for an m/n = 1/1 sawtooth in an L-mode,
Ip = 900 kA discharge with 3.4 MW of counter NBI, and an m/n = 2/1 sawtooth in an Ip = 800 kA
enhanced confinement discharge with 10 MW of balanced NBI. The m = 2 sawtooth-like event has aninversion at 0.2 m which is close to the calculated q = 2 radius.

Section 2.
They have only a small perturbation on axis and an
inversion close to the calculated location of the g = 3/2 or 2/1
surfaces. Figure 3 shows an m = 2 sawtooth-like event which wasobserved in: an enhanced confinement discharge with I = 0.8MA,
q j = 7.6 and 10MW of balanced NBI; the calculated q = 2 radius
for this discharge was about 0.25m. In Pig. 3 the line-averaged
soft x-ray inversion radius was about 0.20m for the m = 2
sawtooth, which was larger than the m = 1 inversion radiusobserved at an m/n = 1/1 sawtooth in an I = 0.9MA, cr ^ = 7.3,
and P^ni = 3.6MW counter-injection L-mode discharge.
The mode
amplitude and the phase on the x-ray diodes clearly show an m = 2
mode structure, also a small amplitude odd
m-mode was present
near the center. The magnetic signals at the wall show a clear
m/n = 2/1 mode with no coupled mode structure.
These m = 2
sawteeth are infrequent, with an irregular period, and they do notappear to affect the confinement of the discharge. The off-axis
sawtooth events can occur during a period of continuous m/n = 3/2
or 2/1 activity (detected by magnetic measurements at the wall)with no interruption of the oscillation.
Similar type sawteeth
have also been observed in pellet injection discharges during aperiod of continuous m = 1 mode activity.
For identical operating conditions, enhanced confinement
discharges can vary substantially, depending on the MHD activity.
Figure 4 shows two such discharges, one with m/n = 2/1 activity,
the other without MHD activity.
The correlation of the MHD
activity with the drop in total stored energy and neutron emission
is very clear. This can occur with m/n = 2/1 alone, or m/n = 3/2
alone. Figure 5 shows the coherent MHD activity versus the rate

426

McGUIRE et al.

1

1

1

1

7

j

§

1

ml
•

\

g

s •00

(J)

O"

*

II

ii

u
>~
00

QT

Q.
•—<

1

1

1

1

1
OJ

A1ISN3Q 39VM3AV—(SlINfl "8aV) AIIAI13V AONdlWO

IAEA-CN-47/A-VII-4

427

dwi I A /dt (MW)FIG. 5. MHD activity between 5 and 55 kHz versus dW°^A/dt, 50 ms before the end of the beam pulse.
These data include only enhanced confinement discharges. No similar correlation of MHD activity andconfinement is seen for L-mode-type discharges.

of change of stored energy (calculated from the diamagnetic
signal) 50ms before the end of the beam pulse. It is clear that alarge amplitude of MHD activity is correlated with negative
dWj^
/dt.
Prom delta-prime calculations on the j(r) profiles
calculated by the TRANSP code [9] it is found that the profiles
are sometimes unstable to the m/n = 2/1 and/or m/n = 3/2 modes.
This implies that the j(r) profile may play an important role in
the energy balance inside the q = 2 surface for the enhanced
confinement discharges on TFTR [10],At lower q cy ^ << ~ 7)> bursts of m = 1 activity are seen in
the center of the plasma on the soft x-ray detectors. This mode
couples out as m/n " 6/1 on the Mirnov coils, similar to the
fishbone oscillations observed on PDX. This is in contrast withthe 2/1 and 3/2 modes which appear as m/n = 2/1 and 3/2,
respectively, on the Mirnov coils, even though the mode amplitude
peaks at a similar radius (0.1 - 0.3 m ) . There is no measurable
effect on the neutron emission, due to m = 1 activity at the levels
observed to date, although we have observed coherent (5 kHz)oscillations in the charge-exchange neutral signals.
Enhanced confinement discharges which disrupt can lose mostof the plasma energy in as little as 20 - 100 us, with only a low
amplitude or no detectable MHD activity preceding them.
Thetemperature on the inner wall, as measured by 1.0 ym radiation,
can rise to over 1200°C in 20 us after a high 6 disruption. In a

428

McGUIRE et al.

0.5

10
Icy I

FIG. 6, The present enhanced confinement regime ofTFTR on a plot of&n versus q t The square symbols show discharges that disrupted during the neutral beam pulse. Also shown is a curve representing
the high q t data better than the optimized Troyon limit.

few disruptive discharges, a 40 kHz mode growing on a 50 us time'scale, with an m = 1 structure in the center and m > 4, n = 1 on
the magnetic signals, has been observed. Although the bulk of theplasma energy is lost rapidly, the current disruption occurs up to
0.5 s later, after a narrowing of the T e and X-ray emission
profiles.
It is also observed that few major disruptions takeplace for q^,r1 < 5; here the continuous 3/2 and/or 2/1 modes arer»dCythe energy is lost slowly, as in Fig. 4.
detected anc
4.

BETA Limits

Enhanced confinement discharges with q cy i < 5 have values ofnormalized toroidal beta, 3 = $A/$C
W „ = u Q I /aBo,), which approach
the limit found from optimized ideal MHD kink stability theory[11,12],This limit has previously been confirmed on otherdevices with NBI (Dili, PDX, ASDEX, ISX-B, for example).
The
experimental limit from these devices is 3 n = 2.2 - 2.5,
approximately independent of plasma shape, but only for a small
range of aspect ratio and qcy^«
Figure 6 shows B n for theenhanced confinement regime of TFTR versus q c v l .
Diamagnetic
3 i D I A (assuming a circular cross-sectional plasma* is used in Fig.
6 and, since the NBI is tangential, this underestimates the total 3(< 10% at high 3 ). It is seen that there is a clear fall of the
There is little published data fromachievable 3. wft
th q cylother auxiliary heated tokamaks in the high-6 n high-q region[13],
At high q c y l , the limit appears to be disruptive, but atq
1 < 5 the 6 n limit is associated with coherent MHD activity.
Attempts to reach higher 3 n by reducing (I , B± ) at constant q
have not been successful. For a current ramp-down experiment wi

429

IAEA-CN-47/A-VII-4

0.5

1.5

FIG. 7. Data from TFTR profiles on an S-a diagram forvarious types ofdischarge. The pressure profile
used is the total pressure, (Pt + PJ/2, calculated by the TRANSP code [9].

NBI, 3 reached 2.65 before the discharge disrupted; this 6 was
not maintained in equilibrium.
There appears to be a hard 6
limit at 2 - 2.2 or an eB » 0.7 limit, as illustrated in Fig.
6. However, the range of the inverse aspect ratio e for the TFTR
data is from 0.28 to 0.33 and it is not clear from this data
whether e is an important parameter in the 3 limit.
The profiles of enhanced confinement discharges have been
examined for violation of the high-n ballooning mode stability
criterion.
Figure 7 shows points on the profile in the S-adiagram [14] (where S = r(dq/dr)/q, a = -2 y o g 2 R(dP T /dr)/B^ 2 ) for
various types of discharges. The pressure profile, (P||+p^)/2 fromthe TRANSP code [9] , which includes calculated beam and thermal
ion pressure profiles and anisotropy, is used for dP^/dr (at,eam <<I ne
0.7 a total^*^ b eam ~driven an<3 bootstrap currents have been
included in the calculation of q.
The profile of maximum 6
hovers close to the curve for the large-aspect-ratio circular
tokamak n + °° ideal ballooning-mode stability limit [14]. Though
the profiles may be at the ballooning limit, these discharges are
not the optimum for n + "> ballooning stability [15] (j (r) is not
optimum). It is well known [16] that FLR, finite-n, toroidicity,and pressure anisotropy effects can all lead to significant
movement of the stability boundary, and it is likely that TFTR isstable to modes with n > 10 under these conditions.
Higher-nmodes (m > 3 and n > 3) have been observed on soft x-ray diodes
viewing the center of the plasma in TFTR at q . " 5 - 7 and athigh injection power. These high-n modes are not observed before
disruptions at high 6 .

430
5.

McGUIRE et al.
Turbulence in TFTR

Langmuir probes in the scrape-off region outside the lastclosed flux surface have revealed broadband turbulent fluctuationsin the frequency range 10 - 200kHz, with fluctuation levels
typically n/n = 0.3 - 0.5 and 9/T = 0.1,
Imaging of D-alphalight at the inner wall also shows large amplitude fluctuations[17] in this frequency range, with typical poloidal wavelengths
X - = 0.03 - 0.07m, but with X> 1.0m, Measurements of themagnetic fluctuations 0.11m outside the plasma edge show that theamplitude at 100kHz increases by a factor of 4 to Eg /E^ = 5 x 10
during neutral beam injection for both L-mode and enhanced
confinement discharges. Since the distance of the coils from theplasma boundary varies for different coils, a radial e-folding
distance can be measured; for most conditions this distance isabout 0.05m. A strong dependence of the turbulent fluctuations on
local edge conditions is found.
For example, a gas puff of
5 Torr* 1/s or a burst of low frequency m/n = 5/1 or 4/1 mode
activity in an ohmically heated discharge can decrease theturbulent fluctuations by a factor of 10. From this result it isclear that turbulent fluctuations measured outside the plasma are
not generally appropriate for direct correlation with global
energy confinement.
The broadband spectra of "n, ? and 1^ in the frequency range
10 - 250kHz have an approximate power-law dependence P(f) « f\where -1 > Y > -2. Short radial decay lengths are found for n and
Eg in the range corresponding to "mode numbers" m > 15, i.e.,
small-scale compared with the plasma size. The amplitude of thebroadband fluctuations of n and Eg increased with NBI both in the
L-mode and enhanced confinement mode.
6.

Conclusions

On TFTR, three different types of sawteeth have been
identified.
The main differences are in the precursor andsuccessor oscillation amplitudes. Possible explanations include
extensions of the Kadomtsev model with two q = 1 surfaces andpartial reconnection.
Turbulence measured in the scrape-off
region shows large amplitude fluctuations in the frequency range
10-250kHz, with fluctuation levels typically "n/n = 0.3-0.5
and?/T e = 0.1.
It is also found that the measured edge
fluctuations of 'n and "Eg can be extremely sensitive to local edge
conditions.
The enhanced confinement discharges on TFTR have values of
3 n at low q j which approach the limit found from ideal kinkstability theory. At higher q 1 it appears that the beta limit
is better described by e6 D I A - 0.7. In addition, dPT/dr in thesedischarges is close to the n + » large aspect ratio ideal
ballooning stability limit. High frequency modes at > 60kHz withm > 3, n > 3 have been observed in the core of discharges with

IAEA-CN-47/A-VII-4

431high B and high neutral beam power. Deterioration of confinementin the enhanced confinement discharges appears to be associatedmore with coherent modes (m/n = 2/1 and 3/2). Major disruptions
at high S can have fast thermal energy decay (< 100 usec); n = 1
modes (with many coupled m-modes) are sometimes observed, but
high-n modes are not detected by the magnetic coils or soft x-ray
diodes during these disruptions.

ACKNOWLEDGEMENTS
The authors are grateful to D.J. Grove and J.R. Thompson for their advice andsupport, and to J. Strachan for his many contributions.
This work was supported by the United States Department of Energy, underContract No. DE-AC02-76-CHO3073. The ORNL participants were also supportedbytheUnitedStatesDepartmentofEnergy,underContract

No. DE-AC05-840R21400 with Martin Marietta Energy Systems, Inc.

REFERENCES[1] PFEIFFER, W., Nucl. Fusion 25 (1985) 673.
[2] McGUIRE, K., et al., in Controlled Fusion and Plasma Physics (Proc. 12th Europ. Conf.
Budapest, 1985), Part 1, European Physical Society (1985) 134.
[3] CAMPBELL, D., et al., Nucl. Fusion 26 (1986) 1085.
[4] KADOMTSEV, B., Fiz. Plazmy 1 (1975) 710.
[5] DENTON, R., et al., Phys. Rev. Lett. 56 (1986) 2477.
[6] PARAIL, V., PEREVERZEV, G., Fiz. Plazmy 6 (1980) 27.
[7] HAWRYLUK, R.J., et al., Paper IAEA-CN-47/A-I-3, these Proceedings, Vol. 1.
[8] GOLDSTON, R.J., et al., Paper IAEA-CN-47/A-II-l, these proceedings, Vol. 1.
[9] HAWRYLUK, R.J., in Physics of Plasmas Close to Thermonuclear Conditions (Proc. Course
Varenna, 1979), Rep. EUR/FU/BRU/XD7476/80, CEC, Brussels (1980) 503.
GOLDSTON, R.J., et al., J. Comput. Phys. 43 (1981) 61.
[10] FURTH, H., Plasma Phys. Controll. Fusion 28 (1986) 1305.
[11] TROYON, F., et al., Plasma Phys. Controll. Fusion 26 (1984) 209.
[12] MANICKAM, J., et al., Nucl. Fusion 24 (1984) 595.
[13] ROBINSON, D.C., et al., in Plasma Physics and Controlled Nuclear Fusion Research 1984
[14]
[15]
[16]
[17]

(Proc. 10th Int. Conf. London, 1984), Vol. 1, IAEA, Vienna (1985) 205.
LORTZ, D., et al., Phys. Lett., A 68 (1978) 49.
CHOE, W., et al., Phys. Fluids 29 (1986) 1766.
SYKES, A., et al., in Controlled Fusion and Plasma Heating (Proc. 1 lth Europ. Conf. Aachen,1983), Pt. 11, European Physical Society (1984) 363.
BISHOP, C , et al., Nucl. Fusion 24 (1984) 643.
TANG, W., et al., Nucl. Fusion 21 (1981) 891.
ZWEBEN, S., et al., J. Nucl. Mater. 145-147 (1987).

IAEA-CN-47/A-VII-5

SAWTEETH AND DISRUPTIONS IN JET
D.J. CAMPBELL, P.A. DUPERREX1, A.W. EDWARDS,
R.D. GILL, C.W. GOWERS, R. GRANETZ, M. HUGON,
P.J. LOMAS, N. LOPES-CARDOZO2, M. MALACARNE,
M.F. NAVE, D.C. ROBINSON3, F.C. SCHULLER,
P. SMEULDERS, J.A. SNIPES3, P.E. STOTT, G. TONETTI1,
B.J. TUBBING2, A. WELLER4, J.A. WESSONJET Joint Undertaking,
Abingdon, Oxfordshire,
United Kingdom

Abstract
SAWTEETH AND DISRUPTIONS IN JET.
The behaviour of magnetohydrodynamic phenomena associated with sawteeth and major disruptions has been studied for a wide range of discharge conditions with Ohmic, neutral beam and ion cyclotron heating. The sawtooth collapse has been investigated in detail and found to be inconsistent withmodels based on resistive reconnection. Disruptions set limits to JET operation at low q and high density, and the detailed evolution of disruptions at these limits has been studied. A hard operational limit
is found at q^ = 2.0 ± 0 . 1 . The highest value of the Murakami parameter achieved in Ohmic plasmas
is nR/B T = 7.2 x 1019 m^T" 1 , but neutral beam heating leads to a substantial increase ( - 5 0 % ) in
the density limit. Finally, during high power neutral beam heating combined with ion cyclotron heating,'monster' sawteeth are observed in which MHD activity of modes with low m and n becomes quiescentfor periods ~ 1 s.

INTRODUCTION
Sawteeth and disruptions are well known phenomena in all
tokamaks and define the limits on the mhd stability of thedischarge. They have been studied extensively in JET and there
are some important new features in the results which will bereported here.
Our studies of these phenomena have used all of the JET
diagnostics, but in particular
we have madedetailed
measurements of the mode structures using:magnetic pick-<up
1

Centre de recherches en physique des plasmas, Ecole polytechnique federate de Lausanne,
Lausanne, Switzerland.
2
FOM Instituut voor Plasmafysica, Nieuwegein, Netherlands.
3
UKAEA, Culham Laboratory, Abingdon, Oxfordshire, UK.
4
Max-Planck-Insthut fur Plasmaphysik, Garching, Fed. Rep. Germany.

433

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CAMPBELL et al.

2 §

(O

IAEA-CN-47/A-VH-5

435

coils and loops outside the plasma boundary; electron cyclotron
emission to measure local electron temperatures; and orthogonalarrays of X-ray diodes combined with tomographic analysis tomeasure the soft X-ray emissivity.
SAWTEETH
Sawteeth are observed in almost all JET discharges. Theirgeneral features which have been discussed in earlier papers[1,2] will only be summarised here.
In _phmically heatedplasmas the modulation of the central electron temperature is
up to 20S& and the period is between 30 and 250ms.
Indischarges with ion cyclotron heating (Fig. 1) or co-injected
neutral beams the modulation is increased up to 50$ and theperiod extended up to 600ms,
whereas with counter injectedbeams both the modulation and the period are reduced.
With
combined ion cyclotron and neutral beam heating we havesometimes seen "monster" sawteeth lasting for more than 1s. We
have attempted to determine the scaling of the period withvarious tokamak parameters, but the only clear trend is that inohmic discharges the period increases with plasma density.
Most JET sawteeth show a compound structure with one or morepartial collapses occuring between the main collapses.
As we have reported elsewhere [1] the main collapse ofsawteeth in JET occurs on a much shorter time scale (typically- 100ys) than would be consistent with resistive reconnection
models.
In ohmic discharges the collapse usually occurs
spontaneously without any precursor oscillations but often we
see successor oscillations. The behaviour is very similar inion cyclotron heated discharges, but with neutral beam heating
we usually see a precursor either as a growing oscillation orsometimes as a saturated mode. When there are precursors, themain collapse appears to be a continuation of the precursor.
This suggests that the precursor and main collapse are the same
instability, but this has not been proved conclusively.
The collapse mechanism appears to be the same in all JET
discharges despite these differences in precursor activity.
Detailed studies [3] with the X-ray and ECE diagnostics showthat the collapse has an m=n=1 structure and involves thedisplacement in minor radius of the hot central core of plasma(Fig. 2a). The displaced core spreads poloidally around themagnetic axis into a crescent shaped region (Fig. 3) and at thesame time there is a rapid outflow of plasma energy.
The
successor oscillations are seen to be the rotation of aresidual asymmetry in the collapsed state.
The soft X-ray
emission profiles after the collapse are sometimes slightly
hollow.
The time scale of the collapse and the detailed

436

CAMPBELL et al.

437

IAEA-CN-47/A-VH-5

m

-I

438

CAMPBELL et al.

displacement and spreading of the hot core are in agreementwith the predictions of the model by Wesson [H] where thesawtooth collapse is an ideal mhd instability.
A magnetic
signal with n=1 is detected simultaneously with the collapse[5].
Partial collapses show a wider range of behaviour in
precursor and successor oscillations and are more difficult toclassify, but there appear to be at least two general types.
One is similar to the main collapse, with a sudden (-lOOps) but
smaller displacement of the hot core (Fig. 2b), and thesubsequent propagation of a heat pulse outside the central
region.
The second type is a much softer relaxation on alonger timescale - 10ms with an m = 1 oscillation which shifts
inwards but does not lead to a central collapse. Both types ofpartial collapse have very little overall effect on the central
temperature.

Pnf=5MW

8.2
Time (s)6.6

FIG. 4. Evolution of central electron temperature Te(0), line integral of electron density \nedt, total
plasma energy from diamagnetic measurements Wdia, and amplitude of total MHD activity during 5 MA
discharge with a monster sawtooth lasting 1.2 s.

439

IAEA-CN-47/A-VII-5

T.(0)A(n=1)

Jl^tJiWlUV}^^,

FIG. 5. After monster sawtooth collapses at time A, large amplitude n = 2 mode persists until end ofneutral beam heating pulse at time B.

MONSTER SAWTEETH
In discharges with combined ion cyclotron and neutral beam
heating some sawteeth have very long duration (up to 1.26s).
These were first seen in relatively low current discharges (I_- 2MA), but similar long duration ("monster") sawteeth have now
been seen in discharges with an internal separatrix (I =• 2MA)
and at high currents (I = 5MA) as shown in figure 1. These
are not simply sawteeth with particularly long periods, as they
are characterised by low levels of coherent mhd activity
suggesting that the m=1 mode has been stabilised.
However
after the sawtooth collapse the plasma remains in an unstablestate characterised by a large m = 3, n°2 mode which oscillates
for several seconds and often persists until the end of theneutral beam pulse (Fig. 5). This seems to prevent the growth

440

CAMPBELL et al.

x

0.6 --

"

Low-q
Ohmic
-

Model Prediction(B T =2.8T , Z e t l =2, a=1.2,•^Pellet

^
0.4 -

®
s

s

Ohmic anc
RF Data —OhmicHelium\0.2 -Neutral BeamHeating//+

n R/BT(x101"m-2.T-1)

FIG. 6. Hugill diagram summarizing the limits of stable operation in JET. Preliminary results of single
pellet injection into Ohmic discharges which did not disrupt are also shown. Low q limit is more consistently plotted in terms of q^ as shown in upper figure.

of a second monster sawtooth In the same pulse, although there
may be several normal sawteeth.
The conditions for monster
sawteeth have not yet been fully explored, but simultaneous ion
cyclotron and neutral beam heating appear to be required. The
ratio of ion cyclotron to neutral beam power does not appear tobe critical but there does appear to be a threshold of total
additional power which increases with plasma current.
A
possible explanation is that the current profile is modified by
the beam driven current which may be enhanced by the ion
cyclotron heating.
LOW q DISCHARGES AND DISRUPTIONS
Reproducible, stable ohmic discharges with I = 3.5MA,
B T = 1 .7T, R = 2.97m, a = 1 .2m, b/a = 1 .35, q^, = 2.2 andq c = ( 5 X i 0 6 abB T /RI )-1.59 have been studied to simulate theIAEA-CN-47/A-VII-5

441

planned operation of JET at I = 7MA and B T = 3.4T. The
sawtooth inversion radius is u.55a.
The global energy
confinement time of O . ^ s is consistent with normal JET
scaling.
These discharges reach the highest value of the
Murakami parameter n R/B T = 7.2X10 19 m~2-Tl obtained for JET
ohmic plasmas (Fig. 6 ) .
When q. is lowered either by raising I , lowering B T , orby reducing a or b/a, a lower limit to stable operation isalways found at q. = 2.0 + 0.1. These low q disruptions are
initiated by a sawtooth collapse and have precursors only on atime scale shorter than the sawtooth period.
The totalradiated power remains less than the ohmic power input andthere is no contraction of the current profile.

DISRUPTIONS AT THE DENSITY LIMIT
Disruptions at the density limit in discharges with q. > 3
are preceeded by signs of deterioration for up to 1s before thefinal energy quench.
The radiated power rises to equal the
total power [6], and although the ohmic power increases due toincreased plasma resistance, it fails to keep pace with theradiation.
The discharge contracts until just before thequench there is virtually no current flowing outside the q=2
surface. Combining these observations with the assumption that
all the energy loss outside q=2 is due to low Z impurity
radiation gives an expression for the density limit in JET. An
earlier calculation [7] has now been extended to include theeffect of additional heating and a more accurate model [8] of
the impurity radiation. The predicted high density limit for
low Z dominated discharges with gas refuelling is"crit " 3.5X10-(Z e f f -1) Rab(q c - 2a/b))
where P is the total power input (watts). This model clearly
breaks down when Z e f f is close to 1 and other coolingmechanisms become important. There is good agreement between
this expression and observed density limits for disruptions
during the current flat top [6], but disruptions during thecurrent rise and decay phases generally occur at lower andhigher densities respectively than the predicted values. The
model also indicates that the weakness of the increase in thedensity limit with ion cyclotron heating is due to the observed
increase in Z e f f with increasing power, whereas the substantial
increase in the density limit with beam heating is slightly
assisted by the reduction of Z e f f by the beam particles.

442

CAMPBELL et al.

-0.5

3.6-

3.7-0.6

1

3.8'

5

0.7

5

3.9"
-0.8
4.0"

¥¥V V
^M-V^^-^V-^^.I^V^)
f?*~ '.

10ms

FIG. 7. Contours of electron temperature (at 100 eV intervals) versus time during magnetic activity
which precedes disruption. Lower traces show (a) electron temperature at R = 3.95 m (r/a = ~0.8),
and amplitudes (b)ofm = 2, n = 1 oscillation (c) of locked mode.

DETAILED STUDIES OF DISRUPTIONS
Detailed studies are being made of the sequence of events
which precede a density limit disruption [9].
In a typicalcase (I = 3MA, B T = 3-^T, q^=5.7) the significant magnetic
activity, shown in Figure 7, starts about 700ms before theenergy quench with a mode which is predominantly m=2, n=1.
This activity grows for about 30ms.
The mode then stops
rotating and locks in a fixed toroidal position when Bg at thewall is 10~ 3 T.
During the locked phase the mode continues togrow in amplitude. ECE measurements show that the temperature
profile shrinks before the start of the mhd activity. After themode locks the temperature profile flattens slightly at a minor

IAEA-CN-47/A-VII-5

443

radius of 0.95 m which is close to the calculated position of
the q=2 surface.
This is consistent with the growth of a
non-rotating magnetic island (the ECE diagnostic views close to
the calculated position of the '0' point) and there may besimilar evidence from the soft X-ray emission, but the analysis
is at a preliminary stage and there could be otherinterpretations. The spatial resolution of the ECE diagnostic(at present *)0mm in the radial direction) is not quite
sufficient to determine whether the '0' point is colder than
the surrounding plasma as predicted by the thermal instability
model of disruptions [10].
The locked mode persists for several hundred ms until the
final energy quench. This has been studied in detail for thelast few ms before the energy quench using the soft X-ray diodearrays. A contour plot (Fig. 8) of the X-ray emission from thevertical camera shows that 2ms before the quench the central
core of the discharge is eroded. The emission following thefinal termination is probably due to contact of the plasma withthe walls and limiter.
Tomographic analysis of the X-ray
emission shows the growth of a large m=1 asymmetry in theplasma centre.
This structure appears to be locked in a
specific toroidal phase for most disruptions which suggeststhat it is n=1 and coupled to the locked m=2 mode observed
magnetically.

MAGNETIC FLUCTUATIONS
The magnetic pick-up coils detect incoherent fluctuations
as well as coherent mhd activity. These incoherent fluctuations have a broadband spectrum over the whole measured
frequency range from 5 to 60kHz.
During typical conditions the frequency dependence of thespectrum is f~' • 6 ± 0 . 2 and the total normalised fluctuation
amplitude is B Q /B Q -10~ 5 . The magnetic fluctuations have ashort growth and decay time. A strong correlation is observedin the direction parallel to B. The normalised amplitude isseen to increase with additional heating power which may
indicate a possible link with reduced confinement [11].
CONCLUSIONS
Detailed studies of sawteeth, disruptions and magnetic
fluctuations are being made in JET with the aim of improving
our understanding of these phenomena. The detailed mechanisms
involved in the collapse of sawteeth have been investigated andfound to be in agreement with an ideal mhd model of the444

CAMPBELL et al.

8
o

7

o
II

-o

v
(ui))M6|3H# jojoejeaejiIAEA-CN-47/A-VII-5445

collapse. A new regime which appears to be stable to low m andn modes for more than 1s has been observed with combined ion
cyclotron and neutral beam heating. The operational limits tostable operation at low q and high density have been explored
and the detailed mechanisms leading to disruptions have beenstudied. This work is proceeding with the objective of fully
understanding the nature of disruptions and attempting tostabilise them by various means.

ACKNOWLEDGEMENTS
It is a pleasure to acknowledge contributions to this workand many helpful discussions with our colleagues in the JET
project.

REFERENCES[1] CAMPBELL, D.J., et al., Nucl. Fusion 26 (1986) 1085.
[2] GILL, R.D., et al., in Controlled Fusion and Plasma Heating (Proc. 13th Europ. Conf.
Schliersee, 1986), Vol. 1, European Physical Society (1986) 21.
[3] EDWARDS, A.W., et al., Phys. Rev. Lett. 57 (1986) 210.
[4] WESSON, L, Plasma Phys. Controll. Fusion 28 (1986) 243.
[5] DUPERREX, P.A., et al., in Controlled Fusion and Plasma Physics (Proc. 12th Europ. Conf.
Budapest, 1985), Vol. 1, European Physical Society (1985) 126.
[6] WESSON, J., et al., ibid., p. 147.
[7] SCHULLER, F.C., et al., ibid., p. 151.
[8] BEHRINGER, K., et al., Paper IAEA-CN-47/A-IV-1, these Proceedings, Vol. 1.
[9] HENDER, T.C., ROBINSON, D . c , SNIPES, J.A., paper IAEA-CN-47/A-V-3, these
Proceedings, Vol. 1.
[10] REBUT, P.H., HUGON, M., in Plasma Physics and Controlled Nuclear Fusion Research 1984
(Proc. 10th Int. Conf. London, 1984), Vol. 2, IAEA, Vienna (1985) 197.
[11] MALACARNE, M., DUPERREX, P.A., JET Preprint-P(86)31, JET Joint Undertaking,
Abingdon (1986).

Session F
PLASMA HEATING AND CURRENT DRIVE
Chairmen of Sessions
Session F-I

O.S. PAVLICHENKO

USSR

Session F-II

R.R. PARKER

USA

Session F-III

F.G. PREVOT

FranceSession F-IV (Posters)

IAEA-CN-47/F-M

RADIOFREQUENCY HEATING ON JETJET TEAM1(Presented by J. Jacquinot)
H. Altmann, R.J. Anderson, J. Arbez, W. Bailey, D.V. Bartlett, B. Beaumont',
G. Beaumont, K. Behringer, P. Bertoldi, E. Bertolini, C.H. Best, V. Bhatnagar2,
R.J. Bickerton, A. Boileau3, F. Bombarda4, T. Bonicelli, S. Booth, A. Boschi,
G. Bosia, M. Botman, H. Brelen, H. Brinkschulte, M.L. Browne, M. Brusati,
T. Budd, M. Bures, P. Butcher, H. Buttgereit, D. Cacaut, C. Caldwell-Nichols,
J.D. Callen5, D.J. Campbell, J. Carwardine, G. Celentano, C D . Challis,
A. Cheetham, J.P. Christiansen, C. Christodoulopoulos, P. Chuilon, R. Claesen,
J.P. Coad,
S.A. Cohen6,
M. Cooke, J.G. Cordey, W. Core, S. Corti,
A.E. Costley, G.A. Cottrell, M. Cox 7 , C. David, J. Dean, L. de Kock,
E. Deksnis, G.B. Denne, G. Deschamps, K.J. Dietz, J. Dobbing, S.E. Dorling,
D.F. Diichs, G. Duesing, P.A. Duperrex8, H. Duquenoy, A.W. Edwards,
J. Ehrenberg9,
W. Engelhardt,
S.K. Erents7,
F. Erhorn,
B.T. Eriksson,
M. Evrard2, H. Falter, N. Foden, M.J. Forrest7, C. Froger, K. Fullard,
G. Fussmann9,
M. Gadeberg10, A. Galetsas, A. Gallacher,
D. Gambier1,
R. Giannella4, A. Giannelli4, A. Gibson, R.D. Gill, A. Goede, A. Gondhalekar,
C. Gordon, C. Gormezano, N.A. Gottardi, C.W. Gowers, R. Granetz, B. Green,
S. Gregoli, F.S. Griph, R. Haange, J.H. Hamnen", C.J. Hancock, P.J. Harbour,
N.C. Hawkes7, P. Haynes7, T. Hellsten, J.L. Hemmerich, R. Hemsworth,
F. Hendriks, R.F. Herzog, L. Horton12, J. How, M. Huart, A. Hubbard13,
J. Hugill7, M. Hugon, M. Huguet, B. Ingram, H. Jackel9, J. Jacquinot,
Z. Jankowicz14, O.N. Jarvis, E.M. Jones, P. Jones, T.T.C. Jones, C. Jupeii15,
E. Kallne, J. Kallne, O. Kaneko16, A. Kaye, B.E. Keen, M. Keilhacker,
G. Kinahan, S. Kissel, A. Konstantellos, M. Kovanen", U. Kiiphnapfel9,
P. Kupschus, P. Lallia, J.R. Last, L. Lauro-Taroni, K.D. Lawson7, E. Lazzaro,
R.C. Lobel, P.J. Lomas, N. Lopes-Cardozo18, M. Lorenz-Gottardi, C. Jj>wry13,
G. Magyar,
D. Maisonnier,
M. Malacarne,
V. Marchese,
P. Massmann,
G.M. McCracken7,
P. McCullen,
M.J. Mead,
P. Meriguet,
V. Merlo,
V. Mertens9, S. Mills, P. Millward, A. Moissonnier, P.L. Mondino, D. Moreau',
P.D. Morgan,
E.R. Muller9,
D. Murmann9,
G. Murphy,
M.F. Nave,L. Nickesson, P. Nielsen, P. Noll, S. Nowak, W. Obert, M. Olsson, J. O'Rourke,
M.G. Pacco, J. Paillere, S. Papastergiou, J. Partridge7, D. Pasini12, N.J. Peacock7,
M. Pescatore, J. Plancoulaine, J.-P. Poffe\ R. Prentice, T. Raimondi, J. Ramette1,
C. Raymond,
P.H. Rebut,
J. Removille,
W. Riediker,
R. Roberts,
D.C. Robinson7, A. Rolfe, R.T. Ross, G. Rupprecht9, R. Rushton, H.C. Sack,
G. Sadler, J. Saffert, N. Salmon13, F. Sand, A. Santagiustina, R. Saunders,M. Schmid, F.C. Schiiller, K. Selin, R. Shaw, A. Sibley, D. Sigournay,
R. Sillen18, F. Simonet1, R. Simonini, P. Smeulders, J.A. Snipes7, K. Sonnenberg,
L. Sonnerup, M.F. Stamp, C.A. Steed, D. Stork, P.E. Stott, T.E. Stringer,
D.R. Summers, H.P. Summers19, A.J. Tagle, G. Tallents20, A. Tanga, A. Taroni,
A. Terrington, A. Tesini, P.R. Thomas, E. Thompson, K. Thomsen10, F. Tibone,
R. Tivey, T.N. Todd7, P. Trevalion, M. Tschudin, B.J. Tubbing18, P. Twynam,
E. Usselmann, H. Van der Beken, J.E. Van Montfoort, M. von Hellermann,
J. von Seggern21,
T. Wade, C. Walker,
B.A. Wallander,
M. Walravens,

449

450

JET TEAM
K. Walter, M.L. Watkins, M. Watson, D. Webberley, A. Welter', J.A. Wesson,
J. Wilks, T. Winkel, C. Woodward, M. Wykes, D. Young, L. Zannelli,
J.W. Zwart

JET Joint Undertaking,
Abingdon, Oxfordshire,
United Kingdom

Abstract
RADIOFREQUENCY HEATING ON JET.
High power ICRF, P < 7.2 MW, has been coupled to the JET plasma in a variety of conditions:outboard limiters, inner wall, combined operation with neutral beams and pellet injection. Long pulse
modulation experiments (10 s) yield a precise deposition which is extremely localized, in agreement withexpectations. The best heating efficiency is obtained when the minority resonance crosses the magnetic
axis. With off-axis heating, the ion temperature profile broadens more than the electron temperature
profile. A new operating regime called 'monster sawtooth' has been found during combined RF and
NBI heating. The MHD quiet regime can last 1.26 s and the central electron temperature rises to
7.4 keV with a D-D reaction rate of 4 x 1015 s"1. The temperature profile steepens over the entire profile and r E is enhanced by a factor of 2.5 for a combined power of 10 MW. RF heating of plasma
fuelled with a single mildly penetrating pellet exhibits the normal L-mode behaviour. In combination
with NBI, the input power reaches 16 MW and the plasma stored energy reaches 5.8 MJ at the highest
plasma current of 5 MA. Recent experiments in single-null X-point configuration with combined heating
have shown the transition to an enhanced confinement regime (H-mode).

1

CEA, Centre d'&udes nucle'aires de Fontenay-aux-Roses, France.
Plasma Physics Laboratory, Association "Euratom-Belgian State" ERM/KMS, Brussels, Belgium.
3
Institut national de la recherche scientifique, Ste-Foy, Quebec, Canada.
4
ENEA, Centro Ricerche Energia, Frascati, Italy.
5
Department of Physics, University of Wisconsin, Madison, WI, USA.
6
Princeton Plasma Physics Laboratory, Princeton University, NJ, USA.
7
UKAEA, Culham Laboratory, Abingdon, Oxfordshire, UK.
8
Centre de recherches en physique des plasmas, Ecole poly technique f&le'rale de Lausanne,
Switzerland.
9
Max-Planck-Institut fur Plasmaphysik, Garching, Fed. Rep. Germany.
10
Ris0 National Laboratory, Roskilde, Denmark.
11
Swedish Energy Research Commission, Stockholm, Sweden.
12
Natural Sciences and Engineering Research Council, Ottawa, Ontario, Canada.
13
Imperial College of Science and Technology, University of London, UK.
14
RCC CYFRONET IAE, Otwock-Swierk, Poland.
15
University of Lund, Sweden.
16
Department of Physics, University of Nagoya, Japan.
17
Department of Physics, Lappeenranta University of Technology, Lappeenranta, Finland.
18
FOM Instituut voor Plasmafysica, Rijnhuizen, Nieuwegein, Netherlands.
19
University of Strathclyde, Glasgow, UK.
20
Research School of Physical Sciences, Australian National University, Canberra, Australia.
21
Kernforschungsanlage Julich GmbH, Fed. Rep. Germany.
2

IAEA-CN-47/F-I-1

451

1. INTRODUCTION
Initial ICRF heating experiments on JET have recently been reported [1,2]. Performed at a
maximum power level of 5MW, these experiments revealed a degradation in energy confinement with increased power, following a behaviour previously observed in Tokamaks with acold plasma edge. It appeared difficult to raise the electron temperature outside the q = 1 surface above the ohmic heating value.
The heating sources have now been increased up to 7.2 MW using three ICRF antennagenerator units [3] and to 10 MW with NBI [4]. Combined ICRF and NBI heating to a maximum power of 16 MW has successfully been achieved, and a wider variety of operating regimes have been investigated. These results are summarised below. Experiments performed
with ICRF alone are addressed in the first three sections, in which emphasis is placed on power modulation to deduce the deposition profile, on off-axis heating and on the effect of antenna phasing. The last sections described specific aspects of combined ICRF, NBI and Pellet
operation. In particular, injected beam ions appear to be accelerated when the RF is appliedand, in some conditions, the internal sawteeth are stabilized and a high central electron temperature is achieved ('Monster' mode).

2. MODULATION EXPERIMENT AND POWER DEPOSITION PROFILE
The power deposition profile can be inferred from the change of slope of the temperature evolution during the sharp rise of the applied heating source. A variant is to observe the slopeof internal sawteeth [5]. The signal-to-noise ratio is generally mediocre and the method could
only be applied to heating on axis at high power density. Experiments with long and amplitude modulated heating pulses offer the opportunity to use noise rejection techniques, suchas the box-car or the fast Fourier transform. The frequency of modulation VM can be scanned
across the characteristic energy exchange and transport frequencies to reveal the dominantenergy channels. In these experiments, VM was between 5-62Hz, the heating pulse lasted during 10s and the peak-to-peak modulation was about 1MW. The local perturbation of the electron temperature was measured with a 12-channeI polychromator.
When the modulation frequency is high enough, the response of the electron species is adiabatic and it is possible to deduce directly the power density Pe transferred to the electrons,
as shown in Fig. la,b. In both cases, the peak response is close to the location of the cyclotron
resonance and the width of the deposition (FWHM) is ~ 60cm for on-axis heating and - 20cm
for off-axis heating.
The experimental result is compared with theoretical descriptions using either the ray tracing technique [6] or full wave solutions [7] .Both models take into account the D-shape crosssection and the actual plasma parameters. These both predict that most of the coupled poweris first transferred to the minority species, as observed, and is subsequently redistributed bycollisions according to Table I. In the ray tracing model, part of the power (~ 20%) is not absorbed and is assumed lost in the wall or the edge. A very small part ( = 5%) goes directly tothe electrons and ions. In the full wave model, the pressure anisotropy of the minority is taken
into account. A significant fraction of the power is directly coupled to the deuterons (10 to 20%).
Note that electron heating prevails with heating on-axis, but, as the power density decreases
with off-axis heatingjon heating increases. At high power, electron heating tends to dominate
in all cases. Both models give a good representation of the observed deposition profile (Fig.la.b).
The agreement is excellent for off-axis heating. Central heating is somewhat broader than
predicted.
The observed temperature modulation decreases rapidly with vu(Te<x VM '2). Such a law
can only be obtained theoretically if the RF power first transits via the minority before reaching the electron species (direct electron heating would imply fe « VM' ')• The measurement

452

JET TEAM
(a)
C

#9255

e

[kW/m3]A

Pe/3M 50"

[kW/m3]+ Modulation
sxperiment100

lay tracing» -ull wave

80

0

40
*

30
20

- 60

•

A

40

r

A

10--

n

«
¥ s
X

3.1

4

3.4

3.3

Z.2

20

f
D

R(m)

b)

"kW/m3]
10-{-Modulation
experimentO

iD[kW/m3]
-30
-

'

a

o Ray tracing
Full wave

A

o

D

-Jl~'

5-

i

n

'

'

8 1

I

•

3.6

3.7

- 15

""

3.8

3.9
R(m)

FIG. 1. Net power density Pe transferred to electrons, deduced from the adiabatic modulation of Te
during long pulse ICRF modulation experiments: 1.3 MW peak modulation, 12.5 Hi, H minority.
Comparison with theory: (a) on-axis heating B^ = 3.4 T, v = 47.7 MHz. (b) off-axis heatingB^ = 2.9 T, v = 37 MHz. Heating via the minority is demonstrated by comparison with theory.

of fe- KVM) has been used to determine the power coupled to the minority. Results are given
in Table I.
3. OFF-AXIS HEATINGThe importance of power localisation has been studied in two series of experiments. In thefirst [2,8], the minority resonance position is scanned by ramping down the toroidal magnetic field during the RF pulse. In the second, the same effect is obtained by varying the wave
frequency in the same plasma conditions. Moreover, the experiment was carried out after a

453

IAEA-CN-47/F-M

TABLE I. TRANSFER OF POWER
Comparison of two models with experiment
Power input (Pc)
1.3MW

Directly to
Minority H(Pm/Pe)To electrons(Pe/Pc)To deuterons(Prf/Pc)

On-axis
Ray tracing
Off-axis

0.75

0.42

0.39

0.71

0.15

0.60

On-axisFull wave
Off-axis

0.80

0.34

0.66

0.90

0.25

0.75

On-axisObserved
Off-axis

0.6 ± 0.12

>0.31 ± 0.06

0.7 ± 0.15

>0.15 ± 0.03

notmeasuredheavy carbonisation to avoid the radiated power increase when the resonance is far off-axis
and with long pulses (~ 8s) to avoid transient effects. Both series give similar results, for example:• Large amplitude sawteeth on both Te and 7] were observed only when the ion-ion hybrid
resonance was inside the inversion radius [2, 8]. The sawteeth do not disappear with far
off-axis heating, but become small and short.
• The largest increase of central temperature and stored energy was achieved with central
heating (Fig. 2). The increase of Te was only weakly dependent on the resonance location[2].
The deuteron temperature profile (from NPA) appears to broaden with heating off-axis (Fig.
3b). On the contrary, the electron temperature profile is only slightly broadened (Fig.3a) outside the q = 1 volume.
Transport code simulations of off-axis heating cases indicate that the energy transport coefficient, xe, re-adjusts itself to try to conserve a similar Te profile, by lowering \e inside the heating region and by further increasing it outside the heating zone.
4. ANTENNA PHASING AND EDGE EFFECTS
Among the three antennas, two are of the dipole-quadrupole type [1] and internal phasingof the two coupling elements allows the launching of a kn spectrum which is peaked either
for ArM =0 or for kn =7m"'. In addition, these two antennas, located on either side of a large
horizontal port, can be phased relative to each other. When the dipoles are out of phase, twopeaks at kn =2m -1 and 7 m"1 are emitted; 90° phase difference produces a small portion of
travelling wave (directivity « 5%). All these phasing conditions have been compared, using
a target plasma which maximizes the impurity effects with ICRF (low density = <ne> =1.1 x 1019 m~3, no carbonisation for several months). The following observations were made:•Phasing between the two adjacent antennas, each being in the dipole mode (kn =0) did
not produce any measurable difference on wave coupling, edge plasma or plasma heating.

454

JET TEAMAW
D (H)B,= 2.3 & 2.5T
Ip = 2.5 MA

APTOT

Is)

' after heavy
carbonisation

0.3

Q2

0.1

2.0

25
Rres (m)

3.0

|

35

Plasma centre

40
Plasma edge

FIG. 2. Incremental confinement time versus position of the H resonance. The frequency was varied
between 25 and 48 MHz. Data points following a heavy carbonisation are identified.

•

Phasing internally in the quadrupole mode produced not only reduced coupling resistance[1 ] but also a marked reduction of specific ICRF effect on the edge plasma. In particular,
the flattening of scrape-off-layer density and temperature profiles observed with dipoles
disappears with the quadrupole, and the increase of metallic ions originating from theactivated screen (Cr for one antenna, Ni for the other [9]) is about halved. The density
evolution was also different in the two cases: the average density increased more with thequadrupole. Despite this higher density, the radiated power Pmd was significantly lower
and the carbon released from the wall was typically x 2-3 smaller. The heating efficiency
defined by the incremental confinement time rinc = (AW/AP) was higher in the quadrupole phasing by typically 25 % for similar plasma conditions. Table II summarizes thiscomparison between the two phasing conditions, in which lNi, I o are the intensities of
Ni XXVI and Cr XXII lines and <t> is total neutral flux.

It is clear from these experiments that the edge modification induced with dipole phasing
is considerably reduced with the quadrupole. The metallic impurity emitted by the activatedscreen and the insensitiveness of phasing between adjacent antennas suggest that the RF field
pattern immediately in front of each antenna screen plays a key role.
5. HIGH POWER EXPERIMENTS : OPERATING REGIMESWith ICRF alone, it is difficult to couple large amounts of power when the plasma current
is low (e.g. at Jp=1 MA). The plasma disrupts at only 1MW, as the density rise during ICRF
approaches the density limit. At Ip> 2MA, no power limit was found. Combined ICRF andNBI operation removed the power limitations at 1 MA.
Experiments with long pulses (5MW, 8s) have shown that quasi-stationary conditions arereached after about 2s. No serious accumulation of high-Z impurity was observed and theplasma central temperatures remained constant after the initial rise.
Neutral beam injection increased the antenna coupling resistance. This increase can be highly
beneficial; for instance, when the plasma is in contact with the inner wall and 5cm away fromthe antenna, the coupling resistance more than doubled. The technical difficulty of copingwith large changes of coupling resistance meant that it can be difficult to maintain constant

455

IAEA-CN-47/F-I1

TABLE II. COMPARISON OF DIPOLE AND QUADRUPOLE PHASING
1 MW coupled power and identical starting plasma conditions
Me,

A/M

<ne>
(a.u)

<ne>

APnd
(MW)

AZ

<//

A(j>
(1021 s" 1 )Tine(10"m-2)

A nedl(s)(a.u)

Dipole

1.8

3.1

0.74

0.34

0.49

0.05

0.17

Quadrupole

1.2

1.5

0.57

0.14

0.79

0.58

0.23

ao

a2

34

3.6

as

4.0 4.2
R[m]RF ofl axis[klv] 3.0

2&

30

32

3.4

3.6

3.8

4.0

4.2

R(m)

FIG. 3. Electron and ion temperature profiles during the ohmic phase (before RF) and during on-axis
or off-axis RF heating. The main effect of off-axis heating is a broadening of the 7} profile,on-axis: B^ = 2.3 T, v = 37 MHz, Pe = 3 MW, lp = 2.5 MA, q(a) = 3.25
off-axis: B^ = 2.5 T, v = 31.8 MHz, Pe = 5 MW, Ip = 2.5 MA, q(a) = 3.55.

456

JET TEAMf(E)'

1 #8909'

NBI (D) (o)10"Deuterium

ICRF* NBI (D) (•)

HydrogenICRF* NBI ( D ) M

\
•

\

\

1

10
\
\ / \
\0
\

io 1 6

\

F
1s

10

15

20

,
40

\
\
\

.1 80
60

\

100 120
Energy [keV]

FIG. 4. Deuterium and hydrogen energy spectra before and during a 'monster sawtooth'(PNBl = 2.5 MW, 65 keV; P^ = 7MW). Ip = 2 MA, B^ = 2.13 T, v = 31.8 MHz. Acceleration ofbeam injected ions up to 120 keV is observed.

RF power when the beams are switched on during a RF pulse. Injecting RF power during a
NBI pulse is normally straightforward. The variation of coupling resistance with the crashof the large sawteeth is often a factor limiting the power during the pulse.
6. COMBINED RF AND PELLET INJECTION
Combining ICRF with pellet injection did not raise any special technical problems. 6MW RF
could be applied at the time of injection of a mildly penetrating pellet (3.6 mm, lkm-s"1). The
power was ramped up over 0.5s. The change of coupling resistance due to the pellet was hardly noticeable. The ratio of radiated power to input power increased only slightly by 10-20%
despite a doubling of the average density. The pellet increased the density from<ne> =1.9xlO 19 m~3 to 3.7xlO 19 m"3. Without RF, the density decayed typically on a Is
time scale. With RF, the density was sustained at its maximum level and even slightly increased
for the duration of the RF pulse (2s). The increase of 7X0) and 7;(0) with PRF/<ne> wascomparable with the largest increments obtained in the similar discharges without the pellet.
The energy confinement during RF was not modified by pellet injection. The incremental confinement time with the pellet becomes Tmc = 0.2 s. Large sawteeth appear early in the reheat
phase. The pellet penetration was not changed by the presence of ICRF.
7. COMBINED HEATING AND BEAM ACCELERATION
Generally, ICRF plasma heating on the outboard limiters behaves according to the L-mode
scaling law [1,2]. Recent combined NBI and ICRF heating up to 16MW has been performed.
The energy confinement scaling is reported in [10]. Again L-mode behaviour is found, although
inner wall operation and high -q operation perform somewhat better than the 'Goldston Lmode scaling'. Both H and He 3 minority scenarios have been tested, but the stored energy
increase is not appreciably different in the two cases.
In combined RF and NBI heating, the H minority scenario has a distinctive feature: power
can be coupled to the energetic D beam ions at the harmonic cyclotron resonance. It is calculated [11] that the NBI current drive will be enhanced by ICRF. The beam acceleration alsoenhances the fusion reactivity.

457

IAEA-CN-47/F-M
04

x
D( 3 He) I P = 3MA- D (H) Ip = 3MA•D(H) JI.P'2,3MA
I inner wall _XD (H) 'M'discharges•TOT[io9 s'w"1;

Q2

0.1

02

0.4

0.6

0.8

1.0

FIG. 5. Fusion D-D reaction rate normalized to the total input power versus the ratio of RF to total
additional power. The beam plasma reaction is maintained proportional to the additional power,
independently of its origin, provided P^iP^j+ P^) ^ 0.5. Monster sawteeth give a larger
fusion yield.

Figure 4 demonstrates beam acceleration by ICRF. The slope of the energy distribution function above the injection energy increases from 5keV to 16keV. Simulation by a Fokker-Planck
calculation [12] requires that a substantial fraction of the incident RF power is coupled to thebeam ions. However, note that the high energy tail of the proton still exceeds the D tail (Fig.
4)andmostof coupled power is still transferred to the protons. The fraction of power directlytransferred to the beam ions is estimated [13] to be about 10%. It is expected to increase nonlinearly with RF power.
The fusion reactivity in these D ->D injection experiments is dominated by beam-plasma
reactions. RF heating increases the reactivity roughly in proportion to the total heating power (cf. Fig. 5) provided the RF power does not exceed the NBI power. Comparison between
the He3 and H minority shows that a higher reactivity is obtained in the H minority experiments {nH/nD~2x\0~2).
8. MONSTER SAWTOOTH

Unusually long sawteeth and high fusion reactivity have been observed in the so-called 'Monster sawtooth' regime in the following conditions:— IP=2 - 2.2MA, 5^=2.1 - 2.2T, Plasmas lying on the inner wall. H minority, RF alone
or combined with beams— IP=5MA, 5*=3.4T, He3 minority, combined heating— 5MW <(PRF + PNBi) < 15MW
— 2xl0 I9 nT 3 <ne < 3.5xl019m"3
Fig. 6 illustrates such a regime when the RF is applied on top of NBI injection. The internal
relaxations stop and the plasma is remarkably free from MHD activity for - Is. 7;(0) morethan doubles and a peak temperature of 7.4 keV has been obtained with 10MW of additional
power. The Te gradient steepens in the entire profile. The inversion radius appears to expand
up to mid-radius. In many cases, the stored energy does not saturate before the crash, even458

JET TEAM

[keV]

#8909
70

Te

/

\

6.0
5.0 -/
t

/

/

/
/

\

/

\

i

y:

<

4.0
f
3

3.0

5

\

5

\

\

^ ^ N B I
^"~
X
N

t

•RF

7.

I7
Ohmic^

2.0
1.0 i

IS

.

1

i

3.0

3.2

3.4

.

1

1

3.6

3B

4.0

4.2
R[m]

FIG. 6. Te profile before heating (ohmic) during PNBI (2.5 MW) and during a 'monster sawtooth'(NBI = 2.5 MW, PRF = 7 MW). In each case, the curves represent the top and bottom of a sawtooth.
Note the increase of the inversion radius.

Time[s]FIG. 7. Evolution of the stored energy during the longest 'monster sawtooth' (1.26 s). Although Te
saturates, the density and the stored energy keep increasing. PNBl = 2.6 MW (80 keV), Pgp = 5.5 MW,
Ip = 2.2 MA, B^ = 2.18 T.

after 1.26s (Fig. 7). The incremental confinement time T,nc = AW/AP^Fis somewhat improved
over the usual conditions. Typically, T,nc=250ms at the end of the 'Monster' instead of 200ms
in normally sawteething discharges. The pressure anisotropy is mild: f}pi = 1.13 Ppn =0.5. The
crash of the monster releases 30% of the stored energy. Thereafter, the MHD activity is high(mostly n = 2 modes) and the confinement appears degraded. In these experiments, the poloidal beta reached 0.5 and the ion temperature was between 5 and 7keV.

IAEA-CN-47/F-I-1

459

9. TRANSITION TO AN 'H-MODE1
Preliminary experiments with combined heating in a single null X-point configuration exhibitedthe change of edge and bulk plasma parameters in the Asdex H-mode behaviour. The transition occurred wiht 7MW of combined heating in the Hydrogen minority regime. After thetransition, the density rose, the temperature remaining constant, and the confinement timeincreased from 0.35s to 0.6s.

10. CONCLUSIONS
After a year of operation on JET, ICRF heating has reached record levels of RF power andelectron temperatures in a Tokamak. Eight water cooled antennas will be installed in 1987 andconnected to a 32MW power source. Recent results give confidence that JET will demonstrate
a fusion output comparable to the heating power input.

REFERENCES[1] JACQUINOT, J., et al., Plasma Phys. Contrail. Fusion 28 (1986) 1.
[2] LALLIA, P.P., et al., Plasma Phys. Controll. Fusion 28 (1986) 1211.
[3] KAYE, A.S., in Fusion Engineering (Proc. 11th Symp. Austin, TX, 1985), Vol. 2, IEEE, NewYork (1986) 1204.
[4] DUESING, G., et al., Plasma Phys. Controll. Fusion 28 (1986) 1429.
[5] GAMBIER, D., et al., in Controlled Fusion and Plasma Physics (Proc. 12th Europ. Conf.
Budapest, 1985), Vol. 9F, Part n , European Physical Society. (1985) 152.
[6] BHATNAGAR, V.P., KOCH, R., GEILFUS, P., KIRKPATRICK, R., WEYNANTS, R.R.,
Nucl. Fusion 24 (1984) 955.
[7] HELLSTEN, T., APPERT, K., in Controlled Fusion and Plasma Heating (Proc. 13th Europ.
Conf. Schliersee, 1986), Vol. IOC, Part II, European Physical Society (1986) 129.
[8] SAND, F., et al., ibid., p. 197.
[9] BEHRINGER, K., et al., Paper IAEA-CN-47/A-IV-l, these Proceedings, Vol. 1.
[10] CORDEY, J.G., et al., Paper IAEA-CN-47/A-1I-3, these Proceedings, Vol. 1.
[11] HAMNEN, H., et al., in Controlled Fusion and Plasma Heating (Proc. 13th Europ. Conf.
Schliersee, 1986), Vol. 10C, Part II, European Physical Society (1986) 425.
[12] CORE, W., et al., ibid., p. 73.
[13] COTTRELL, G., private communication, 1986.

DISCUSSION
R.J. GOLDSTON: Do you have any idea why the duration of the 'monster sawtooth' before the crash is so long?
J. JACQUINOT: We first thought this might be due to NBI current drive
enhanced by the local increase in Te produced by ICRF. The recent observation that
good 'monster sawteeth' with similar input powers are also obtained with RF alone
weakens this model very much and leaves us without a clear physical picture to offer
as an explanation.

460

JET TEAM

K. ITOH: During the 'monster sawtooth', the time behaviour of the electron
temperature indicates that the electron energy confinement is not improved, while thetotal stored energy increases continuously. Do you observe any increase in ion energy
which is due to the appearance of the 'monster sawtooth'? And do you conclude that
ion energy confinement is improved by 'monster sawteeth'?
J. JACQUINOT: The improvement of confinement during a 'monster sawtooth' seems to be related with the continuous increase in electron density and in thesize of the q = 1 surface. The additional stored energy is mainly in the electrons.

IAEA-CN-47/F-I-2

ION CYCLOTRON HEATING
AND LOWER HYBRID EXPERIMENTS
ON ASDEX
K. STEINMETZ,
F.X. SOLDNER,
D. ECKHARTT,
G. JANESCHITZ,
F. LEUTERER,
F. MAST,
K. McCORMICK,
J.-M. NOTERDAEME,
C. SETZENSACK, F. WAGNER, F. WESNER, I. BAUMLER, G. BECKER,
H.S. BOSCH, M. BRAMBILLA, F. BRAUN, H. BROCKEN, H. DERFLER,
A. EBERHAGEN,
H.-U. FAHRBACH,
G. FUSSMANN,
R. FRITSCH,
O. GEHRE, J. GERNHARDT, G. von GIERKE, E. GLOCK, O. GRUBER,
G. HAAS,
J. HOFMANN,
F. HOFMEISTER,
K. HUBNER1,
A. IZVOZCHIKOV2,
F. KARGER,
M. KEILHACKER3,
O. KLUBER,
M. KORNHERR,
K. LACKNER,
M. LENOCI,G. LISITANO,
E. VAN MARK, H.M. MAYER, D. MEISEL, V. MERTENS, E.R. MULLER,
E. MUNICH,
H. MURMANN,
H. NIEDERMEYER,
Z.A. PIETRZYK4,
W. POSCHENRIEDER, S. PURI, H. RAPP, H. RIEDLER, H. ROHR, J. ROTH,
F. RYTER5,
K.H. SCHMITTER,
F. SCHNEIDER,
G. SILLER,
P. SMEULDERS3, E. SPETH, K.-H. STEUER, T. VIEN, O. VOLLMER,
H. WEDLER, D. ZASCHE, M. ZOUHAR

Euratom-IPP Association,
Max-Planck-Institut fur Plasmaphysik,
Garching, Federal Republic of Germany
Abstract
ION CYCLOTRON RESONANCE HEATING AND LOWER HYBRID EXPERIMENTS ON
ASDEX.
A summary is given of recent experimental investigations with ICRH and LH on the divertor
tokamak ASDEX. Hydrogen second harmonic heating (20 CH at 67 MHz), hydrogen minority heating(D(H) at 33.5 MHz), and ICRH in combination with neutral beam injection (NI) are compared in detailat different parameters (PIC < 2.6 MW, PNI < 4.4 MW, IP = 250 to 450 kA, ne = 2 to 9 x 1013
cm" 3 ). Heating and current drive with waves have been studied in Ohmic target plasmas and in combination with NI heating up to a total heating power of 4 MW (PNI < 3 MW plus P LH < 1 MW).

'
1
3
4
5

University of Heidelberg, Heidelberg, Fed. Rep. Germany.
Academy of Sciences, Leningrad, USSR.
JET Joint Undertaking, Abingdon, Oxfordshire, UK.
University of Washington, Seattle, WA, USA.
CEA, Centre d'etudes nucl6aires de Grenoble, France. Present address: Centre de recherches
en physique des plasmas, Ecole polytechnique fddeYale de Lausanne, Lausanne, Switzerland.

461

462

STEINMETZ et al.

ION CYCLOTRON RESONANCE HEATING
Various heating schemes have been investigated on ASDEX:
Hydrogen second harmonic heating (2n, 67 MHz, in purehydrogen or D/H mixtures), hydrogen minority heating (D(H),33.5 MHz, nu/n+

= 5%), neutral beam injection (H•»• H ,+

H° + D , D° •»• H , D° -> D ) and the combination of NI with
ICRH. Thus, a thorough comparison with respect to heating,confinement, and plasma performance in a wide range ofparameters can be provided.

H-mode with pure ICRF heating
The H-mode has been realized - for the first time - with ICRH
alone in the D(H)-minority scheme at an absorbed RF power ofabout 1.1 MW / 1 , 2/. Thus, it appears that the dominant confinement regimes of auxiliary heated plasmas are found to be
independent of the heating method. At that power level the Hmode is achieved just marginally as shown in Fig. 1 where atypical example of an ICRF heated discharge is displayed incomparison to an L-type plasma. The H-mode, developing out ofan L-phase with sawteeth, shows all characteristic features
like a distinct drop in the H -signal in the divertor chamber(which is mostly a measure of the reduced power flux into thedivertor), rising electron density (indicating an improved particle confinement), enhanced plasma energy content (owing to
improved energy confinement), a significantly enhanced edge
electron temperature (rising from about 50 eV to 250 eV at
r/a = 0.98; not shown in Fig. 1 ) , and the existence of thetypical MHD-phenomenon, called ELMs (edge localized modes) /3/.
The sudden occurrence of ELM-free phases (marked by H*) changed
the antenna

coupling to such an extent that always one of thegenerators terminated operation by voltage breakdown. Unfortunately, this technical difficulty could not be avoided so far.

463

IAEA-CN-47/F-I-2
n H /n e -97.

1.6

FIG. 1. Plasma energy, line averaged electron density, and Ha signal of 1CRF heated discharges
with/without H-mode transition.

Hydrogen minority and second harmonic heating

Wall carbonisation prevented disruptions otherwise caused byexcessive impurity radiation and allowed the ICRH to work up to
maximal power (PTr, = 15 PJLU) without NI assistance. Thus, con1UUn
finement conditions of purely ICRF heated plasmas could bestudied. At high power in standard (double-null) configuration
both RF scenarios follow essentially an L-type scaling where
the D(H)-scheme appears to be slightly better. Figure 2
summarizes the energy confinement times normalized to thevalues of the Ohmic phases (which eliminates isotope effects onrg) for the D(H)-scheme and for 2 n ^ in hydrogen andhydrogen/deuterium mixture plasmas. The confinement degradationof both RF schemes seems, however, less severe than with beam
heating. This is supported by the observation that, with ICRH,
the particle confinement does only weakly degrade with respect
to the Ohmic phase which is indicated by the constancy of theplasma density during the ICRH pulse, and is supported by anincreased impurity confinement during ICRH with respect to the
NI L-phase (see also Fig.3).
Figure 3 illustrates the decay (transport) times of TiXX,
injected into the plasma by laser blow-off, indicating similar

STEINMETZ et al.

464».A NII = 60....95ms

_
o

+
• on
TVX
*

2 il CH
DIH)
2HCH*NI
0IH1-NIa = 0.6
a = 0.6
a = 0.7
a = 0.7

i.o-j,D(H|.IN

4A»

2n CH *NIIH°-H',H°-D*l

0.5-

FIG. 2. Energy confinement times (normalized to the value of the Ohmic phase) versus total power.

trends for impurity/particle transport and confinement. The
transport analysis suggests an enhancement of the inward drift
rather than a reduction of the diffusion coefficient withrespect to NI. It is not yet clear whether the improved
particle confinement or the absence of plasma rotation during
ICRF heating has an influence on the degree of energy confinement degradation.

Combination of ICRH and NI

Both RF heating schemes have been combined with neutral beam
injection up to total heating power of P

= 4.5 MW = 50 P' .
tOt
UnThe injection of either deuterium or hydrogen together with
2 a -launching clearly indicated bulk heating as the main ICRF
On
absorption scenario. Heating by direct coupling of the wave
with the injected particles was found to be negligible / 1 , 9/.
When combining ICRH + NI the degradation of confinement at highpower seems to be again less severe than with beam heating
alone (Fig. 2 ) . In particular, ICRH + NI data indicate theexistence of a substantial residual energy confinement at high465

IAEA-CN-47/F-I-2

TE*(2nCH]/TE*(0H)

0

0.5

1.0

+

ICRHI2ACHI*

NI.ICRH I2UCH]

1.5

2.0

P,C + PNI (MW)FIG. 3. Transport times of Ti XX compared to global energy confinement.

power (T*-plateau). The corresponding values T|/T|(OH) =0.43 - 0.50, where T*+ aP I C + 0.85 P N I ) , areconsistent with a T"-analysis modelling the increase of plasmaenergy content following an RF power step by W (t) = W (OH)TC
+AW • (1-exp(-t/Tp)) which gives - under certain assumptions P&
a measure of the ultimate energy confinement time at "infinitely" large auxiliary heating power. In these experiments theplasma energy content has been increased by a factor of 5, from
26 kJ at P.,, = 0.4 MW to 123 kJ at 4.4 MW total heating power.
The improved confinement of the combined scenario is also
supported by a reduced impurity transport (Fig. 3)Hydrogen second harmonic heating of two-species plasmas
Second harmonic heating has also successfully been applied to
H/D mixtures of hydrogen concentrations where conventionalhydrogen minority heating at the fundamental frequency does nolonger work. In H/D mixtures where the resonant species (H ) isa relatively small fraction of the total particle content the
RF heating efficiency, defined as increase of plasma energyICcontent per MW launched power (AWD / P T P ) , is found to bestrongly enhanced beyond that of pure hydrogen operation(Fig. 4) probably due to the isotope effect on confinement as

STEINMETZ et al.

466
40-

. .

2 f l C H * NI(D°—D')

D(H].NI(D*—D*]
2nCH.NI(H'-D'l
2nCH.NI(H'-H'|
PN,= 1.9MW
PIC =1.0 MW

0

0.2

0.4

0.6

0.8

1.0

nH/nc

FIG. 4. ICRF heating efficiency at various hydrogen concentrations.

observed also with NI heating. At the lowest hydrogenconcentrations (n /n = 0 . 1 ) the heating efficiency of the"2fi,, -minority" scenario /1 / ( = 36 kJ/MW) compares well withthat of the conventional D(H)-hydrogen minority mode ( = 3 5
kJ/MW) and is well above that of 2n -heating in pure hydrogenUn(= 21 kJ/MW). Unsuccessful application of 2finu to helium andpure deuterium plasmas clearly ruled out the possibility of 4th
harmonic heating.
On the basis of these results second harmonic heating appears
as an attractive scenario for a reactor-like DT-plasma where
efficient heating is expected to be performed much easier inthe 2flp -scheme than in a three species minority scenario(e.g. hydrogen minority in a DT-mixture) owing to the probably
difficult control of the minority concentration within a tightwindow of n /n .
H

6

Impurity behaviour
The ICRF related reduction of particle and impurity transportwith respect to NI (see Fig. 3) leads to an enhanced impurity
level which is dominated by iron released from the torus walls/4/. The power radiation of about 35 %, mainly originating fromthe plasma centre (Fig. 5 ) , limits high power ICRF operation to467

IAEA-CN-47/F-I-2\ .

wait carbonization**r• 0-15

0

10

20

30

40

r(cm)
F/G. 5. Radiation density profiles for different torus wall conditions.

P< 1.5 MW if no wall carbonisation or NI preheating isprovided. Carbonisation of the torus wall allowed an almost
iron free plasma with a radiation level of about 15 % (Fig. 5)
where the residual peaking sometimes observed in the radiation
profile results mainly from an inperfect carbon coating. In
addition, another mechanism of impurity release has been
determined which is related to wave absorption: Conditions ofpoor central wave absorption have always been accompanied by aparticular large central radiation as verified at e.g. 2flinhelium or pure deuterium discharges. The anti-correlation ofcentral wave absorption and central radiation level indicatesthat RF power absorbed in the outer region of the plasma,leading to strongly enhanced CX fluxes at the plasma edge /9/
and thereby increasing the metal impurity release, might play adeleterious role in terms of ICRH induced impurity generation.

Lower Hybrid Experiments

Heating and current drive with Lower Hybrid (LH) waves at f=

1.3 GHz have been studied in Ohmic target plasmas and in combination with Nl-heating (P< 3 MW) for various LH wave

468

STEINMETZ et al.
4.0

3.5-

Te
(keV)

# 18424
O:0H
• :LH

3.0-

n e = 06 x 1013crrf3
2.5-

PLH=375kW(Aip=90°)

2.0-

1.5 •

1.0 •

0.5-

-50

-40

-30

-20

-10

0

10

20

30

r(cm)

FIG. 6. Radial electron temperature profiles during Ohmic and complete LH current drive (Uj = 0);
lp = 500 kA.

spectra with transmitted power up to P. „ = 1 MW and pulse
duration T & 1.5 s. Efficient modification of the current density profile

can be achieved by appropriate shaping of the LH

wave spectrum.
Lower Hybrid Heating
Strong electron heating is obtained when LH waves are absorbed
by suprathermal electrons below the onset of ion interaction(n< 3 x 10 ^ cm

on ASDEX). At low density the electron

temperature profile peaks with the central value strongly increasing. In Fig. 6 the T (r) profiles are shown during the
Ohmic phase and during complete steady state LH current drive(U. = 0) at na = 0.6 x 10 1 3 cm"3 (P 0H = 240 kW, P^gs = 375 kW).
A.

6

With LH current drive much higher temperatures are observed in
the central region inside r = 10 cm (a = 40 cm) with T (0) :2 •* 3.5 keV. Qualitatively similar results are obtained in the

IAEA-CN-47/F-I-2

469

LH heating mode with stationary wave spectra. The zone oftemperature increase in all these cases seems to be restricted
to the region inside the q = 1 surface. The observed strong
central heating is not due to the suppression of sawteeth which
is only attained in the case of LHCD but not in the heating
mode.
Power balance calculations are carried out for the LH heating
and current drive modes (LHH, LHCD). The absorption coefficient,
as determined from the rate of change of the plasma energy content at the beginning of the RF pulse, decreases from about 0.8
at n= 1 x 1 0 1 3 cm" 3 to about 0.3 at n g = 2 x 10

cm" 3 . This

could be explained by a decrease of the RF power accessible tothe plasma centre /5/. The absorption coefficient also decreases with increasing RF power, thus indicating saturation
effects. The energy confinement with LHH and LHCD is improved
at low density above the Ohmic values (n = 0.6 x 10
cm ,
OH
IH
I = 300 kA: li! = 25 ms -> tZ, = 50 m s ) . With increasing RF
P

CJ

CM

power the energy confinement time decreases. The region of
improved confinement seems to be limited to the parameter range
where the total power input with application of LH is reduced
below the power input in the Ohmic phase.
LHH and LHCD have been combined with NI at high power levels:
P. < 1 MW, P..T < 3.H MW. At low density the same L scalingl_»ri
NX
for the confinement is found for NI and for the combination
with LH as shown in Fig. 7. LH heating itself is improved bypreheating from NI, mainly due to an additional direct interaction of LH waves with the beam ions in the plasma centre.
These are accelerated up to twice the injection energy as
observed in charge exchange measurements /9/.
Lower Hybrid Current Drive
Efficient saving of voltseconds in the primary OH transformer
can be obtained in the mode of combined RF and inductive

470

STEINMETZ et al.

100-

wp
(kJ)

50o : OH
A : LH• : NlA : NULH

FIG. 7. Plasma energy content WP versus total injected heating power PM for Nl, LHCD and the
combination of both (ne = / . / x W'3 cm~3, Ip = 300 kA).

current drive. The fraction of OH power replaced by LH power/ P A U ) a n d t n e corresponding conversion rate

(-/IP
LH

(-APnu /P r u ) are shown in Fig. 8 as functions of P.„ for different densities. The total plasma current is kept constant inall these experiments by a feedback regulation of the OH primary circuit. High efficiencies for replacement of OH power by
LH power can be achieved at low P

where the LH generated
LH

suprathermal electrons increase the electrical conductivity andthereby improve the OH current drive efficiency /6/. At low
densities a large reduction in OH power can be obtained with a

LH OH
high conversion r a t e for the LH power: - APUnu/P_0 Hu = 1 with
H
_
i Jo
_o
nH
-APj,
/P
*
1
for
n
<
0.5
x
10
cm
.
T
H
un
Liti
e
In combination with Nl the LH current drive efficiency is
reduced, as shown in Fig. 9 where the drop in OH power upon

471

IAEA-CN-47/F-I-2

LH
"OH

-AP

200

400

—I

1

600

800

1000

FIG. 8. Drop in OH power input —AP^/P^ during LHCD and corresponding conversion rate ofLH
versus P^; Ip = 300 kA.

power -LHCD is plotted versus P N T : It decreases with increasing beam
power, probably due to the direct absorption of LH power by thebeam ions which then reduces the power flow to the electrons.
This is also indicated by the strong drop in hard X-ray
intensity and suprathermal ECE emission from fast electrons
during the presence of fast beam ions in the plasma.

Current Profile Modification

The current profile j(r) during LHCD is no longer determined by
Spitzer resistivity. The T (r) and the j(r) profiles are de6

coupled by LHCD in OH target plasmas and in NI heated plasmas.
The current profile broadens while the T (r) profile peaks. The
broadening of j(r) as derived from the drop in internal
inductance 1. normally increases with the amount of the RFdriven current. A different behaviour is found in experiments

472

STEINMETZ et al.
NI +LHCD
IT, =1.1«1013cm"3l p =300KA
B, =2.2T

200-\

H'—D'

"APOH

(kW)

• \^

P LH =720kW

100 -

PLH=540kWi

2
P NI (MW]FIG. 9. Drop in OH power input with LHCD in combination with NI as junction of PNI.

where the form of the LH wave spectrum was changed gradually,by shifting power from high to low phase velocities, with thetotal power kept constant. This leads to a gradual decrease of
the total RF-driven current whereas the drop in 1 increases inthe same way /7/. The LH waves with lower phase velocities
therefore seem to be absorbed closer to the plasma periphery.
The local off-axis LH-current drive then leads to the stronger
broadening of j(r). An external control of the local plasma
current distribution by LH-current drive with an appropriate
wave spectrum and the related deposition profile seems
therefore possible.
By broadening j(r) such that the central q-value increased
above 1, sawteeth could be stabilised with LH-current drive in
OH- and Nl-heated plasmas /8/. With normal current drive

IAEA-CN-47/F-I-2

473

FIG. 10. Waveforms of density ft,, total radiated power Prad and diamagnetic @p for a sawtoothing NI
heated discharge (dashed lines) and for a sawtooth-free discharge with NI and LHCD (solid lines);
Ip = 380 kA, B, = 2.2 T, q(a) = 2.75.

spectra for this about half the power of complete LH-current
drive is necessary when LH is applied to OH plasmas. With NI
the LH power required for sawtooth suppression decreases withincreasing beam power. Stabilisation of the sawteeth results inhigher central electron temperatures and in slightly improved
global energy confinement as seen from Fig. 10.

SUMMARY

The application of waves in the frequency range of ICRH or LH on
ASDEX has been proved as a useful tool for either heating or
current drive. ICRH efficiently heats the plasma in various
2n u -and D(H)-regimes where the H-mode has also been achieved.
In combination with NI, t* is improved above the usual L-type
scaling where at high power a substantial residual energy confinement is observed. Strong electron heating is found with LH in
the suprathermal electron regime with confinement better than474

STEINMETZ et al.

Ohmic at low density. In combination with NI,acceleration of beamions is observed. Decoupling of T (r) and j(r) by LHCD allows
independent control of current and heat deposition profiles.

REFERENCES
[1]
[2]
[3]
[4]
[5]
[6]
[7]
[8]
[9]

STEINMETZ, K., et al., in Controlled Fusion and Plasma Heating (Proc. 13th Europ. Conf.
Schliersee, 1986), Vol. 2, European Physical Society (1986) 21.
WAGNER, F., et al., Plasma Phys. Controll. Fusion 28 (1986) 1225.
KEILHACKER, M., et al., Plasma Phys. Controll. Fusion 26 (1984) 49.
JANESCHITZ, G., et al., in Controlled Fusion and Plasma Heating (Proc. 13th Europ. Conf.
Schliersee, 1986), Vol. 1, European Physical Society (1986) 407.
SOLDNER, F., et al., in Controlled Fusion and Plasma Physics (Proc. 12th Europ. Conf.
Budapest, 1985), Vol. 2, European Physical Society (1985) 244.
LEUTERER, F., et al., Plasma Phys. Controll. Fusion 27 (1985) 1399.
LEUTERER, F., et al., in Controlled Fusion and Plasma Heating (Proc. 13th Europ. Conf.
Schliersee, 1986), Vol. 2, European Physical Society (1986) 409.
SOLDNER, F. ( et al., Phys. Rev. Lett. 57 (1986) 1137.
RYTER, F., et al., in Controlled Fusion and Plasma Heating (Proc. 13th Europ. Conf.
Schliersee, 1986), Vol. 1, European Physical Society (1986) 101.

DISCUSSION
S.-I. ITOH: In your summary you said that the confinement structure was notdependent on the heating method. On the other hand, in the case of the particle confinement time (obtained by T; line radiation), there is a discrepancy between neutral
injection and ICRF heating. What kind of confinement is it that is not dependent on
the heating method?
Secondly, with respect to LH heating you stated that the heating was restricted
inside the q = 1 surface. Did you see any MHD instabilities due to the sudden risein Te and/or v Te inside the q = 1 surface.
K. STEINMETZ: Non-dependence of the confinement structure on the auxiliary heating method means the general existence of an L-type and H-mode
behaviour with NI as well as with ICRH. With respect to TE or TP, both heating
schemes exhibit the same type of scaling, as shown in Fig. 3 of the paper, although
the confinement for ICRH is slightly improved compared to that for NI. So there isno discrepancy in terms of the global confinement structure.
No MHD stabilities due to the central electron heating by LH waves were seen.
For application of LH alone, the heating efficiency is similar in the current drivemode without sawteeth and in the heating mode in the presence of sawteeth.

IAEA-CN-47/F-I-2

475

H. MATSUMOTO: In your conclusions you said that you have observed an
improvement in confinement during ICRF second harmonic heating and during NBI
heating through an increase in the heating power. However, this may simply be due
to an improvement in absorption of the wave at the second harmonic resonance when
the plasma beta was increased. What is your opinion?
My second question is, did you use the constant power absorption coefficient awhen evaluating the RF heating power input to calculate the global confinement timein Fig. 3? And further, how did you determine a?
K. STEINMETZ: When the plasma beta was increased by neutral beam injection, the experimentally derived effective ICRH power absorption was found toincrease from aIC = 0.6 to ai C = 0.7. Since the experimental values of a are used
to calculate TJ!, the observed improvement in confinement of ICRH plus NI, withrespect to NI only, cannot be attributed to the measured change of a.
For pure ICRH heating, a = 0.6 is taken, while for the combined ICRH plus
NI scheme, a = 0.7, as shown in Fig. 2 of the paper. The effective power absorption is derived via the rate of change in the plasma beta, d/3/dt, when the RF is
switched on or off, and by modelling the increase in the plasma energy content following an RF power step (see the paper presented by our group at the 13th European
Conference in 1986).
R.R. WEYNANTS: Does the effectively absorbed power, a, change significantly when going from one ICRH heating scenario to the other?
K. STEINMETZ: No. The ICRH absorption coefficient aIC = 0.6 is about
equal in the 2QCH and the D(H) minority regimes. When ICRH is combined withneutral beam injection, the effective ICRH absorption increases to aIC = 0.7. Ishould note that we have cross-checked our procedures used to determine a by applying them to purely Nl-heated plasmas and found good agreement between the numerically computed values using a beam deposition code and the experimentally derived
values of aNI = 0.9-1.

IAEA-CN-47/F-I-3

ION BERNSTEIN WAVEHEATING EXPERIMENTS ON PLT
M. ONO, P. BEIERSDORFER, R. BELL, S. BERNABEI,
A. CAVALLO, A. CHMYGA1, S.A. COHEN, P. COLESTOCK,
G. GAMMEL G. GREENE, J. HOSEA, R. KAITA,
I. LEHRMAN, G. MAZZITELLI2, E. MAZZUCATO,
D.H. McNEILL, M. MORI3, K. SATO4, J. STEVENS,
S. SUCKEWER, J.R. TIMBERLAKE, V. VERSHKOV1,
J.R. WILSON, A. WOUTERS
Princeton Plasma Physics Laboratory,
Princeton University,
Princeton, New Jersey,
United States of America
Abstract
ION BERNSTEIN WAVE HEATING EXPERIMENTS ON PLT.
Ion Bernstein wave heating (IBWH) has been investigated on PLT, with up to 800 kW of
RF power applied to the antenna. A plasma antenna loading of 2-4 J) has been observed, resulting
in 80-90% of the RF power being coupled to the plasma. An ion heating efficiency ofATi(O) ^ / P R P = 6 eV X 1013 cm" 3 /kW, without high energy tail ions, has been observed, up to the
maximum coupled RF power of 650 kW (P 0 H = 550 kW). The particle confinement during highpower IBWH increases significantly (as much as 300%). Associated with this increase are a longer
confinement time of the injected impurity, a reduced drift wave turbulence activity, frequency shifts ofdrift wave turbulence and the development of a large negative edge potential. The energy confinement
time, however, shows some degradation from the Ohmic value, which can be attributed to the enhanced
radiation loss observed during IBWH. The ion heating and energy confinement time are relatively
independent of plasma current.

1.

INTRODUCTION

Plasma heating by externally launched ion Bernstein waves (IBWH) has been
investigated as an alternative to conventional ICRF heating. Because of theirrelatively short wavelength, ion Bernstein waves can heat the bulk ions even at
relatively high harmonic frequencies [1].
' Kharkov Physico-Technical Institute, Ukrainian SSR.
Associazione Euratom-ENEA sulla Fusione, Centro Ricerche Energia Frascati,
Frascati, Rome, Italy.
Japan Atomic Energy Research Institute, Tokai-mura, Naka-gun, Ibaraki-ken, Ja
Institute of Plasma Physics, Nagoya University, Nagoya, Japan.

2

477

478

ONO et al.

To preferentially excite ion Bernstein waves (as opposed to fast waves), a
B<>-Ez type loop antenna [2] was used to simulate lower-hybrid-like waveguidefields. The RF current for this antenna flows in the toroidal direction, whereas thecurrent for the fast wave antenna flows in the poloidal direction. In 1985, experiments were performed with such a loop antenna, and various efficient heating
regimes were identified, including 5UD, H minority, 3He minority, and 3/2 QD[3, 4]. The RF power in these experiments was 100-150 kW. The observed ion
heating quality factor ATj(O) n<,/PRF was about 3-6 eV x 1013 cm~3/kW for theseregimes. These initial experiments demonstrated that:— IBWH heats relatively high harmonics of the ion cyclotron frequency — aproperty that is attractive for a compact IBWH waveguide launcher;— IBWH interacts non-linearly with subharmonic frequencies [2, 4], giving riseto new heating scenarios;— IBWH produces very few high energy tail ions. The heating is essentially bulk
ion heating, which is of interest for an understanding of plasma confinementwith auxiliary heating.

2.

HIGH POWER IBWH ANTENNA

Since these results were encouraging, in 1986 further IBWH experiments were
performed on PLT to investigate the higher power regime (PRF > POH) whereplasma confinement issues can be tested during IBWH. It was of particular interest
to study the confinement properties for bulk ion heating in comparison with those forhigh energy ion heating such as fast wave ICRF heating or neutral beam heating. Twohigh power IBWH antennas were installed on PLT. The new high power antennas
are 60 cm long and 5 cm wide. The antenna design is very similar to that of the PLT
high power ICRF antenna. The distance from the current strip to the back plate is
5 cm, providing adequate flux linkage to the plasma. The antenna is covered withMacor cover plates. Between the Macor plates and the plasma there are two layersof stainless steel Faraday shields. The antenna is protected by graphite plates placed
toroidally on both sides of the antenna. Up to 500 kW of RF power can be applied
to one antenna. The plasma loading for these antennas is 2-4 fi, resulting in 80-90%
of the applied RF power being coupled to the plasma. The overall antenna
performance is similar to that of the PLT high power fast wave antenna.

3.

IBWH ION HEATINGIBWH at high power levels was performed on PLT in the 3He and 3/2 fiD
regimes utilizing the existing ICRF 30 MHz transmitter. A typical heating result isshown in Fig. l(a) where the time evolution of the ion temperature is shown for479

IAEA-CN-47/F-I-3
1

1

1

T 1

1

1

T

2.0 -

0 Doppler

(a)
•"•

a A

I
1.5 -

A

,

/I

Cx

• Neutron

I

-

I A
A

A

A
I

A

1.0 -

A
A
A
r

A

J-A

A .

a A

0.5 "A A

-

Prf A

n

500 T
(kWV

300

«

400

500

1

600
700
TIME(ms)

1

1

800

900

0.3
0.4
Prf (MW)

FIG. 1. (a) Typical time evolution of ion temperature. Pgf = 500 kW, F = 30 MHz, Bo = 29 kG,
3/2 QD, ne = 3.0 x 1013 cm'3. Other parameters are given in Fig. 2(a-c).
(b) Ion heating as a function of RF power for 3He minority (Bo = 32 kG, 10% 3He) and 3/2 QD(Bo = 29 kG, 0% 3He) regimes.

3/2 QD heating at 500 kW. Three Tj diagnostic techniques — Doppler broadening(Ti XXI), charge exchange (CX), and neutron emission — are in good generalagreement within the uncertainties of each diagnostic. The charge exchange velocity
distribution is Maxwellian, without any high energy tail ions up to the highest power
level. The ion temperature profile as measured by the Doppler broadening of various
ion lines is similar to the Ohmic profile. The ion heating efficiency as shown in
Fig. l(b) is nearly linear up to 650 kW, with a heating quality factor of
ATifle/PRp = 6-7 eV X 1013 cm~3/kW. It is interesting to note that 3He heatingand 3/2 QD heating have similar heating efficiencies. The toroidal magnetic field forONO et al.

480
i

i

\

i

i

r

(o)
3

IBW
OH

'E
<j

o 2

^ « % ^ ^ ^ ^ % % % % % ^ ^ ^ : OH
GaSPuff|

rf

,IBW

200

300 400 500 600 700 800 900
TlME(ms)

200

300

6

"40

400 500 600 700
TlMEIms)
3r

-20
0
20
RADIUS (cm)

40 -30800 900-20
0
20
RADIUS (cm)40

FIG. 2. (a) Density time evolution for the IBWH case (solid curve) shown in Fig. l(a) and for the
Ohmic case (dashed curve) in which gas puffing was used to simulate the density rise of the IBWH case.
(b) Time evolution of Ha emission for the IBWH case and the simulated Ohmic case, near theantenna-limiter region.
(c) Density profiles at t = 600 msfor the IBWH case and the simulated Ohmic case shown in Fig. 2(a).
(d) Corresponding electron temperature profiles.

3/2 fiD heating is about 10% lower than that for 3He heating, and, of course, for the
3/2 QD regime no 3He injection is required. In the 500-600 kW range, the central
Doppler ion temperature reaches about 2 keV, which exceeds the central televisionThomson scattering (TVTS) electron temperature. The time evolution of the plasma
density is shown in Fig. 2(a). The density and electron temperature TVTS profiles
at t = 600 ms are shown in Fig. 2(c,d). The dashed curves are for an Ohmic casewith the gas feed programmed to yield a similar density evolution. As can be seen,
the density and electron temperature profiles are very similar for the two cases.

IAEA-CN-47/F-I-3
4.

481

IMPROVED PARTICLE CONFINEMENT DURING IBWH

During IBWH, the particle confinement is significantly improved [5]. A good
example is shown in Fig. 2(a). The plasma density increases by more than a factorof three, without active gas puffing and without increasing particle recycling, as
indicated by a drop in the H a emission. The largest drop occurs near the antenna,
limiter and gas valve regions. As shown in Fig. 2(b), the Ha emission near theantenna-limiter region (where much of the recycling takes place) during IBWH (solid
curve) is considerably less than in the simulated OH case (dashed curve). The density
profiles during IBWH are similar to those for the simulated Ohmic case, as shownin Fig. 2(c). The rapid density drop after IBWH as compared with the Ohmic casedemonstrates the reduced particle recycling ratio resulting from the reduced total
particle inventory in the system. It should be emphasized that the density rise during
IBWH is not correlated with the impurity influx as measured by various spectroscopic
instruments — a bremsstrahlung based 'Z'-meter and a scanning bolometer. Also, thelarge increase in the neutron level (a factor of 500) shows that the incremental density
is due mainly to deuterium.
In connection with this improvement in particle confinement (or reduced
recycling) a number of interesting phenomena were observed. During IBWH, the low
energy neutral flux levels (at energies less than 1 keV) decrease significantly [6].
This drop may be related to the drop in the H a emission. For a further understanding of the particle transport processes during IBWH, a very small amount of selenium
was injected with a laser blow-off impurity injector. As shown in Fig. 3(a), the decay
time of the Se XXV radiation increases by a factor of two from the Ohmic level (from50 ms to 100 ms). Since during the observation time the central electron density andtemperature are similar and nearly constant for both cases, the selenium line
behaviour gives an indication of the central selenium confinement. A similar
behaviour was observed with a helium gas puff. When a given amount of helium is
injected with a pre-programmed fast gas valve, the net density rise during IBWH is
at least twice that in the case without RF, which again suggests an improvement in
the particle confinement time. The high ionization states of carbon (C V, C VI) do
not rise during IBWH (they may even drop); this represents an exceptional case [6].
Another striking effect of IBWH is the change of the low frequency turbulence
activity in the plasma. A microwave scattering system was used to investigate theturbulence in the half-radius region between the q = 1 and q = 3 surfaces(r = 25 + 10 cm). Figure 3(b) shows the time evolution of the scattering signals at
100 kHz for the IBWH case and the simulated Ohmic case (the scattered wavenumber is 6-8 cm" 1 ). One can see a significant drop in the turbulence level during
IBWH as compared with the simulated Ohmic case. In the insert, the frequency
spectrum for each case is shown at the end of RF (t = 700 ms). A frequency shift
as well as an amplitude reduction occur during IBWH. Since the plasma density
profiles are very similar in the two cases, the possibility of a change in the scattering
volume can be ruled out. From the frequency shift, one can infer a net increase of

482

ONO et al.

-300-200-100 0 100 200 300
f (kH2)

XI 400 500

600

1

1

o
_
2

(C)

1

-100 -

700 800 900 1000 MOO
TIME(ms)
1

1

1

1
-

g-200

\

V

^-300

i
i
1
1
1
300 400 500 600 700 800 900 1000 MOO
TIME(ms)
1

iFIG. 3. (a) Time evolution of the selenium XXV resonance line intensity for the IBWH case and thesimulated Ohmic case. The selenium was injected at t = 550 ms.
(b) Time evolution of the integrated low frequency turbulence level at 100 kHz for the two cases;k(scatt) = 6-8 cm'1 at r = 25 ± 10 cm.
Insert: Frequency spectrum at t = 700 ms for the two cases.
(c) Time evolution of the plasma edge floating potential.

the poloidal rotational velocity of approximately 5 x 104 cm/s in the electron
diamagnetic drift direction. The net result is a near doubling of the electron diamagnetic drift velocity. This drift velocity corresponds to an electric field of 15 V/cm
in the radially inward direction. It should be noted that this reduction in the drift wave
activity correlates well with the long impurity particle confinement time. Measurements with an RF-shielded Langmuir probe in the plasma edge have shown that

IAEA-CN-47/F-I-3

483

during IBWH a large floating potential develops which is negative with respect to thechamber wall (Fig. 3(c)). This is correlated with the appearance of the frequency
shift in the turbulence spectrum.

5.

ENERGY CONFINEMENT TIME DURING IBWH

Plasma confinement during IBWH is of interest because of the bulk ion heating
properties of ion Bernstein waves. To investigate the scaling of energy confinementwith plasma current, the current was varied between 200 kA (q^e = 10) and
630 kA ( q ^ = 3), with a constant RF power level of 250 kW. Note that the ratioof the RF power to the Ohmic power increases with decreasing current, reaching 1
for the 200 kA case. For a given RF power, the ion temperature increase remains
constant down to the lowest current of 200 kA. The normalized energy confinement
time, TEI%, also remains constant. This may be due to the characteristic of IBWH
bulk ion heating that fast ion losses are negligible, even for the low current case. For
a given current, Ip = 500 kA, the power scaling shows a gradual degradation ofconfinement time with increasing RF power. At 650 kW, the confinement time isabout 40 ms at 4.5 x 1013 cm" 3 , which is significantly below the Ohmic value of
50-55 ms but is higher than the L-mode scaling time of about 33 ms. This drop in
the global confinement time can be attributed to an increase in the electron loss
channel. However, in view of increased radiation losses during IBWH, the electron
thermal diffusivity still appears to be Ohmic like.

6.

IMPURITY INFLUX DURING IBWH

During IBWH, an increase in the influx of metallic impurities (predominantlyiron) was observed. This high-Z impurity influx, together with the longer particle
confinement time, can result in non-negligible central radiation losses during highpower (high density) IBWH. Thus, for a successful high power IBWH experiment
it is important to reduce the iron influx level as much as possible. To find the sourceof the iron impurities, the outer Faraday shields of one of the antennas were coatedwith a 20 pm thick carbon film. This antenna was compared with an uncoatedantenna. For the same RF power level, the carbon coated antenna produced half as
much iron influx (and half as much central bolometric radiation) as the uncoatedantenna. The ion heating efficiency and the density rise, however, remained similar
for the two antennas. This indicates that at least half of the iron influx originated fromthe Faraday shields. The other half may have come from the vacuum chamber wallor the uncoated inner Faraday shield. An antenna with both inner and outer carbon
coated Faraday shields is being prepared and will be used in an investigation of thispoint.

484
7.

ONO et al.
CONCLUSIONS

Efficient ion heating by IBWH was observed in the 3He and 3/2 flD regimes,
up to the highest power level (PRF = 650 kW). With the application of IBWH,
significant improvement in the particle confinement time was observed. Associatedwith this improvement were a longer confinement time of the injected selenium, a
reduced low frequency turbulence level, frequency shifts of drift wave turbulence andthe development of a large negative edge potential. The global plasma energy
confinement time at the highest power levels showed some degradation from the
Ohmic level, which is attributed to the increases in the electron loss channel that are
due mainly to the radiative losses. The ion heating and the energy confinement
remained relatively constant with plasma current. The ion energy distribution
remained Maxwellian for the entire power range.

REFERENCES
[1]
[2]
[3]
[4]
[5]
[6]

ONO, M., WURDEN, G.A., WONG, K.L., Phys. Rev. Lett. 52 (1984) 40.
ONO, M., WATARI, T., et al., Phys. Rev. Lett. 54 (1985) 2339.
WILSON, J.R., et al., in Radiofrequency Plasma Heating (Proc. 6th Top. Conf. Callaway
Gardens, GA, 1985), American Institute of Physics, New York (1985) 8.
ONO, M., et al., ibid., p. 83.
HOSEA, J.C., et al., Plasma Phys. Controll. Fusion 28 (1986) 1241.
WILSON, J.R., et al., J. Nucl. Mater. 145&146 (1986).

DISCUSSION
R. KOCH: Where does the power limitation come from in your experiment?
M. ONO: At present, the power limitation is due to an arc in one of the twoion Bernstein wave antennas. With an improved antenna and with more operational
experience, the power limit can be raised relatively easily.

IAEA-CN-47/F-I-4

CONFINEMENT STUDIES ON TEXTOR
WITH HIGH POWER, LONG PULSE ICRH
P.E. VANDENPLAS,
T. DELVIGNE,
P. DESCAMPS,
F. DURODIE,
M. JADOUL, R. KOCH, D. LEBEAU, A.M. MESSIAEN1, D.I.C. PEARSON,R. VAN NIEUWENHOVE,
G. VAN OOST,G. VAN WASSENHOVE,
R.R. WEYNANTS1

Laboratoire de physique des plasmas — Laboratorium voor Plasmafysica,
Association Euratom-Etat beige — Associatie Euratom-Belgische Staat,
Ecole Royale Militaire — Koninklijke Militaire School,
Brussels, Belgium
H.L. BAY, G. BERTSCHINGER, W. BIEGER2, P. BOGEN, W. BRUSSAU,
G.A. CAMPBELL3, Y. CAO 4 , R.W. CONN 3 , H. CONRADS, K.H. DIPPEL,
H.G. ESSER, K.H. FINKEN, G. FUCHS, B. GIESEN, D.M. GOEBEL3,
E. GRAFFMANN, H. HARTWIG, E. HINTZ, F. HOENEN, K. HOTHKER,
W.L. HSU 5 ,A. KALECK,
H. KEVER,
L. KONEN,
M. KORTEN,
W.K. LEUNG 3 ,L. LI 4 ,Y.T. LIE,
A. POSPIESZCZYK,
D. REITER,
A. ROGISTER, G. ROSS 6 , D. RUSBULDT, U. SAMM, J. SCHLUTER,
B. SCHWEER,
H. SOLTWISCH,
G.J. THOMAS 5 ,F. WAELBROECK,
G. WAIDMANN, P. WIENHOLD, J. WINTER, G.H. WOLF

Euratom-KFA Association,
Institut fur Plasmaphysik,
Kernforschungsanlage Jiilich GmbH,
Jiilich, Federal Republic of Germany

Abstract
CONFINEMENT STUDIES ON TEXTOR WITH HIGH POWER, LONG PULSE ICRH.
A detailed analysis of the confinement properties of TEXTOR with ICRH is made by means of(i) power, density, plasma current and toroidal magnetic field scans, for which the other plasma
parameters are chosen to be as constant as possible, and (ii) RF amplitude modulation at different
frequencies. The results can be interpreted on the assumption that the total energy can be split into twoparts, each following its own confinement law; the residual OH part E 0 H , obeying the neo-Alcator typescaling of TEXTOR, and the additional heating contribution EAUX, depending on an auxiliary confinement time rAUX. The observed time evolution of the total energy during the discharges can be wellaccounted for by two uncoupled first order differential equations describing, respectively, E 0H andEAUX- In this way, a complete power or current scan during a single shot can be simulated accurately.
1

Senior research associate at Fonds national.de la recherche scientifique, Brussels, Belgium.
Fachhochschule Julich, Jiilich, Fed. Rep. Germany.
3
University of California, Los Angeles, CA, USA.
4
Institute of Plasma Physics, Academia Sinica, Hefei, China.
5
Sandia National Laboratories, Livermore, CA, USA.
6
INRS-ENERGIE, Varennes, Quebec, Canada.
2

485

486

VANDENPLAS et al.

It is found that TAVX increases linearly with Ip, is insensitive to the density and decreases slightly with
the RF power. The effectiveness of an increase of OH power at constant density is compared to thatof ICRH.

1.

INTRODUCTION

The general characteristics of ICRH and of stationary long pulse (~ 1 s), low
impurity ICRH operation have been reported previously [1-3]. This paper reports on
a detailed analysis of TEXTOR confinement by means of scans of the main
parameters of the heated discharge: plasma current, Ip; central chord density, Neo;RF power radiated by the antenna, PRF; and toroidal magnetic field, Bt. The
measurements are made either in steady state plasmas or during the dynamic evolution of the discharges through comparison with simulation programs. For the present
analysis, TEXTOR was operated with main and antenna carbon limiters and with a
hot carbonized liner [4]. The heating regime is either the minority (or 2WCD) regime(CH/CD < 1%) or the mode conversion regime (CH/CD = 5%). While the relativeelectron and ion energy increases are different for the two heating regimes [3], thetotal plasma energy increases are nearly the same for identical main dischargeparameters; the two regimes are, therefore, not distinguished. The operating range
is 1.5 < N^ < 6 X 1013 cm" 3 , 0.2 < Ip < 0.5 MA, 0 < PRF < 2.2 MW,
B, = 2 T; small radius a = 0.46 m, large radius RQ = 1.75 m.

2.

INTERPRETATION 'ANSATZ'

The very different behaviour of the total plasma energy content E with
OH power, P 0H = IpVf (Vf is the loop voltage), and with auxiliary ICRH power has
suggested to us [2] to split E according to the 'ansatz'
E = EQH + EAUX

(1)

where EQH is that part of the energy that is due to the remaining OH power,
POH = Vf'Ip, and EAUX is the part due to Pup. V,' is the reduced loop voltage in thepresence of ICRH. We further assume that EAUX = TAUXPRF and E 0H = TOHP6H
where TOH = F c i i N ^ is the neo-Alcator OH confinement law of TEXTOR. The
auxiliary confinement time rAUX is determined for steady state conditions fromrAux = (E -V;I P T O H)/PRF

and will be shown to be well approximated by Eq. (5).

(2)

IAEA-CN-47/F-I-4

487

When the plasma parameters are time dependent, EQH and EAUX are describedby:dEoH/dt = PQH — EOH/^OH;

dEAUX/dt = PRF -E A U X /T A U X(3)These two equations are only coupled through V,, which depends on PRP. The
single law dE/dt = Ptot — E/T E , where Ptot = P 0H + PRF a n d rE is the global
confinement time, cannot describe all aspects of the observed energy evolution.
When PRF is amplitude modulated, the set of Eqs (3) allows not only the value of
TAUX to be checked, but, if the total radiated power PRF is well coupled to theplasma, is also relevant in the following manner: (i) at a sharp switchoff of PRF, wehave A(dE/dt) = A(PRF + P O H); (») immediately after the switchoff, the energy
decay is essentially governed by dEAUX/dt = - E A U X / T A U X ; (iii) with sine modulation of PRF = PRPQ + PRF exp (-iwt), we have
EAUx = PRF^AUX (1 + C ^ A U X ) - * exp (i0), with <j> = arctan (a>rAUX)(4)

and a similar expression for EOH, which is negligible for the cases considered. Thenthe phase difference between E = EAUX and P ^ is a measure of TAUX and, if we use
the value of T AUX , the amplitude of E determines IPRFI. If the value of I PRP I
obtained in this way agrees with that actually radiated by the antenna, it is confirmedthat the radiated power is well coupled to the bulk of the plasma and represents the
PRF value to be used in Eqs (2) and (3). Another evidence for the coupling of PRP
to the plasma bulk results from the fact that the thermal load remains, with ICRH,
at least proportional to Ptot [2].
Let us, finally, note that, when Ip and Neo are held constant, the incremental
confinement time Tinc is linked to TAUX by
Tinc = dE/dPtot = (TQH dPon/dP^ + rAUX) (dPtot/dPRF)"'3.

STEADY STATE ANALYSISThe energy content is measured by four different methods: (i) diamagnetism,(ii) /3p computation from an equilibrium code using as input the current in thevertical field coils and the plasma current profile obtained from polarimetry, (iii) the
cosine coil method [5], (iv) kinetic evaluation using the data from the HCNinterferometer-polarimeter, from ECE and the neutron yield for Tio, identical T;
and Te profiles being assumed. The four methods agree well to within ±7 kJ.
TAUX is computed from Eq. (2) by using for TOH the neo-Alcator law defined in
Eq. (1) with a = 0.8 and the value of F 0 H , which fits the observed TOH value justbefore the RF pulse (depending on the particular shot, 10 < F 0 H < 14). Note thatlower values of a (e.g. a = 0.5) with the appropriate adjustment of FOH would also
give a reasonable fit of the OH behaviour.

488
(i)

VANDENPLAS et al.
Power scan

To decouple the effect of the density increase from the effect of the total
power increase, a careful selection of shots is made with the same value of Ne0
during ICRH. As was already stated in Refs [2, 3], the results agree well witha constant value of rAUX, and the global confinement time obeys the relation
TE = (1 - PRF/Ptot) TOH + (PRF/PIOI) TAUX, resulting from Eq. (1). For some setsof data, there is, however, a modest decrease of rAUX with power (<30% for
0.1 < PRF < 2MW).
(ii) Density scan
The global confinement time rE = E/Ptot is shown versus N^ in Fig. l(a) for
different values of Ptot = P 0H + PRF- The OH curve exhibits the linear dependenceof r 0H on NeQ. The density dependence of xE diminishes as PRF increases, and thecomputed value of rAUX, which corresponds to the limit PRF — oo, does not, within
the accuracy of measurement, depend on the density.
(iii) Current scanFigure l(b) exhibits the linear dependence of TAUX on the plasma current. We
havewith 50 < FAUX < 70 (ms, MA). In contrast to this, ToH/Neo decreases «q a a when
Ip increases. This is also shown in Fig. l(b).
(iv) Magnetic field scan
For a given plasma composition, the RF power deposition profile can be crudely
tailored by moving the mode conversion surface over the plasma cross-section, andconditions can be achieved where the major fraction of power is no longer depositedin the bulk. The central confinement then increases such that the overall energy
increment, nevertheless, remains fairly constant.

4.

DYNAMIC ANALYSISThe energy evolution is simulated by integrating Eqs (3) with N ^ t ) , Ip(t),PRF(t), Bt(t) as input data. An approximate empirical expression, derived fromsteady state data, is taken for V,' (i.e. V,' oc (1 - BPRFXI + CIp) in the operating
range).

489

IAEA-CN-47/F-I-4

>

I

.

i

i

(ms)
PtofOH

100•

ICRH

| *

TAUX |^+

0.35MW

x

0.6MW

. •«

0.8MW

80^. *

60-

+

40• •

+ #

,

, ^

20^21881 -23120

0-

1

1

1

1

r

1

1

Neo (10 13 cm- 3 )FIG. 1. (a) Density scan results for different values ofP,0, = P'0H + Pgf and constant plasma
current, Ip = 350 kA. Curve labelled OH corresponds to pure Ohmic heating case; curve TAUX
is computed from Eq. (2).

l

AUX

3p

(s/1013cm'3)

(ms)

N eo

0.04

300.0320-0.02

10-•001^21881-23120

0.6

Ip(MA)
FIG. 1. (b) Evolution ofTAVX and TOH/Ne0 versus Ip.

490

VANDENPLAS et al.

E
(kJ)
60 •

\

20 •

E

n
0

1 Time (s) 2

FIG. 2. (a) Complete dynamic current scan made in one shot (No. 23 114). Edia is the diamagnetic and E the corresponding simulated energy (FOH = 11; FAUX = 53; P^ = 1.57 MW). The
part corresponding to EAUX is also displayed.

FIG. 2. (b) Corresponding time evolution of Ip, Ne0 (HCN), Te0 (ECE) and Ti0 (neutrons).
Curve Ti0A is computed equipartition value from N^, Te0, Ip and B, measurements.

We take r 0H = FonNeoq"-8and rAUX = FAUXIp, where the exact value of
F 0 H is adjusted around 12 and that of FAUX around 60 (ms, 1013 cm" 3 , MA) to provide the best fit with the experimental evolution described by the diamagnetic signal.
A careful measurement of the time response of the raw diamagnetic energy signal
Edia,r (see Appendix) gives a time constant of Tdia = 12 ms, which is used to obtainE
dia = Edia>r + TdiadEdilM/dt as corrected experimental energy signal.

IAEA-CN-47/F-I-4

491

TIME{s)FIG. 3. (a) Dynamic power scan in one shot (No. 24 257). Edu, is compared to simulated E(F0H = 12.3; FA(JX = 58.5; Ip = 480 kA); Pgp is also shown. Decay of Ip starts at end of
RF heating.

1

TIME (s)FIG. 3. (b) Edia compared to simulated E (F0H = 12.8; FAVX = 58; Ip = 280 kA) for
amplitude modulated RF pulse. Pgf is also shown. Inset shows enlargement of Eiia decay ascompared to E after abrupt ending of RF pulse (No. 24 467).

Figure 2(a) shows a comparison of the simulation with a complete current scanin one shot; EAUX is also displayed. The time evolution of Ip, Neo, the central
electron temperature T^ and the ion temperature Ti0 is shown in Fig. 2(b). The
value of T i0A as computed from equipartition with the electrons agrees well with theexperimental value of Ti0 (obtained from the neutron yield) during the OH parts ofthe shot. The simulated E and the observed Edia are compared in Fig. 3 for shots

492

VANDENPLAS et al.

with an amplitude modulated ICRH pulse. Figure 3(a) displays a complete power
scan, and Fig. 3(b) the results of sinusoidal amplitude modulation. The insert of
Fig. 3(b) shows the good fit during the energy decay after the sharp switchoff of the
RF pulse. Also, the phase shift between Edia and PRF is in agreement with thesimulated phase shift.

5.

CONCLUSIONS
From our investigations, we conclude:(1) The chosen 'ansatz' in which the effects of PRF and POH are decoupled, canaccurately describe the evolution of E during the shot in a large variety of situations.
(2) The validity of the parametric dependence taken for TAUX OC Ip andT0H « Neoqa" for TEXTOR is confirmed.
(3) The value of 7AUX is very different from that of rE = E/Ptot, showing thatthe supplied auxiliary energy has its own behaviour.
(4) It is verified that essentially all radiated PRF power is coupled to theplasma bulk when optimum values of B, and the H/D ratio are chosen.
The different nature of the confinement properties of OH and of RF additional
heating is again very apparent in the comparison of the incremental heating efficiencies 77 = AE/APtot for OH and ICRH. Figure 4 displays, at constant Ne0, the OH
energy evolution versus P 0H as Ip is increased, together with the different evolutionof the energy curve at the onset of ICRH, starting from OH for various values of Ip.

E
Ikj)

303 4 0 ^ - ^ * ^ 8 0 kA
— - RF
20-

10-

0.2

0.8

0.6
13

P... IMWl
3

FIG. 4. Evolution, at constant density N^ = 2.35 x 10 cm , of E0H versus P0H whenplasma current is increasedfrom 215 to 280, 340, 480 kA successively (plain line arrows). Dottedarrows also indicate, for each value of plasma current, how total energy E evolves from E0H at
onset of ICRH.

IAEA-CN-47/F-I-4

493

While for the Ne0 value considered r]0U > T;ICRH at low Ip (200 kA), and
Von = I?ICRH (Ip = 350 kA), we have 7?ICRH > Von at high Ip (~500_kA).
Apparently, the heating efficiency of ICRH exceeds that of OH when Ip/Ne0 issufficiently large.

Appendix
MEASUREMENT OF THE TIME RESPONSE
OF THE DIAMAGNETIC SIGNALThe time constant of TEXTOR's diamagnetic signal [6] has been determined byvarying the frequency Q of the amplitude modulation of the ICRH pulses (Fig. 3(b))
and measuring the amplitude ratio and the phase difference between Edia r and PRF.
From these measurements, using a least squares fit procedure, it follows thatEdia.r/PRF <* [(1 + iOri)(l + IQT2)] ~', with T2 — TAUX and r, = 12 ms. The timer,, whose value is close to the wall time constant, is considered to be the time
constant Tdia intrinsic to the diamagnetic measurement system.

ACKNOWLEDGEMENTS
The authors would like to thank Drs T. Shoji and S. Okamura for suggesting
the power modulation experiment in combination with the diamagnetic
measurements.

REFERENCES
[1]
[2]
[3]
[4]
[5]
[6]

WEYNANTS, R.R., et al., Radiofrequency Plasma Heating (Proc. 6th Top. Conf. Callaway
Gardens, GA, 1985), American Institute of Physics, New York (1985) 40.
MESSIAEN, A.M., et al., Plasma Phys. Controll. Fusion 28 (1986) 71.
WOLF, G.H., et al., Plasma Phys. Controll. Fusion 28 (1986) 1413.
WINTER, J., et al., J. Nucl. Mater. (1986).
SWAIN, D., NEILSON, G., Nucl. Fusion 22 (1982) 1015.
FUCHS, G., et al., Rep. KFA-IPP-IB-I/84, Kernforschungsanlage Jiilich GmbH (1984).

IAEA-CN-47/F-I-5

EXPERIMENTS ON ICRF HEATING
AND FAST WAVE CURRENT DRIVE
IN JIPP T-IIU
T. WATARI,
K. OHKUBO,
R. AKIYAMA,
R. ANDO,D. ECKHARTT1,
Y. HAMADA,
S. HIROKURA,
K. IDA,
E. KAKO, O. KANEKO, K. KAWAHATA, Y. KAWASUMI,
S. KITAGAWA, T. KURODA, K. MASAI, K. MATSUOKA,
A. MOHRI, S. MORITA, M. MUGISHIMA, N. NODA, I. OGAWA,
Y. OGAWA, Y. OKA, M. SAKAMOTO, K. SAKURAI, M. SASAO,
K. SATO,
K.N. SATO,
S. TANAHASHI,
Y. TANIGUCHI,
T. TETSUKA, K. TOI, K. YAMAZAKI
Institute of Plasma Physics,
Nagoya University, Nagoya,Japan
Abstract
EXPERIMENTS ON ICRF HEATING AND FAST WAVE CURRENT DRIVE IN JIPP T-IIU.
The paper summarizes two years' progress on ICRF heating and current drive experiments in
JIPP T-IIU. In the heating experiments, a new ion Bernstein wave heating regime called Mode II was
examined; efficient electron heating was observed. In high power ICRF heating experiments, a five element antenna array was used, improving the quality of the heating. High power ICRF heating(2 MW-m" 1 ) was successfully utilized by suppressing high Z impurity contamination, using a carbonization technique. The energy confinement time scaling has been established, with moderate degradation compared to that of neutral beam injection. Current drive was demonstrated by injection of thefast waves in the lower hybrid frequency range. A four element dipole antenna array was used, and acurrent drive efficiency as high as that of slow wave current drive was achieved. The fast wave wasapplied to achieve stabilization of them = 2/n = 1 tearing mode, causing suppression of major disruptions in various conditions. In another effort undertaken to find a density limit free current drive method,electron beam injection was also examined.

1.

ION BERNSTEIN WAVE HEATING (MODE II)

At the previous IAEA Conference (1984), a success in ion Bernstein wave heating was reported (Mode I) when a layer with w = 3COCD (or u> = (3/2)ooCH) wasestablished at the central plasma vertical chord [1]. Unexpectedly, a half integer
cyclotron harmonic heating process was found to exist. A new regime called ModeII is dealt with in this paper where the oi - wCH layer is aligned on the central
On leave from Max-Planck-Institut fur Plasmaphysik, Garching, Fed. Rep. Germany.

495

496

WATARI et al.

0 . 5 -

- 2 0

- 1 0

0

M i n o r

R a d i u s

10

20

(cm)

FIG. 1. Electron temperature profile changes for varying phasing angle.

chord. The outstanding feature observed here is a rapid rise in the electron temperature just after the RF power has been turned on, followed by a gradual rise in theion temperature [2]. This constitutes a pronounced difference with respect to theMode I experiment. Through a power balance analysis of the experimental data, theconclusion is reached that the input RF power is mostly absorbed by electrons.

2.

CONFINEMENT STUDY IN HIGH POWER ICRF HEATED PLASMAS

A five element antenna array was installed for k r tailoring on which the qualityof heating may depend. One expectation was the reduction of impurity content at
Ap = 180°, associated with the co-axial mode. In the experiment, however, changes
are observed neither in the line radiation from impurity ions nor in the spatiallyresolved bolometer signals. However, the heating efficiency clearly depends on thephasing angle. The central electron temperature shows maximum rise and a peaked
profile for A<p = 180° (Fig. 1). The neutron yield and the ion temperature also rise
to maximum values for A<p = 180°. The stored energy as determined from magnetic
measurements supports these kinetic data. A computer code developed by Fukuyama
et al. [3] is used to study the wave propagation and the power deposition profile. A
wave with large k| tends to be absorbed centrally and is produced whenA<p = 180°. Therefore, the observed effects are understood in the framework ofwave physics as kj is controlled through the phasing of the antenna.
JIPP T-IIU high power experiments have been applying new methods of impurity control such as carbon limiters, synchronized gas puffing, and a second current
rise [4]. Their effectiveness was, however, decreasing with increased injection power

497

IAEA-CN-47/F-I-5
2.0

0.0
0.0

0.5

1.0
PRF (MH)

1.5

2.0

F/G. 2. Radiation loss versus injected RF power (open circles: before carbon coating; solid circles:
after carbon coating).

(2 MW max.), i.e. with high power density (2 MW-cm 3) which has not been
attained elsewhere. A carbonization technique [5] was then employed and found tobe very effective in impurity suppression (Fig. 2). Fe lines, which have been themain radiation loss channel in the hot core region, were reduced in intensity by anorder of magnitude.
The key question is whether or not ICRF heated plasmas show degradation ofconfinement as do NBI heated plasmas. The parametric dependences of TE are
studied by varying Ip from 200 kA to 320 kA and PRF up to 2 MW with ne fixedaround 7 X 1013 cm" 3 (Fig. 3(a)). There is a degradation of confinement withincreasing power and a marked improvement with increasing IP. A power law fit to
the data yields a scaling TE OC Ip P~f with a ~ 1 and 13 « 0.42. This is generally
in good agreement with Kaye-Goldston scaling [6]; the encouraging feature of thisresult is its indication of milder degradation than that due to Kaye-Goldston scaling.
An offset linear plot [7] is likewise suitable to the same data set (Fig. 3(b)). The
offset linear TE is determined to be 8 ms. While xE, from this definition, predicts amilder degradation with increasing power, there is the problem that it may not be
improved by increasing Ip. With TE values provided by various machines of different
size, the JIPP T-IIU data contribute to the determination of the rE scaling to theplasma volume. The offset TE seems to have a V * dependence on the machine sizeas Kaye-Goldston scaling has with constant q.
Based on precise measurements of Te(r), n,,(r) and sawtooth activity, the power
deposition profile P(r) and the electron heat diffusivity Xe(r) have been studied. The
deposition profile is important because not only may it deeply affect the confinement

498

WATARI et al.

2.0

2.5

P 0 H + P R F (MW)FIG. 3. Confinement time TE of ICRF heated plasma: (a) power law fit and (b) offset linear fit;I » 280 kAand ne ~ 7 x 1OU cm'3.

properties but also provides an interesting interface between experiment and wave
physics. It is found to be a peaked deposition profile for medium density plasmas( S 5 x 1013 cm" 3 ) and a broader profile in the higher density cases(~(7-8) x 1013 cm" 3 ). The results agree qualitatively with the analysis from the
RF power deposition code [3].

3.

FAST WAVE CURRENT DRIVE EXPERIMENTS

Attention has been focused on fast wave current drive [8-10], because of its
attractiveness as a current drive method which may be utilized in high temperatureand density plasmas.
The fast wave current drive is studied experimentally by using a four element
dipole antenna with a double Faraday shield. The RF frequency is 800 MHz, andBT is 2.6 to 2.96 T. The antenna is designed to produce only Ey and Bz fields sothat, preferentially, the fast wave is excited/ Global reflections of the plasma loaded
antenna show a minimum at A<p = 0°, where A<p is the phase difference between the
dipoles. The reflections decrease also with increasing n,, and by moving the antennacloser to the plasma. These types of behaviour are those of a fast wave, where adecrease in the evanescent region improves the loading characteristics. The pump RF

IAEA-CN-47/F-I-5
i
80
40

1

499
1

1
---

^

1 1

i

I

I

1

1

|
1

I
I

i
i

1
1

I

0
-1
0.6

.1 0 /
3"

1

3.

J
200
5 0

0

80
0

1

i

|

i

1

i

1
1

A*
/,
_V

^—'

1

{
K
"•

^

^\, I

i

i

I

i

1

i

0.8

fJ-

i^

i
i
~
80-

-

1

i

1
1

-

1
i

_

1
1
i
100
200
Time f (ms)-I

FIG. 4. Time behaviour of plasma parameters for injection of fast waves: plasma current, lp, loop
voltage, VL, line averaged electron density, ne, microwave emission near the second electron cyclotron
harmonic, [fCe, integrated counts of X-ray (10-300 keV), Nx, peripheral electron temperature deduced
from the ratio of the O V lines, Tep, central electron temperature, Te0 and RF net power, PRF.

amplitude is monitored with an electrostatic RF probe in the scrapeoff region, 90°
toroidally from the antenna. The pump RF field, normalized to the injected RF
power, increases by about 15 dB as Ay? is shifted from 180° to 0°. This is ascribedto the increased portion of excited waves not satisfying the accessibility condition.
When 80 kW of net RF power, PRF, is applied with A<p = 0°, a plasma
current Ip of about 50 kA is sustained with a negative loop voltage at constantfie = 3 x 1012 cm" 3 (Fig. 4). The efficiency of current drive by fast waves,
7? = IpneR/PRF, is 1.7 x 10" 2 (1020 kA-m- 2 -kW- 1 ), which is as high as that ofslow waves in the JIPP T-IIU tokamak.
In the low OH current discharge, a density limit (ne = 8 x 1012 cm" 3 ) is
observed, contrary to theoretical expectations based on linear wave propagation. We
conclude that the disappearance of the slide-away electron tail which exists before
applying RF and the power flow to the ion tail via a parametric instability resulting
from mode conversion must be responsible for the density limit [10].

500

WATAR1 et al.

80 120 160 200
Time(ms)

80 120 160 200
Time(ms)

FIG. 5. Discharge characteristics of RF injected plasma in the high density regime where ii>w
(r = 0) > o>.

When fast waves are injected into a high OH current plasma, the hard X-ray
emission in the energy range higher than 30 keV is strongly enhanced in the high density regime wLH(r = 0) > w (Fig. 5). In this discharge, an enhanced electron cyclotron emission is also observed, while the bulk electron temperature (Thomson
scattering) remains constant or decreases slightly. The hard X-ray emission is sharply
localized near the plasma centre. These results indicate that fast waves propagate into
the high density plasma and interact intensely with the high energy electrons which
exist already in a plasma with high OH current. This result is quite encouraging, inthe sense that the density limit encountered in the low current plasmas can be overcome if high energy seed electrons are amply provided by slow waves and/or electron
cyclotron waves, in place of an inductive electric field.
The RF current drive has a large potential for controlling the current density
profile, which governs tearing mode stability and confinement properties. The suppression of m = 2 tearing mode and major disruptions was studied using RF current
drive and/or production of high energy electrons by fast waves [11]. Stabilizing
effects are found in the following regime: q(a) = 2.5-3.5, B, = 1.2-2.4 T and
fie = 0.5-1.5 x 1013 cm" 3 . The RF power required to suppress the major disruption is about 20 to 40% of the Ohmic input power, which is close to the theoreticalpredictions [12].

4.

ELECTRON BEAM INJECTION (EBI)

A current drive experiment with EBI has been performed by using a newly
developed injector working for a long duration ( > 4 ms). Here, the 'plasma anodemethod' was used to generate the electron beam. The parameters of the target plasma

IAEA-CN-47/F-I-5

501

were B, = 2 T, ne = 1.5 x 10 l3 cm" 3 and Ip = 80 kA. When an electron beam ofcurrent I,. = 240 A at an energy 1.1 keV is injected, I p rises quickly by 4 kA [13].
The driven current remains constant for 4 ms, and the multiplication factor, AIp/Ik,
is about 20.

ACKNOWLEDGEMENTS
The authors thank Dr. Y. Sakamoto and all collaborators who contributed to thesuccess of the carbon coating experiment.

REFERENCES
[1]
[2]
[3]
[4]
[5]
[6]
[7]
[8]
[9]
[10]
[11]
[12]
[13]

ONO, M., et al., Phys. Rev. Lett. 54 (1985) 2239.
KAWAHATA, K., et al., in Controlled Fusion and Plasma Heating (Proc. 13th Europ. Conf.
Schliersee, 1986), Vol. 10c, Part II, European Physical Society (1986) 169.
FUKUYAMA, A., et al., Nucl. Fusion 23 (1983) 1005.
TOI, K., et al., in Plasma Physics and Controlled Nuclear Fusion Research (Proc. 10th Int. Conf.
London, 1984), Vol. 1, IAEA, Vienna (1985) 532.
NODA, N., et al., Jpn. J. Appl. Phys. 25 (1986) L397.
KAYE, S.M., GOLDSTON, R.J., Nucl. Fusion 25 (1985) 65.
ODAJIMA, K., et al., to be published in Phys. Rev. Lett.
GOREE, G., et al., Phys. Rev. Lett. 55 (1986) 1669.
McWILLIAMS, R., et al., Phys. Rev. Lett. 56 (1986) 835.
OHKUBO, K., et al., Phys. Rev. Lett. 56 (1986) 2040.
TOI, K., et al., in Controlled Fusion and Plasma Heating (Proc.13th Europ. Conf. Schliersee,1986), Vol. 10c, Part II, European Physical Society (1986) 457.
RUTHERFORD, P.H., Rep. PPPL-2277, Plasma Physics Lab., Princeton Univ., NJ (1985).
MOHRI, A., et al., in High-Power Particle Beams (Proc. 6th Int. Conf. Kobe, 1986) P-A-l.

DISCUSSION
J.-M. NOTERDAEME: Was there any indication in the phasing experimentsthat radiation losses depend on phasing as the carbonization wears off, or in an uncarbonized machine?
T. WATARI: We carried out our phasing experiment under three sets of conditions, namely before carbon coating, after carbon coating and after removal of thecoating. In all three cases, the impurity contents were insensitive to the phasing
angle.

IAEA-CN-47/F-IM

ELECTRON HEATING AND MHD MODES
DURING LOWER HYBRID CURRENT DRIVE
ONPLTS. BERNABEI, R. BELL, A. CAVALLO, T.K. CHU,
P. COLESTOCK, J. FELT, W. HOOKE, J. HOSEA,
F.C. JOBES, E. MAZZUCATO,

E. MESERVEY, R. MOTLEY,

J.A. MURPHY, M. ONO, R. PINSKER, J. STEVENS,
S. VON GOELER, J.R. WILSON
Princeton Plasma Physics Laboratory,
Princeton University,
Princeton, New Jersey,
United States of America

Abstract
ELECTRON HEATING AND MHD MODES DURING LOWER HYBRID CURRENT DRIVEON PLT.
Lower hybrid waves of up to 1 MW power from a 2.45 GHz source were applied to ohmic discharges in the PLT tokamak. Highly efficient current drive (Ifl]3 R/P « 1.4) was obtained from long,multi-element grills (8 and 16 element) producing a narrow, high velocity wave spectrum. Above a certain power level, dependent on the plasma density, the lower hybrid waves suppress sawtooth oscillations. Further increase in power level to about 500 kW (the original ohmic power level) and beyond
suppresses the m = 1 oscillation at fie » 1013 cm"3 and results in strong electron heating in the coreregion (Te increases from 2 to 4-6 keV). The m = 2 oscillation does not manifest itself unless thelimiter q is reduced to below 2.5. Enhanced sawtooth amplitude driven by up to 1 MW of ICRF heating(D- 3 He minority) can also be suppressed by the lower hybrid waves. Preliminary experiments to drive
current with fast waves by a phased loop array are described. These experiments, however, did notexhibit efficient current drive effects.

1.

INTRODUCTION

The
applicationoflower
hybridwaves
totokamakdischarges in the past few years has led to a variety of new,incompletely understood phenomena, including current drive,
sawteeth
suppression, and strong electron heating[1-5].
These
effectsarethe resultofthe formationof asuprathermal electron tail by the RF waves which creates a new
tokamak equilibrium in which the current distribution and theelectron temperature distribution are uncoupled.
The PLTexperimenthas been
devotedto the studyof this new
equilibrium, which is still largely uncharted because of thelarge number of free parameters, such as the direction and503

BERNABEI et al.

504
1.5

2.45 GHz
I6-Waveguide Grill

'E

0.5

0.5

1.0

1.5

n c (x|0 l3 cm" 3 )FIG. 1. Current drive figure of merit, In^R/P, versus electron density, for 800 MHz, 6-element,
wide and narrow grills, and 2.45 GHz, 8-element and 16-element grills.

phase of the lower hybrid waves, the intensity of the induced
electric field in the plasma and the conditions of the target
plasma.
2.

THE EXPERIMENTThe PLT Tokamak, with major and minor radii of R = 132 cm
and a = 40 cm, has been operated with a toroidal field B = 2032 kG, a plasma current I of 300-700 kA (2 < q L < 6) in thedensity range fie = 0.2-4 x 10
cm . The RF system consistsof three 2.45 GHz klystrons capable of delivering 1 MW ofpower to the plasma. The waves are coupled to the plasma bymeans of two grills, one an 8-waveguide array positioned on
the top and the other a 16-element array on the outside of thetorus.
The principal diagnostics include multi-point laserThomson scattering for T , soft and hard X-rays, 2JL,
radiation, and Mirnov loops.
3.

RESULTS

CURRENT DRIVE. Comparison of results with the present
2.45 GHz system with those of the past 800 MHz, 6-waveguidegrill [6] enable us to check the scaling of the density limit
and the effect of the wave spectrum. On the first point, thedensity limit at 2.45 GHz is ~ 4 :> 1 0 1 3 cm" 3 , at least 6 times
greater than at 800 MHz.
As shown in Fig. 1, the current
drive figure of merit In., jR/P is greatest for the higher
frequency
systemlaunching
the fastest waves with the

505

IAEA-CN-47/F-II-l
0.4

1

1

1

1

, m =0 at r = 1 cm^/
(Sawtooth);o.3/

fc r =-5cm

-

+ 6cm

UJ

80.1

<f
c

\

1

200

\

\i

\
400
Prf(kW)

1

1 .
600

1

..I

.

800

FIG. 2. Mode amplitude of m = 0 (sawtooth) at r = 0, and of m = 1 at r = -5 cm and 6 cm, asa function of lower hybrid power as determined from 2Qce emission; ne = 1013 cm3, B = 29 kG,
I = 500 kA.

narrowest wave spectrum (6 WG, An.. = 1.5 and 2.7; 8 WG, A n =1.4, and 16 W G , A n = 0.5). In a l l cases the phase velocity
w a s optimized for maximum efficiency.
T h e decrease inefficiency with density is probably due to diminished wave
accessibility to the center of the torus. Experiments aboven g = 1 x 10
cm
show that the efficiency improves withincreasing plasma current and electron temperature.
STABILITY. If the current drive efficiency is h i g h , lower
hybrid power can strongly affect the M H D modes of the tokamak
plasma. Above a certain power threshold, which depends on thed e n s i t y , the R F waves suppress sawteeth activity a t the centerof the plasma while the off-center m=1 mode becomes d o m i n a n t ,as shown in Fig. 2.
Further increase in the R F power,
however, leads to suppression of the m=1 mode a l s o .
ELECTRON HEATING. In spite of the broadening of thecurrent profile, as determined from the change of internal
inductance,thecentralelectrontemperature
rises
dramatically, forming a peaked temperature profile apparently
decoupled from the current profile (Fig. 3 ) . The stability
characteristics of this type of discharge are still largely
unexplored by theorists.
As shown in Fig. 4, the temperature rise continues u p to
~ 500 kW, and then appears to saturate. Stored energy for theelectrons increases with power, b u t a t a rate less than
expected from ohmic d i s c h a r g e s .
The temperature increase
becomes smaller as the plasma density rises, as shown in
Fig. 5. Above n e = 3 x 1 0 1 3 c m " 3 the available RF power isinsufficient to eliminate the sawteeth oscillations.

BERNABEI et al.

506

-40

0

20

RADIUS (cm)FIG. 3. Temperature profiles during ohmic discharges (lower curve) and lower hybrid discharges
measured by laser Thomson scattering. B = 30 kG, I = 500 kA and ne = 1013 cnf3. (The Te profiles
are measured along a fixed vertical chord which cannot follow any outward shift of the axis.)

1

1

I

1

1

i '

-

-

.'i

1

200

400

600

1

1

800

1

1000

FIG. 4. Laser Thomson scattering measurements of TJO) with RF power input. ne ~ 1&3 cm'3,
B = 30 kG, I ~ 500 kA.

-

1 •}

—3

The electron energy confinement time at n g = 10cm
rises initially with RF power and then slowly decreases to theohmic level at an RF power level comparable to the ohmic
power.
The electron energy confinement time, which is
proportional to n e for ohmic discharges for n e < 3 x 10
cm" , increases with rie at a rate of ~ 1/3 of that of theohmic discharge during lower hybrid current drive.
During operation with q L > 2.5, the RF power suppresses
the m=0 and m=1 modes without driving up the m=2 mode or
causing a disruption.
As q L is reduced below 2.5, the m=2
mode tends to grow in bursts during the RF pulse. Operationnear q L = 2 results in strong m=2 activity and disruptions inboth ohmic and RF drive.
Ion cyclotron minority heating (D- He) at 1 MW increases
the sawteeth amplitude from 0.20 to 0.5 keV because of the

IAEA-CN-47/F-II-l507

Prf=770kW-60° •-75° »o 2ne (ICTcm"3)

FIG. 5. Electron temperature measured with soft X-rays before and during lower hybrid heating
at 770 kW.

central power deposition and electron- He coupling [7]. The
application of lower hybrid wave power suppresses theseenhancedsawteeth;the minimum
lower hybrid power for
suppression is almost independent of ICRF power (up to 1 M W ) .
FAST WAVE. A 6-element loop array has been installed on
PLT to launch the fast wave at 800 MHz, to determine whether
current drive is possible beyond the normal density limit.
Measureable currents have been driven with up to 80 kW ofpower, but this power level is too low to establish thedensity scaling of the current drive. The dependence of theloading of the antenna on density and the appearance ofparametric decay spectra at n e > 1 x 1 0 1 3 cm" 3 are similar to
those seen previously with the slow wave coupler, so that theexcitation of the fast wave is in doubt. A 3 x 4 dielectric
waveguide is now being installed on PLT to enable directcomparison of the slow and fast wave characteristics with abetter defined launching configuration.

REFERENCES
[1]

[2]

GORMEZANO, C , BLANC, P., BOTTOLLIER, P., et al., "Experimental studies of lower
hybrid heating and current drive on the Petula-B tokamak", Radiofrequency Plasma Heating(Proc. 6th Top. Conf. Callaway Gardens, GA, 1985), American Institute of Physics, New York(1985) 111.
CHU, T.K., BELL, R., BERNABEI, S., et al., ibid., p. 131.

508
[3]
[4]
[5]

[6]
[7]

BERNABEI et al.
CHU, T.K., BELL, R., BERNABEI, S., et al., Nucl. Fusion 26 (1986) 666.
SOLDNER, F., McCORMICK, K., ECKHARTT, D., et al., Phys. Rev. Lett. 57 (1986) 1137.
HOSEA, J., WILSON, J.R., HOOKE, W., et al., "RF experiments on PLT", Controlled Fusion
and Plasma Heating (Proc. 13th Europ. Conf. Schliersee, 1986), European Physical Society(1986).
BERNABEI, S., DAUGHNEY, C , EFTHIMION, E., et al., Phys. Rev. Lett. 49 (1982) 1255.
MAZZUCATO, E., BELL, R., BITTER, M., et al., in Plasma Physics and Controlled Nuclear
Fusion Research 1984 (Proc. 10th Int. Conf. London, 1984), Vol. 1, IAEA, Vienna (1985) 433.

IAEA-CN-47/F-II-2

RADIOFREQUENCY HEATING AND
CURRENT DRIVE EXPERIMENTS
ON ALCATOR C AND VERSATOR II
M. PORKOLAB, P. BONOLI, Kuoin CHEN, C. FIORE,
R. GRANETZ, D. GRIFFIN, D. GWINN, S. KNOWLTON,
B. LIPSCHULTZ, S.C. LUCKHARDT, E. MARMAR,
M. MAYBERRY1, F.S. McDERMOTT, J. MOODY,
R. PARKER, J. RICE, Y. TAKASE, J. TERRY,S. TEXTER2, S. WOLFE
Plasma Fusion Center,
Massachusetts Institute of Technology,
Cambridge, Massachusetts,
United States of America
Abstract
RADIOFREQUENCY HEATING AND CURRENT DRIVE EXPERIMENTS ON ALCATOR C
AND VERSATOR II.
Lower hybrid heating and current drive experiments on the Alcator C tokamak (R = 0.64 m,
a = 0.165 m, molybdenum Iimiters) were performed at a frequency of 4.6 GHz, with net injected RF
powers up to P rf < 1.5 MW. Recent experiments have focused on energy confinement studies in lower
hybrid current driven (LHCD) and LHRF heated Ohmic discharges, and sawtooth stabilization in combined LHCD-OH driven discharges at densities ne < 1.4 x 1020 m~3. Ion Bernstein wave heating
experiments were also carried out in Alcator C, at a frequency of f = 183 MHz and at power levels
P rf s 200 kW. Significant heating (AT; s 400 eV) was observed at o/uCH - 1.5, 2.5, ando>/«CD = 2.5, at densities ne = 1 x 1020 m"3. In the Versator II tokamak, particle confinement
improvement (by factors of about two) was observed in the presence of 2.45 GHz lower hybrid current
drive.

Energy Confinement in LHCD Plasmas in Alcator C
The energy confinement properties of lower hybrid current driven(LHCD) plasmas, heated and maintained entirely by rf-driven currents
have been investigated in the parameter range ne = 0.1 — 0.8 x 1020
m - 3 , / p = 100-200 kA, BT = 8 - 1 1 T, qe = 8-17,Zeff= 1.5-2.2, hydrogen gas [1]. Electron temperature profile measurements were obtainedwith a 5—channel ruby Thomson scattering system and ion temperaturesby neutral charge-exchange analysis. The energy of the current carrying
1
2

GA Technologies Inc., La Jolla, CA, USA.
TRW Inc., Los Angeles, CA, USA.

509

510

PORKOLAB et al.

superthermal electron population was inferred from plasma hard x-ray
spectral analysis and measurements of the equilibrium quantity j3p + ^ / 2 .
Confinement measurements from LHCD plasmas were compared with very
similar ohmic discharges. The bulk thermal energy content of LHCD
plasma was nearly identical to that of a similar ohmic plasma, while theenergy content of the superthermal electron tail in LHCD discharges could
account for up to half the total kinetic energy of the plasma at the lower
range of densities. The global confinement time, defined as TE = Wiot/Pim
where Wtot is the total kinetic energy of the bulk and the superthermal
particles (LHCD plasmas), and Pj n is the net input rf or ohmic power,
is plotted vs. density in Fig. 1. Over this range of densities, the ohmic
power remained constant, and the ohmic confinement time rises linearlywith density. The rf power required to maintain the rf driven discharge
increases nearly linearly with density [2]. The quantity r'E = WBuik/Pin,where WBUUC is the thermal energy of the bulk LHCD plasma, is also plotted. The difference between T'E and TE of the LHCD plasmas indicatesthe relative importance of the electron tail in the overall energy balance.
At densities ne <, 0.3 x 1020 m~ 3 , the confinement time of LHCD plasmas
is somewhat greater than that of ohmic plasmas, due to the substantial
fraction of energy carried by the superthermal electron distribution. At
higher densities, the LHCD confinement time degrades relative to that of
the ohmic plasmas. The experimental scaling of the LHCD confinement
time with density and input power is comparable with the predictions of
the empirical Kaye-Goldston scaling law for NBI-heated plasmas [3]. Suchan agreement suggests that similar agents may be affecting the transportof different types of auxiliary heated plasmas, including ones purely maintained by LHCD. Combined ray tracing and transport code studies [4]indicate that the present results at high densities could be explained by adegradation of %e (bulk) relative to its ohmic value [1].
The confinement time of energetic tail electrons has been estimated by
combining measurements of hard x-ray spectra and theoretical calculationsof the absorbed power provided by the code [5]. After collisional losses
are subtracted, the confinement time of the fast electrons above 100 keV
is found to be comparable to the global energy confinement time andincreases with energy.
Lower Hybrid Electron Heating Experiments in Alcator C
Heating experiments have been carried out in ohmically maintained
deuterium plasmas for BT - 5.5 - 8.0 T, Ip = 150 - 375 kA, n e <2.0 X 1020 m" 3 , using molybdenum limiters. In previous experiments using
SiC coated graphite limiters, significant electron (AT eo ~ l.OkeV) and ion

511

IAEA-CN-47/F-II-2
Prf (kW)
200

300

LHCD

T

A LHCD

T

600

900

1

1

E

1 0 -" & OH TE

E

A ^
A
A

i

\>V I

i

5 -

K-G

:

^ J

i

10

ne (10 19 nr 3 )FIG. 1. TE versus ne for LHCD (solid circles) and Ohmic (open triangles) hydrogen plasmas,
for BT = 8 T, Ip = 140 kA. T'E = WBull/Pin (solid triangles) for the LHCD plasmas. Vie dashedline represents the Kaye-Goldston scaling law.

(AT;O <, 0.7A;ey) heating was observed [6]. However, these results wereaccompanied by rather large rises in Zeff (AZeff <, 3). Bulk electron
heating due to electron Landau damping and subsequent ion heating due
to collisional relaxation are observed in the present experiments over thedensity range n e = 0.6 — 1.7 X 1020 m~ 3 . Temperature increases of
ATeo = 0.7 keV and ATio = 0.3 keV are recorded for Prf = 1 . 0 MW
at BT = 5.5 T, Ip = 260 kA, ne = 1.4 x 1020 m~ 3 with AZeff < 0.5[7]. As seen in Fig. 2, comparable heating efficiencies are obtained withwaveguide phasings of A(j> = 90° (current drive) and 180° (heating mode),and the electron temperature profiles are similar to those in ohmic plasmas
in both cases. Figure 2 shows that the energy confinement in LH-heatedplasmas is degraded relative to that in ohmic discharges. As in the caseof LHCD plasma, the confinement scaling with rf power is consistent with
the Kaye-Goldston formula [3]. In the current drive mode sawteeth werestabilized [7, 8]; however, no significant improvement in the global energy
confinement is observed when the sawteeth are suppressed. The radiatedpower estimated by spectroscopic means increased with rf power but waslimited to Prad ~ 150kW. Bolometric measurements show a similar trend
but higher radiated power levels. However, radiation alone cannot explain
the deterioration of rjg with power. Numerical code modelling [4] indicatesthat 75% of the injected rf power is absorbed by electron Landau damping,PORKOLAB et al.

512

O-OH
• -RF

%

A<J>=115

B = 5.5 T
1=250 kA
• XiO^m3
,0

V-OH
T-RF
5 '

-

B = 6.2 T
1=250 kA
n. = 1.1X1
0.5

1.0

1.5

2.0

Plot
/7G. 2. Energy confinement times in OH driven plasmas with LH heating mode (A0 = 180°) and with
LH current drive phasing (A<t> = 90°).

with the remainder being lost to collisional absorption near the plasmaedge. Modelling indicates that a 30% degradation of Xe, the electron
thermal diffusivity, is sufficient to explain the present results.
Sawtooth Stabilization with LH Current Drive in Alcator C
Stabilizing sawteeth by means of rf current drive has been studied
in Alcator plasmas at higher densities (n e = 0.6 — 1.2 X 10 20 rn~ 3 ,/ p =250 - 325 kA, qt = 4 - 6 ) than in other tokamaks [9-11]. The major
diagnostic of the sawtooth behavior is a 16-channel soft x-ray diode arraysensitive in the range 1.5 to 15 keV. The sawtooth period is found tolengthen during lower hybrid wave injection, and with sufficient rf power(PrfH £ 550 kW for ne = 1.1 x 1020 m ~3,IP = 250 kA, BT = 6.2
T) the sawteeth are entirely suppressed [7, 8]. Sawtooth suppression has
only been achieved in the current drive mode of operation, i.e. with therelative phase between waveguides of A<f> = +90°; stabilization is notobserved with symmetric (A(f> — 180°) or anti-current drive (A^> = —90°)
phasings for the range of rf powers used in this experiment. It has been
noted that as the rf power is increased above Prf « 900A;W\ the sawteeth
recur after 40-50ms, which is comparable to the characteristic L/R time[7]. In contrast to stabilization experiments on low density plasmas in
ASDEX and PLT [9, 10], in the current drive mode (A<j> = 90°) theIAEA-CN-47/F-II-2

513

0J5

0.10-

0.05-

FIG. 3. Change in A (= 0 p + f/2-/) versus rf power for A<t> = +90° (triangles) and A<f>= -90°(circles); BT = 6.0 T, he = 0.8 x 1020 m3, Ip = 275 kA, q(a) = 4.7.

sawtooth period increases monotonically (by up to a factor of two) with rf
power until stabilization is achieved. In all cases, the sawtooth inversion
radius inferred from the x-ray measurements remains at r/a ~ 0.2 at all rf
powers for which sawteeth are observed. For A<f> = +90° at rf power levels
near the threshold, high frequency oscillations are often superimposed on
the x-ray sawteeth. In these circumstances, the sawtooth crash times,
as observed on the central soft x-ray channel, can be as long as 1 msec
(compared to the usual crash times of r <, lOO/nsec in both ohmic and LH
heated plasmas). The m — 1 oscillations are sometimes observed in fullystabilized discharges as well. As shown in Fig. 2, equally efficient electron
and ion heating are observed for both A<f> — 90° and 180°, as well as forAcf> — -90° (not shown).
The change in fip + £i/2 versus rf power has been measured and isshown in Fig. 3. During rf injection, (3p + £{/2 increases by 0.11 for
A<j> = +90° but only by 0.04 for A<j> = —90°, in contrast to results fromthe low density stabilization experiments on ASDEX [9]. From the kinetic
measurements, the thermal f3p is estimated to increase by 0.04 — 0.05 in
both cases. A likely explanation of the difference in the measured valuesof (3p-\-£i/2 between the two phasings is an increase in the plasma internal
inductance for the current-drive case. This conclusion is supported bycomparison of the time evolution of j3p + £{/2 for the two phases. While/5p + £{/2 increases for both phases, the slowest time evolution, whichmost likely represents the change in the internal inductance, is negative
for A<£ = —90° and positive for A<f> — +90°. A possible explanation of thisbehavior is that rf current is driven at r/a fa 0.2 — 0.4, outside the q — 1
surface. Consequently, the current profile shrinks in the outer regions of514

PORKOLAB et al.

the plasma (r/a ~ 0.6 — 0.8), resulting in an increased inductance, but isflattened near the center such that q(r) > 1 everywhere.

Ion Bernstein Wave Heating E x p e r i m e n t s in Alcator C
Ion Bernstein wave (IBW) heating experiments have been carried out
in Alcator C in order to test the effectiveness of this technique for heating high density plasmas (n, > 1 x 10 2O m~ 3 ). Efficient ion heating with
IBW has been reported at lower plasma densities in tokamaks by Ono andcoworkers [12]. However, the heating mechanism remained somewhat controversial and,in addition to minority heating, nonlinear absorption mechanisms have also been advocated to explain the observed results [13,14].
The present work indicates that both nonlinear absorption and improved
ion energy confinement could play a significant role in the ion heatingprocess.
Ion Bernstein waves were launched by a low field side, center fed,
moveable T-shaped loop antenna, with the center conductor aligned parallel to the toroidal magnetic field, and surrounded by a double layered
Faraday shield. The outer dimensions of the stainless steel antenna are:width <, 0.04 m, length <, 0.25 m, and maximum power transmitted Prf <,200 kW at f = 183 MHz. Due to significant metallic impurity generation
at high powers, most of the data were obtained at P <, 150 kW. The experiments were carried out in three different plasma configurations, withthe minor radius (set by molybdenum limiters) set at a — 0.115 m, 0.120
m or 0.125 m (4.8 < B(T) < 11.0). Optimal coupling was obtained withthe front surface of the Faraday shield being Ar ~ 5 mm behind thelimiter surface. Although heating studies were carried out in a rangeof magnetic fields, here we shall concentrate on discussing three regimes
corresponding to magnetic fields B(0) ~ 5, 7.5, and 9.3-10.3T, such thatl{$ = W / W C H ( 0 ) ^ 2.5, 1.5 or 1.25 (and hence 6£ = w / w c o ( 0 ) ~ 5, 3,
and 2.5 whenever deuterium minority is injected into the hydrogen majority plasma) where £fP(££) designates the hydrogen (deuterium) harmonic
numbers in the plasma interior near r = 0. It should be noted that for£$ = 2.5 the resonant layers £H — 2, 3 are just outside of the plasma
column. In addition, by adjusting the magnetic field the tjj = 3 layer
was moved behind the antenna. In the case of £H = 1.5, both the £u= 1,2 resonances were outside of the plasma column (and again behind
the antenna). However, for the 1$ = 1.25 (iff = 2.5) case, the £H =
1.5 (or ZD — 3) layer could have been in front of the antenna, and in fact
in many successful heating shots it was located 1-2 cm inside the plasma
column. Under such circumstances linear and nonlinear absorption of the

IAEA-CN-47/F-II-2

0515

100 200 300 400 500 600

time (ms)FIG. 4. Time sequence of a typical shot with rf injection in the IBW regime, B = 7.5 T, I = 240 kA.

incoming IBW wave at the plasma edge should be complete. Nevertheless,even in such cases efficient central bulk hydrogen and deuterium minority
ion heating was observed.
A typical shot during IBW launching at B=7.6T is shown in Fig.4.
The hydrogen bulk temperature rises by an amount ATJJ ~ 300 eV for
Prf = 125kW as measured by standard mass-resolving charge exchange
neutral analysis. The density rises by 30%, the loop voltage, Zeff andthe electron temperature remain nearly constant. The ion temperature
profile is centrally peaked, and is quite similar to the initial ohmic profile.
Plots of the increase in the bulk hydrogen temperature versus RF
power at three different values of the magnetic field (corresponding to thethree regimes discussed before) are shown in Fig.5. The most efficient
ion heating is recorded for the case of B = 9.3 T (or ($ = 1.25, £^= 2.5). The typical range of densities corresponding to these data weren e ~ (0.7 — 1.1) x 10 2o m~ 3 . Typical heating rates corresponding to thesecases are AT(eV)/P(kW)~ 2.2-3.3 at a density of ne ~ 1.0 X lO 20 ™" 3 . In
all cases a threshold power is apparent before heating is observed (namely
PtK ~ 20 kW for £^ ) = 2.5, Pth ~ 35 kW for ^ = 1.5, and Pth < 10 kW
for *g> = 1.25).
In general, significant heating at any magnetic field is observed only
at densities n e <, 1.5 x 102Om~3. This is demonstrated in Fig.6, where
heating data are shown at B = 7.6 T. Note that no heating was observed
above the density limit ne ~ 1.5 X 10 20 m~ 3 . Measurements indicate thatat B = 5T (£H = 2.5, £D = 5),even when a deuterium minority species

516

PORKOLAB et al.
500

400

-*

300

200

100

40

60

80

100

120

140

160

PRF(kW)
FIG. 5. AT( versus P^ for three values of B. At B = 9.3 T, lH(0) = 1.25; at B = 7.5 T,(H(0) = 1.50; at B = 5 T, (H(0) = 2.5, ne = (0.7- 1.1) x 1020 m3.

is present, Tp ~ T/j. In particular, the hydrogen ion temperature riseis independent of the presence of deuterium, indicating that nonlinear
absorption is operative in this regime [14]. However, to some extent at
tH = 1.5, and especially at £H = 1.25 the deuterium temperature (withtypical concentrations of 0.5-10%) may achieve significantly higher values than the hydrogen majority component. For example, at B = 9.3T,TD £ 3.5 keV while TH <, 1.5 keV during RF injection. A radial scan of
the charge exchange analyzer indicates that the deuterium component isalways peaking on axis, decays slowly in time, and that the central hydrogen and deuterium heating maximize at the same time. It should be
noted that at n e > 1.5 x 102Om~3, T e often decreased owing to increasing
values of radiation. Nevertheless, Ti does not decrease, and the electron
to ion temperature ratio may approach unity (in contrast to pure ohmic
discharges where Te > Ti). This phenomena is observed at densities up
to n e ~ 2 x 10 2 0 m- 3 at both B = 7.5 T (iff = 1.5) and B = 9.3 T (*g}= 1.25). These results clearly indicate the importance of direct rf heatingof ions (as long as Te > Tj, collisional heating of ions due to confinement
improvement cannot be ruled out).
In most discharges for which ne < 2.5 X 10 20 ro~ 3 , IBW injection results in a significant increase in the global particle confinement time, r p .
Typically, a large increase in electron density, accompanied by a decreasein the particle source rate as inferred from the H a emission, is observed.

517

IAEA-CN-47/F-II-2
400

I

i

i

i

i

i

i

a 30<P HF <60 kW
0 60<F|, F <9d kW
300

.

4

Ai2O<P RF <17OkW -A

200

B=7.6 T
Hydrogen

- r
&

"

O

100

-

0.5

1.0

1.5

2.0

2.5

3.0

3.5

4.0

F/G. (5. A7} vermr r^/or B = 7.6 T; hydrogen ions.

When compared with discharges in which the density is increased by gas
puffing alone, the r p during IBW is as much as three times larger. Improved TP are observed for both B ~ 7.6T and B ~ 9.4T. At B = 5T,smaller(10%-20%) improvement of rp was observed, perhaps due to the lower injected powers. The increases in rp are strongly dependent upon the densityof the target plasma. The maximum fractional increase in TP occurs in target plasmas whose line averaged density is ~ 0.7 x 102Om~3. As the pre-rf
target density is increased, the enhancement in rp is reduced until aboven e > 2.5 X 102Om~3, no effect upon rp is observed. This dependence isshown clearly in Fig.7. The changes in global particle confinement time
were associated with similar improvements in the central plasma (and notjust near the periphery which dominates the global rp). Central impurity
particle confinement times were measured by injecting trace amounts of
Si using the laser blow-off technique. The impurity confinement times
also increased during IBW injection, typically by a factor of ~ 1.5 - 2,
but sometimes by as much as a factor of 2.8. Before IBW injection, rp isfound to be 3-6 ms at ne ~ 0 . 9 x 10 20 m~ 3 , with large fractional increases
occurring with IBW injection, as described above. Using these values ofTp, at lower densities (n e < 1 x 102Om~3) the convection term could be
important in the ion energy balance.
In the present experiments, the excitation, propagation and absorption of directly launched ion Bernstein waves (IBW) were studied using
CO 2 laser scattering [15]. The scattering volume was located at the same
toroidal location as the IBW antenna. The identification of the IBW wasPORKOLAB et al.

518
3.5

O B,=9.2-9.5 T• B,=7.4-8.0 T

3.0

I"
o
for P rf >80 kW

t
a.

O

1 .0
0.5
0.0

1.0

2.0

3.0

O

O -

4.0

target n e (x10 2 0 m" 3 )FIG. 7. Particle confinement time with and without IBW injection versus density. B = 9.4 T.

made by mapping out the dispersion relation. The excitation of the IBW
is optimized when In evaluated at the antenna is just below an integer
value, as expected from theory. The scattered CO 2 signal intensity at
x/a ~ +0.4, B — 7.5T, is shown as a function of density in Fig. 8a. A
sharp decrease of the scattered signal is observed above n e ~ 1 X 10 20 m~ 3 .
This density corresponds to the ion heating "density limit", and the lossof ion heating may be attributed to the inaccessibility of the IBW to theplasma interior. To check this hypothesis, the level of low-frequency fluctuations was measured near the plasma edge by CO 2 scattering. The relative magnitude of these fluctuations is plotted on Fig. 8b. A strong correlation between the increased level of low frequency edge fluctuations andthe decreased level of IBW in the plasma interior is observed as the density
is increased above ne > 1 x 10 20 m~ 3 . In the present case a minimum in the
low frequency fluctuation activity is observed at ne ~ 1 x 102Om~3> and at
the same density IBW and ion heating (see Fig.6) maximize in the plasmainterior. Thus, the density limit for IBW heating may be associated withstrong scattering by low frequency density fluctuations [16].
A brief discussion regarding the physics of IBW heating in Alcator C isin order. As shown in Ref. 14, we expect that nonlinear absorption should
dominate (even in the presence of minority species) at the resonances
1$ = 2.5 (m = 5, B = 5T) and ($ = 1.5 (m = 3, B = 7.5T). The
situation at B=9.3T is more complex. In particular, both nonlinear andlinear absorption can take place near the plasma edge where i£ — 1.5(TO = 3) and £p = 3, or nonlinear absorption can take place near the

519

IAEA-CN-47/F-H-2
40

•

•(a)

•

|

i

|

i

•

-20 -

t

t
I

1

2

3

4

ne (10 2 0 m- 3 )FIG. 8. Scattered CO2 power versus nefrom (a) IBW at rla = 0.4, and (b) low frequency fluctuations,(An/n)2 at rla = 0.83. Hydrogen plasma, B = 7.5 T.

center on the deuterium minority where £D — 2.5 (m = 5) (this may
be the only possible process at B>10T). The nonlinear threshold for one
e-fold decay of the transmitted power in a single pass is given by [14]1.14 x1019Wk±2R(m)(P(kW)/S(m2))> 1

where W is the nonlinear matrix element whose values are given in Fig.9 for
m=3, 5. P/S is the transmitted power density, and Y = (fcj./2)(w^i/wpi) X(deR/dk±) is a quantity proportional to the group velocity. Typical valuesof Y vary from Y ~ 0.3 — 0.1 for k±rci ~ 1.2 — 2. The other plasma parameters are in standard notation, including R, the major radius. The
calculated threshold powers for each regime are listed in Table 1. Here
we assumed an effective power transmission area of S=2A, where A is theantenna surface. We see that the calculated thresholds are in reasonableagreement with the observed values at B=5T and B=7.5T. The situation
at B=9.3T is complicated by the presence of the surface absorption layer
at r ~ 10 cm (r/a ~ 0.83). In order to explain the observed energetic
deuterium component, at least part of the wave power penetrates to theplasma center. This phenomena is not well understood at present, especially since linear absorption is also effective at the plasma edge. If edge
absorption is operative, it also produces significant central bulk heating

PORKOLAB et al.

520
0.10

0.08 -

0.06 -

0.04 -

0.02 -

0

1

2

3

4

5

6

7

8

FIG. 9. Non-linear cyclotron damping matrix element, W, versus k±rciforIBW.

of hydrogen, apparently by improving energy confinement of ions near thecenter. In addition, tens of kW s of power would be transferred to the H
majority by collisional equilibration from the energetic deuterium minority
component. Further complications arise from significant ponderomotive
forces at the surface. Detailed theoretical and transport code studies ofthese complex phenomena are underway.

Particle Confinement in LHCD Plasmas
in the Versator II Tokamak
We have reported previously that significant density increases (up to
factors of two) observed during combined ohmic and 800 MHz LHCD atdensities ne <, 6 x 10 1 8 TO~ 3 were due to the improvement of global particle confinement [17]. Recently the RF frequency was raised to 2.45 GHz,and RF current drive was demonstrated at densities n e ;$, 2 x 10 19 m~ 3[18]. Significant density increases ( (An e /n e ) ~ 1) have also been ob-

521

IAEA-CN-47/F-H-2

TABLE 1. THRESHOLD POWERS

B(0) (T) m

Resonanceat (r/a)Ion Species
absorbingTi(r) Pth (kW)
n(r)
20
3
(keV)
(10 m- )

5.2

5
5

0.5
0.5

H (maj)D (min)

0.8
0.8

0.4
0.4

16
0

7.6

3
3

-0.3
-0.3

H (maj)D (min)

0.8
0.8

0.7
0.7

36
0

9.3

3
3
5

0.9
0.9
0

D (min)H (maj)2% D (min)

0.1
1.0

0.1
1.0

0
0.1
30

served during injection of the 2.45 GHz RF power, even at densitiesn ( > l x 10 19 cm~ 3 . In Fig. 10, the line-averaged density, n e , is shown toincrease from n e ~ 1.3 X 1019 to ~ 1.9 x 1019cm~3 when 60 kW of 2.45
GHz RF power is applied in the current drive mode (Acf> — 90° relative
waveguide phasing). The density profile is broadened and increased while
the Ha profile height is decreased during RF injection. Since emissions
from carbon and oxygen impurities did not show any increase during RF
injection, the electron source term S is dominated by ionization of hydrogen neutrals. The global confinement can be calculated from the particle
balance equation, dNejdt = S — Ne/rp. The particle confinement time isfound to increase by a factor of ~ 2.2 during RF injection. The density
increase is largest for a relative waveguide phasing of A<j) = TT/2 (current drive direction) and is smallest when A<j> = — TT/2 (anti-current drive
direction). It should be noted that the improved confinement regime coincides with the current drive density regime, i.e., n e <, 2 X 10 1 9 TO~ 3 . The
upper value of the density at which particle confinement improvement isobserved also depends on the toroidal magnetic field. For example, whenthe magnetic filed is raised from 0.9T to LIT, the upper density limit

PORKOLAB et al.

522

CA

PRP=60kW
20

25

30

Time (ms)FIG. 10. Temporal evolution of the line-averaged density ne, the total electron number Ne, the ionization source term S deduced from Ha emissivity profiles, and ike calculated global particle confinement
time rp in Versator II during LHCD.

value is increased from 1.6 x 1019cm~3 to 2.0 x 10 19 cm~ 3 . This dependence may be explained by improved wave access when the magnetic field
is increased, in agreement with other current drive effects. The exact
physical mechanism responsible for the improved confinement during RF
current drive has not been identified yet.

ACKNOWLEDGEMENTS
The authors are grateful to the Alcator and Versator technical staff for their help
in these experiments. They also thank Dr. S. McCool and Dr. R. Watterson for theircontributions to the Thomson scattering experiments.
This work was supported by the United States Department of Energy, underContract No. DE-AC02-78ET51013.

IAEA-CN-47/F-II-2

523

REFERENCES
[1]
[2]
[3]
[4]
[5]
[6]
[7]
[8]
[9]
[10]
[11]
[12]
[13]
[14]
[15]
[16]
[17]
[18]

KNOWLTON, S., et al., Phys. Rev. Lett. 57 (1986) 1529;
TAKASE, Y., et al., Nucl. Fusion 27 (1987) 53.
PORKOLAB, M., et al., Phys. Rev. Lett. 53 (1984) 450.
KAYE, S.M., GOLDSTON, R.J., Nucl. Fusion 25 (1985) 65.
BONOLI, P., ENGLADE, R., PORKOLAB, M., in Heating in Toroidal Plasmas (Pcoc. 4th Int.
Symp. Rome, 1984), Vol. 2, Int. School of Plasma Physics, Varenna (1984) 1311.
TEXTER, S., Ph.D. Thesis, 1986; Rep. PFC/JA-86-10, Plasma Fusion Center, Massachusetts
Inst. of Tech., Cambridge (1986).
PORKOLAB, M., et al., Phys. Rev. Lett. 53 (1984) 1229.
PORKOLAB, M., et al., in Controlled Fusion and Plasma Heating (Proc. 13th Europ. Conf.
Schliersee, 1986), Vol. 2, European Physical Society (1986) 445.
PORKOLAB, M., et al., Bull. Am. Phys. Soc. 30 (1985) 1493.
SOLDNER, F., et al., Phys. Rev. Lett. 57 (1986) 1137.
CHU, T.K., et al., Nucl. Fusion 26 (1986) 666.
VAN HOUTTE, D., et al., Nucl. Fusion 24 (1984) 1485.
ONO, M., et al., Phys. Rev. Lett. 54 (1985) 2339.
HOSEA, J., et al., Plasma Phys. Controll. Fusion 28 (1986) 1241.
ABE, H., et al., Phys. Rev. Lett. 53 (1984) 1153.
PORKOLAB, M., Phys. Rev. Lett. 59 (1985) 434.
TAKASE, Y., et al., Bull. Am. Phys. Soc. 31 (1986) 1587.
ONO, M., Phys. Fluids 25 (1982) 990.
LUCKHARDT, S.C., et al., Phys. Fluids 29 (1986) 1985.
MAYBERRY, M.J., et al., Phys. Rev. Lett. 55 (1985) 829.

DISCUSSION
R.J. GOLDSTON: The Alcator team has previously identified ion thermal
losses as the cause of the deterioration of confinement at high density. Now you havea tool to heat ions or perhaps to improve their confinement. How do the electrons
respond when the ion temperature rises?
M. PORKOLAB: The electron temperature remains nearly constant as the RF
power is injected. In some instances, at the highest power levels, Te may decreaseby 10-20% because of radiation losses associated with injection of high Z; impurities. In all cases, however, Te > T( is observed. Since we heat ions exclusively, it
is hard to say anything about electron confinement.
K. UEHARA: Did you observe a relationship between the energy confinement
time and PRF in the low density regions where TE is better than in the OH case? And
what was TE in the low density region as compared with TE during injection of RF
power?
M. PORKOLAB: Our Thomson scattering equipment does not operate reliably
at densities below 3 x 1019 nr 3 . Hence, we have no information on confinement atvery low densities, i.e. below 1 x 1019 nr 3 . However, if the trend in our dataremained the same, I would expect TE (RF) > rE (OH) at these low densities. This
would be consistent with observations in other machines.

IAEA-CN-47/F-II-3

CURRENT DRIVE BY LOWER HYBRID WAVES
AT 3.7 GHz IN PETULA B
F. PARLANGE1, D. VAN HOUTTE1,
H. BOTTOLLIER-CURTET, G. BRIFFOD, P. CHABERT1,
M. CLEMENT, C. DAVID2, A. GIRARD, C. GORMEZANO2,
W. HESS1, M. HESSE1, G.T. HOANG1, G. ICHTCHENKO,
J.M. RAX1, F. RYTER3, J.C. VALLET1
Association Euratom-CEA,
Departement de recherches sur la fusion controlee,
Centre d'etudes nucleaires de Grenoble,
Grenoble, France
Abstract
CURRENT DRIVE BY LOWER HYBRID WAVES AT 3.7 GHz IN PETULA B.
The density limit for Lower hybrid current drive is extended to 8 x 1013 cm" 3 by increasing thefrequency from 1.3 to 3.7 GHz. When the density is increased above 2 x 1013 cm" 3 , the plasma
current cannot be replaced entirely by RF current, because of limited power and wave absorption saturation. Even with partial current replacement, sawteeth can be suppressed with rather low power. In somecases, this may be due to a broadening of the current profile and the vanishing of the q = 1 surface.
In other cases, there is some evidence that relaxations may be suppressed while q < 1 on axis.

1.

INTRODUCTION

The capability of lower hybrid current drive to produce more stable discharges
than inductive drive is of great interest at present, in view of the high temperaturerelaxation limiting performances in today's experiments [1]. The spatial distributionof wave driven currents is a priori different from that in the Ohmic case. In addition,
the existence of fast electrons having the wave velocity affects the plasma DC
conductivity. Therefore, even partial sustainment of a discharge by waves should
modify the stability of tokamak plasmas against MHD resistive modes and sawtooth
relaxations.
These effects were demonstrated in an early work on Petula [2] and are
currently being observed on other tokamaks [3-5]. Up to now, experiments on Petula
were limited to low densities, r^ < 1.5 X 10 13 cm~ 3 , because of the low RF
frequency of 1.3 GHz and critical density phenomena.
1

CEA, Centre d'&udes nucldaires de Cadarache, Saint-Paul-lez-Durance, France.
JET Joint Undertaking, Abingdon, Oxfordshire, UK.
3
Association Euratom-Confederation Suisse, Ecole polytechnique fe'de'rale de Lausanne,
Lausanne, Switzerland.

2

525

526

PARLANGE et al.

Current drive studies in Petula are now being extended to high densities of up
to 10M cm" 3 , by adding a new lower hybrid equipment at 3.7 GHz [6](500 kW/50 ms). The grill with 2 x 9 waveguides is of the multijunction type andemits a well defined Ny spectrum which can be varied from 2.7 to 1.5.
Experiments were made with the following parameters: R = 0.72 m,
a = 0.165 m, BT = 2.8 T, Ip = 100-170 kA, ne = (0.5-10) x 1013 cm' 3 .

2.

DENSITY LIMITAt low densities (n,, ~ (0.8-1.6) x 1013 cm 3 ), the current drive efficiency
is similar to what was obtained at 1.3 GHz. The usual figure of merit isnIR/PRF ~ 1019 A-m" 2 -W~'. At higher densities, where current drive is no longer
observed at 1.3 GHz because of the density limit effect, a significant decrease in loop
voltage, at constant plasma current, is still measured during the 3.7 GHz RF pulse,
up to the practical operating density of Petula, ne ~ 1014 cm" 3 (Fig. 1).
However, no complete replacement of the Ohmic current is obtained fordensities higher than ne ~ 2 x 1013 cm" 3 , owing to the l/ne density scaling of thelower hybrid current drive and, possibly, other phenomena such as wave accessibility
or absorption saturation.

1.0 -

0.5 -

i

I V
•\ i . 3 GHuz
i

^ .

i

i

D N// = 1.7

PHFo N// - 2.7

PHF

D

-cfi>

;J.7 GHz= 300 k W
- 400 kW•/—eft-.
i

t

1

X

10

3

4

5

6

7

10

?Te(1013cm-3)

FIG. 1. Relative loop voltage drop and parametric decay wave excitation during lower hybrid
current drive.

527

IAEA-CN-47/F-II-3

In particular, the small effect on loop voltage at n<. = 8 x 1013 cm""3 could
equally be explained by the limited power available, P < 400 kW, or by the density
limit. On the other hand, parametric decay wave excitation starts increasing sharply
above noise level, for densities higher than 8 x 1013. By similarity with the
1.3 GHz case, this was considered to be a criterion for the density limit. Thisvalue, when compared to the value for the 1.3 GHz waves, is in agreement with ascaling n = fa, where 1.7 < a < 2, which so far has only been inferred frommeasurements made on other devices [7].

3.

CURRENT DRIVE EFFICIENCY AND SATURATION

If all available power were absorbed, the plasma response to the electric field
E and the RF power launched into the torus, P, would be given by [8, 9]:J « crE + 77 a P + TJ2 Ea P
where a is the Spitzer conductivity, T; the current drive efficiency given in [8] anda the fraction of power fulfilling the accessibility condition. The third term on theright hand side describes the RF enhanced conductivity.
This formula gives reasonable agreement for results obtained with a mean wave
index N| = 2.7, up to some power where the voltage drop saturates (Fig. 2), atleast for densities lower than 4 x 1013 cm" 3 . The power threshold for saturation is

6x10" cm"3

PRF(100kW/div.)

FIG. 2. Loop voltage during the RF pulse for two different densities. Solid line is the result of
simulation.

528

PARLANGE et al.

an increasing function of density. Saturation is also observed for the same power onelectron cyclotron emission [10]. This saturation is interpreted as a limitation in
power absorption when collisions are not sufficient to prevent the formation of aplateau in the electron distribution function.
In short, compared to the linear relation I = n a P, the experimental points in
Fig. 2 show an improved efficiency at low power, due to enhanced conductivity anddegradation at high power because of limited absorption.
For N,| = 1.7, accessibility becomes a severe problem preventing propagationof power to the plasma centre for nc > 4 x 10'3 cm" 3 (a = 0) and giving
a = 5% for ne = 3.2 x 1013 cm" 3 . Nevertheless, the experimental curves arequite similar to the N« = 2.7 case and can be fitted by taking a = 10%, whichcorresponds to accessibility at mid-radius. Saturation is observed, as well.

4.

SUPPRESSION OF INTERNAL DISRUPTIONSStudies of sawtooth activity during lower hybrid current drive were carried out
with an array of soft X-ray diodes supplemented, for some shots, by fast IR
interferometry and ECE emission.
For the whole density range of 1 to 6 x 1013 cm" 3 , the soft X-ray emission isstrongly affected by the lower hybrid current drive and may increase by a factor ofthree above the Ohmic level, on the central chord.
The sawtooth and MHD behaviour during current drive on Petula depends onthe density:
For densities lower than 3 x 1013 cm" 3 , the sawtooth suppression isassociated with the triggering of an m = 2 mode in a way similar to the
1.3 GHz case [2]. When the density is increased, the MHD oscillations are
triggered later and later until, for ne = 3 x 1013 cm" 3 , they cannot be
observed any longer during a 100 ms RF pulse. Soft X-ray emission starts
increasing in the plasma centre but levels off when the m = 2 mode is present.
(2) In the (3-4.5) X 10 13 cm" 3 density range, amplitude and period of thesawteeth increase at low RF power (Fig. 3). As the RF power is raised, the sawteeth disappear for some time and may start again. The sawtooth structure
becomes more complicated. For sufficiently high power, complete stabilizationis achieved from the beginning of the RF pulse. On the central chord, the X-ray
level goes to a maximum within 8 to 15 ms, after which the emission isredistributed to the external parts of the plasma. This results in a highly peaked
emission for the first 10 ms, which relaxes to a profile much broader than the
Ohmic one (Fig. 4). The broadening of the SXR emission seems to be triggeredby an m = 1 mode, which is detected slightly before the maximum at the centre
and may last for the rest of the RF pulse.

(1)

529

IAEA-CN-47/F-II-3
Central
I

soft

X-ray
a

p

=120 kA, n = 3xlO&

emission3

cm~ and <N,,> = 1.7
//

1.2

ISO

FIG. 3. Soft X-ray signal on central chord at three different powers compared to Ohmic case. Noiseon two highest power traces is m = 1 MHD oscillation.

I S.X.R.(a.u')
Ip - 122 kA

fie ' 3.4x10 n cm- 3
N// - 1.7

.2"-

-5

10

15

FIG. 4. Soft X-ray emission profile for he = 3.2 x 1013 cm 3. Before RF pulse (solid circles), 10 ms(circles) and 25 ms (squares) after beginning of RF pulse.

530

PARLANGE et al.
S.X.R | r = 0
3.6 • fo.u.l

/

i

1
2.7

1

~f

1

1

i fT

1

/1

111

1.8

7
9

. i

n j* . i JL* < . i J . * * • I . . i > « •.

l/i/iftfcnjll\fif1.

F/G. 5. 5o^ X-ray signal as a function of time for nt = 6 x / 0 " cm 3. Bottom to top: Ohmic
discharge, 150 kW RF discharge, 240 kW RF discharge. Time-scale is 8 ms/div.

The inversion radius of the sawtoothing discharges has been determined for
different RF powers. Up to the power threshold for stabilization, this radius remains
roughly constant and close to its value for the Ohmic case, although amplitude andperiod of the sawteeth have been increased by a large factor.
These two facts — constancy of inversion radius and presence of m = 1
oscillations — lead to the conclusion that sawteeth can be inhibited with central
q-value smaller than one. Alfven wave measurements [12] indicate, on the other
hand, a decrease in the internal inductance but their interpretation is not yet clear.
(3)

At higher densities, ne = (4.5-6.5) x 1013 cm" 3 , the picture is much
simpler. At low RF level, the total soft X-ray emission increases without mucheffect on sawtooth amplitude and frequency. For sufficiently high power(~240 kW), internal disruptions disappear completely, and the soft X-ray levelmay increase for the duration of the RF pulse without the large emission
rearrangement observed at lower densities (Fig. 5). m = 1 oscillations withlow amplitude are sometimes observed. At the highest densities, the RF pulse
is clearly too short for a stationary situation to be established.

All these observations refer to the case of NB = 1.7. It is to be noted that sawteeth could not be suppressed at high density, with the Nj = 2.7 spectrum, although
the effects were quite similar for low powers, for the two spectra.

5.

CONCLUSIONS

Current drive effects were observed up to 8 X 10 cm 3. The Ohmic current
replacement was limited by the high power required for high density and absorption
saturation, which impairs the apparent efficiency, in the intermediate density range.

IAEA-CN-47/F-II-3

5 31

Even at the highest densities, temperature relaxation is suppressed withmoderate power levels, compared to what would be required for complete replacement of the inductive drive (AV/V ~ 25%).
The power requirements could be further reduced by using the accessibility
properties of the lower hybrid waves to deliberately generate an off-axis current [11].
Experiments along this line in Petula were difficult to interpret. Even at the highest
densities, the efficiency of waves with fast phase velocities, N[ = 1.7, remains highalthough accessibility would localize these waves in the plasma periphery. Moreover,the presence of m = 1 oscillations, already observed on PLT [5], indicates thatsawtooth stabilization mechanisms more complex than the simple decrease in central
current density have to be considered.
ACKNOWLEDGEMENTS
The authors wish to thank Mr. Agarici and Mr. Panzarella for their technicalsupport and the RF team for the design and operation of the RF system.
REFERENCES
[1]
[2]
[3]

LALLIA, P.P., JET TEAM, Plasma Phys. Contrail. Fusion 28 (1986) 1211.
VAN HOUTTE, D., et a]., Nucl. Fusion 24 (1984) 1485.
PORKOLAB, M., et al., in Controlled Fusion and Plasma Heating (Proc. 13th Europ. Conf.
Schliersee, 1986), Vol. 2, European Physical Society (1986) 445.
[4] SOLDNER, F.X., et al., Phys. Rev. Lett. 57 (1986) 1137.
[5] HOSEA, J., et al., Plasma Phys. Contrail. Fusion 28 (1986) 1241.
[6] REY, G., et al., in Controlled Fusion and Plasma Heating (Proc. 13th Europ. Conf. Schliersee,1986), Vol. 2, European Physical Society (1986) 339.
[7] WEGROWE, J.G., TONON, G., Non-inductive Current Drive in Tokamaks (Proc. IAEA
Techn. Comm. Meeting Culham, 1983), Vol. 2, Culham Laboratory, Abingdon (1983) 343.
[8] FISCH, N.J., Phys. Fluids 28 (1985) 245.
[9] RAX, J.M., Saturation Induction and Radiation in Wave-Driven Plasmas, Rep. EUR-CEAFC-1298, Association Euratom-CEA sur la fusion, Centre d'e'tudes nucteaires de Cadarache,Saint-Paul-lez-Durance (1986).
[10] GIRARD, A., et al., in Controlled Fusion and Plasma Heating (Proc. 13th Europ. Conf.
Schliersee, 1986), Vol. 2, European Physical Society (1986) 374.
[11] BRIFFOD, G., et al., ibid., p. 453.
[12] COLLINS, G.A., et al., submitted to Plasma Phys. Contrail. Fusion.

DISCUSSION
F.X. SOLDNER: In your sawtooth stabilization results you report that for somecases there is a continuous increase in the intensity of the soft X-ray emission after
suppression of the sawteeth. You attribute this to the occurrence of one giant sawtooth. Could it not be just a continuous acceleration of the fast electrons generated
by your low NB spectrum in the DC electric field?

532

PARLANGE et al.

F. PARLANGE: The giant sawtooth was mentioned only to stress the fact that
we do not have a stationary situation and that the X-ray level could not go on increasing for ever if the RF pulse was longer. As far as the continuous acceleration of fastelectrons is concerned, it should be noted that this occurs only at high density(6 x 1019 m~3) and not at lower densities ( < 4 x 1019 m" 3 ), and that the hard
X-ray signal is quite flat.
K. McGUIRE: Do you observe improved confinement time or increased stored
energy when you stabilize the sawtooth oscillation?
F. PARLANGE: Neither electron cyclotron emission nor X-rays can be used
to measure the electron temperature profile during current drive because of thepresence of the energetic tail. Furthermore, Thomson scattering was not reliable
during the 3.7 GHz campaign, so that we have no possibility to measure the stored
energy.

IAEA-CN-47/F-II-4

EXPERIMENTS ON
CURRENT DRIVE AND PLASMA HEATING
BY LH AND EC WAVESIN THE T-7 TOKAMAK
V.V. ALIKAEV, A.A. BORSHCHEGOVSKIJ,
V.V. CHIST'AKOV, Yu.A. GORELOV, V.I. IL'IN,
D.P. IVANOV, N.V. IVANOV, A.M. KAKURIN,
P.P. KHVOSTENKO, A.Ya. KISLOV, V.A. KOCHIN,
P.E. KOVROV, K.I. LIKIN, Yu.S. MAKSIMOV,
I.B. SEMENOV, A.F. SHIGAEV, Yu.A. SOKOLOV,
N.L. VASIN, V.V. VOLKOV
I.V. Kurchatov Institute of Atomic Energy,
Moscow, Union of Soviet Socialist Republics
J. DATLOV, V. KOPECKY, L. KRYSKA
Institute of Plasma Physics,
Czechoslovak Academy of Sciences,
Prague, Czechoslovakia
Presented by F. Zdcek
Abstract
EXPERIMENTS ON CURRENT DRIVE AND PLASMA HEATING BY LH AND EC WAVES IN
THE T-7 TOKAMAK.
A microwave set in the ECR range, including two gyrotrons with X = 4.8 mm and a total powerof about 400 kW, was added to the lower hybrid set (600 kW, f = 900 MHz), for further developmentof current drive experiments in T-7. Experimental results on the application of LH and EC power arepresented in this paper. The initial stage, where gas breakdown was performed by an EC wave, wasfollowed by LH current ramp-up. In the experiments on LH current sustainment, attention was paid tothe dependence of current drive efficiency on electron temperature and to a study of the plasma power
characteristics. Application of ECRH allowed a rise of the electron temperature from 1 to 4 keV, atplasma densities within a range of ne = (5-7) x 1012 cm" 3 . In this case, a rise in the current drive
efficiency was observed. The LH current drive was accompanied by substantial electron heating. The
electron energy confinement time for the LH and EC heating techniques is well described by an 'Ohmic'
scaling law.

533

534

1.

ALIKAEV et al.

INTRODUCTION

In further developing the LHCD experiments in the T-7 tokamak [1, 2], grill
waveguide conditioning by glow discharge was used which enabled us to raise the
LH power launched. In addition to the LH system, the tokamak was supplied with
an EC system including two gyrotron generators operating at f = 63 GHz. The
output power of each gyrotron is 200 kW for a pulse duration of 100 ms. 90% of
the gyrotron power is radiated as an electromagnetic wave with a polarization
corresponding to the ordinary wave in the plasma. The power from the gyrotrons islaunched into the tokamak chamber along oversized cylindrical waveguides, 80 mm
in diameter, through a horizontal port. A mode converter, which converts theincident radiation with ordinary polarization into a radiation with a polarization
corresponding to the extraordinary wave, is installed at the chamber wall opposite
to the waveguides. The reflected radiation mainly propagates at an angle 80° to thetoroidal magnetic field.
The main parameters of the T-7 tokamak are: the major and minor chamberradii are 122 and 35 cm. The main ring-like limiter has a radius of 31.5 cm, the rail
limiter is located at the bottom part of the chamber at a distance of a = 25-31.5 cm
from the chamber axis. The toroidal magnetic field in most cases was BT = 2.3 T,
the discharge current was I < 200 kA in deuterium. The magnitude of the longitudinal moderation of the LH wave related to the spectral power distribution maximum
at the grill mouth for a relative phase shift of 90° corresponded to NB « 3.
The set of diagnostics enabled us to record the plasma parameters including thediamagnetic flux, the electron density along six chords, and the X-ray spectra in theenergy range of 2-150 keV (with two germanium detectors). One detector was
located in the horizontal port and the other in the vertical port. The line of sight ofeach detector could be scanned so that the spatial X-ray spectra distributions could
be determined. The radiation losses were recorded by a wide angle pyroelectricdetector. The impurity spectral line intensities were recorded by optical instrumentsin the visible area of the spectrum. The ion temperature at the plasma column axis
was recorded by a five channel charge exchange neutral spectrum analyser.
2.

NON-INDUCTIVE DISCHARGE STARTUP

Experiments, in which gas breakdown, ramp-up and sustainment of the toroidal
current were performed by RF methods only, were recently carried out in sometokamaks [3, 4]. Using these techniques, one can expect startup and continuous
plasma sustainment in tokamaks without inductors or, at least, substantial savings in
the inductor magnetic flux.
Only ECR power — 360 kW — was launched into the tokamak at the first stageof the gas breakdown experiments. A toroidal field of 2.3 T, corresponding to the
ECR zone position near the tokamak chamber axis, was used. The primary windingof the tokamak was short circuited; deuterium was the working gas.

535

IAEA-CN-47/F-H-4
n e (10 12 cm- 3 )•\l-30 -20 -10

0

10 20

#r(cm)

FIG. I. Spatial distributions of average (over vertical chords) plasma density.

-20

-10

0 r (cm)

FIG. 2. Signals of pyroelectric detectors versus position of ECR point.

too

o

50

\

n
is

20

25

JO r (cm)FIG. 3. Spatial distribution of electron temperature.

o2/'o ° ,

0
o

°

"i

-2
I L (A)20 oni ° ° °i

o

o

if"0;

-S3 -40 -?0

0

?0

40

FIG. 4. Toroidal and limiter currents versus B±.

536

ALIKAEV et al.
I, (A),

Z

h

6

&

n e (10 12 cnr 3 ) 'FIG. 5. Rail limiter current versus plasma density.

0

50 100 150 200

050 100 150 200t(ms)FIG. 6. Oscillograms of toroidal current and loop voltage.

Effective gas ionization occurred under ECR power injection into the tokamak
chamber. An almost unchanged plasma density, depending on the initial neutral gas
pressure, prevailed up to the end of the microwave pulse, after a 10-15 ms transient
process. A typical spatial plasma density distribution at a pressure of 3 X 10 " 5 torr
is shown in Fig. 1. At pressures 2: 10 " 4 torr, gas ionization occurred nonsimultaneously in different chamber regions; signals from the interferometerchannels located at r = -15.5 cm and at r = - 6 . 5 cm increased with = 5 m s delay
in comparison with the signals from the other channels.
The signals from two wide angle pyroelectric detectors placed in the upper
diagnostic port versus the toroidal field (and hence the ECR zone position) arerepresented in Fig. 2. The first detector r = 0 recorded the energy losses, P, fromthe plasma by radiation and charged particles, the second detector (r = - 1 0 cm)
measured the losses by radiation only, P r The direction of the magnetic fieldcorresponded to the upward direction of electron toroidal drift. The electron temperature as measured by a double Langmuir probe was 70-80 eV (Fig. 3).
A toroidal current depending on the vertical magnetic field B x (Fig. 4)
was generated when the ECR power was injected in case the tokamak primarywinding was short circuited. In magnitude and direction, this current coincided withthe current which should be driven as a result of electron accumulation on top
due to toroidal drift and their drain-off along the inclined magnetic field lines:

IAEA-CN-47/F-II-4

537

I = a2enevdBT/B± (lines 1 in Fig. 4). For low B±, the driven current should be
limited at a level where the field produced by this current reaches a value « B X :I = acB ± /2, which is observed in the experiment (line 2 in Fig. 4). The current IL
passing between rail limiter and chamber wall is shown in the same figure. The direction of this current corresponds to the toroidal drift direction of charged particles, andits value depends on the plasma density, as is shown in Fig. 5.
The breakdown was followed by a stage of toroidal plasma current ramp-up by
the lower hybrid wave. ECR and LH systems were switched on simultaneously inthis experiment. The ECR system was under operation for 30 ms, and the LH systemfor 150 to 200 ms, at a power of 300 kW.
If the primary tokamak winding was short circuited, the current rose up to
15-20 kA in 30-40 ms, after which the current rise was decelerated substantially(Fig. 6(a)). The current trace virtually did not depend on the duration of ECR system
operation. Probably, ramp-up deterioration was caused by the self-inductive electriccounter field in the plasma which affects hot electrons with energies of mc 2 /2N|.
The time of their appearance must be of the order of magnitude of the Coulomb
collision time for the electrons of such energies, i.e. r « 40 ms.
Changing the sign of the loop voltage to positive (U = 0.5 V) by means ofprimary tokamak winding current control made it possible to retain the high rate of
the plasma current ramp-up during the entire time of LH system operation(Fig. 6(b)).

3.

TEMPERATURE DEPENDENCE OF LOWER HYBRID CURRENT
DRIVE EFFICIENCYThe question of the dependence of LHCD efficiency on the plasma electron
temperature seems to be very important for a better understanding of the LHCD
phenomenon as well as from the point of view of application of LHCD to a
tokamak reactor. The efficiencies obtained in experiments carried out so far (see,e.g. [1, 2, 5, 6]) are not sufficiently high for a reactor, and previous theoreticalconsiderations [7] of the current drive mechanism do not predict any change in efficiency when we go over from temperature values of Te « 1 keV that are typical ofpresent experiments to reactor scale values of Te = 10 keV. However, an analysisof the experimental data suggests that these considerations are not quite accurate andthat the current drive efficiency should depend on the electron temperature in theconditions of a real experiment.
In fact, it was shown in Refs [2, 8] that, in contrast to Ref. [7], the lower velocity limit, V,, of the quasi-linear plateau on the electron distribution function does
not coincide with the short wavelength boundary of the spectrum of the LH waves
launched into the plasma from an external source. The probable reason is thedevelopment, in the plasma, of some mechanism of LH wave spectrum broadening
towards parallel phase velocity reduction. This broadening continues until a balance

538

ALIKAEV et al.

is reached between the input RF power transferred to the resonant electrons and theirpower lost by Coulomb collisions. In the case where the whole current is driven by
LH waves, the value of V, can be found from the known current density whichapproximately corresponds to the condition q(0) = 1, from the equationJ_ en. V j - V?
~ V* VTc
2-v^e _
~

cB
2 TTR

{

'

Here, ne is the electron density, VTe is the electron thermal velocity, and V[ and V2
are the limits of the quasi-linear plateau on the electron distribution function. For
ordinary experimental conditions, Eq. (1) yields the approximate solution
V, ~ 2VTe, which is only weakly dependent on the other plasma parameters. As aresult, an electron temperature dependence appears in the expression for the current
drive efficiency:=

Zeff + 5fl^R=
6
P + ILH Um
V j - V]
2 33
18 ir e L In (V2/V,)

where U is the loop voltage, ILH = IAU/U0 the driven current, PLH the LH power,
and L is the Coulomb logarithm. In formula (4), the driven current is assumed toflow near the plasma column axis, where the plasma density n,. = 1.5ne.
To verify experimentally the dependence of the efficiency ?; on the electron
temperature we compare results obtained in three tokamak operation regimes. In thefirst regime (toroidal field BT = 19 kG, plasma current I = 110 kA, effective
charge Zeff = 2.5, and LH power PLH = 40 kW), the electron temperature was
Te = 0.8 keV. The temperature increased up to Te = 2.3 keV in the second regime(BT = 23 kG, I = 180 kA, Zeff = 2.5, PLH = 180 kW) and reached a value of
3.5 keV in the third regime (BT = 23 kG, I = 180 kA, Zeff = 5, PLH = 180 kW),
which differed from the other regimes in that ECRH was applied. The plasma density
in all three regimes was ne = (3-7) x 1012 cm" 3 . The electron temperature and theeffective plasma charge were determined from the results of X-ray spectroscopy
measurements in the energy range of 3-10 keV.
The results of measurements of current drive efficiency in the above regimes
are given in Fig. 7. The power input into the wave propagating parallel to the current
forming electron flux was used as PLH in determining the efficiency. According tocalculations, this power amounted to =70% of the total LH power input into theplasma, in our case. The result of calculating the dependence of efficiency on
temperature, by formula (4), is also shown in Fig. 7 by the dashed curve. A
characteristic phase velocity of the launched LH wave of V2 = 1010 cm-s~ l waschosen in these calculations. It will be seen from Fig. 7 that in the temperature range
investigated the efficiency tends to rise with Te. This increase agrees, within
experimental accuracy, with the result of calculation by formula (4).

539

IAEA-CN-47/F-II-4

FIG. 7. Dependence of LHCD efficiency on electron temperature. Result from calculation byformula (4) is shown by dashed line.

200
K k A ) tOO

0

U(V)

'

Pr(kW)

TV^-w^-l0.5
0{arSu^f\

0

dt 10
5

100 200 3Q0t(tns)

r

6

-10
1.5

/\

AW±(kJ)

0.5
0

\
200

300 t(ms)FIG. 8. Oscillograms of main discharge parameters under LHCD. LH pulse duration is shown bydashed lines.

4.

LH AND EC HEATING

Here, we consider the energy balance of a plasma when LHCD is applied toa discharge that was formed by the ordinary OH method. Typical oscillograms of theintegral discharge parameters are shown in Fig. 8. A fast drop in the loop voltage
was observed upon LH power launching. This voltage drop was followed by a

540

ALIKAEV et al.

to-.
0

10

SO

120 E (keV)

FIG. 9. X-ray spectra in Ohmic regime and during LH pulse.

r (cm)

W

PO

r(cm)

FIG. 10. Spatial distributions of electron temperature and density in Ohmic regime and during
LH pulse.

transient process lasting =30 ms, and then the loop voltage was established at aquasi-stationary, almost constant level. This level was 0-0.3 V, depending on the
discharge parameters fie, I and on the launched LH power. The loop voltage in the
Ohmic discharge stage was U = 0.8 V in these experiments.
Regimes with a discharge current of I = 170 kA were studied in detail. The
typical X-ray spectra recorded in the Ohmic discharge and during the LH pulse areshown in Fig. 9, where we see that, during the LH pulse, besides the appearance of
X-ray radiation within the energy range of 20-150 keV, the slope of this spectrum

IAEA-CN-47/F-II-4

541

from 2 to 10 keV is changed substantially. The radial electron temperature distributions, obtained from the X-ray spectra, and ne(r) before the LH power launch and50 ms after its triggering, are shown in Fig. 10. We see that a rise in the electron
temperature takes place upon LH power deposition (ATe = 1 keV in the central
part of the plasma column), ne rises by 30-40% and reaches 10' 3 cm" 3 at theplasma column axis. The obtained distributions n,,(r) and Te(r) show a considerable
rise, by a factor of about two, in the thermal electron energy. This result is confirmed
by diamagnetic measurements (Fig. 8).
Measurements with pyroelectric detectors showed that the radiation losses
remained virtually unchanged during the LH pulse and were equal to 10% of the OH
power deposition. No changes were obtained in the charge exchange neutral spectrain this case, the ion temperature being Tj(O) = 250 eV at the plasma column axis.
Thus, in the regimes under study, electron heating was observed besides thecurrent drive. Let us find the LH power absorbed in the plasma, PLH, and theelectron energy confinement time, T^" . PLH can be calculated from the equation forthe change in the plasma column energy:

dw LH
dt

d w0H
dt

Here, dW0H/dt and dWLH/dt are the time derivatives of the plasma total energy in
the OH regime and just after the LH system triggering, obtained from a combined
consideration of the diamagnetic signal and the plasma equilibrium signals. AU is thechange in loop voltage. The electron energy confinement time during the LH pulse
is found from the equation describing the power balance for the quasi-stationary state:= PLH + ULHI

(4)

where WfH is the electron thermal energy obtained by integrating the ne(r) and Te(r)distributions (Fig. 10). We note that the energy transfer from electrons to ions in (6)is neglected. The results of calculations of PLH, WfH and r£f , together with thecorresponding parameters in the Ohmic regime, are given in Table I. An evaluationof the parameters of the electron beam produced by LH waves yields 0.8 U and
0.2-0.3 kJ for the longitudinal and transverse beam energies and 90 kA for thecurrent transported by the electron beam.
The ECRH technique was studied in the next portion of this run. The experiments were done with discharge parameters close to those in which the LH current
drive was conducted. This enabled us to compare the efficiencies of the two heating
techniques.
The oscillograms of the integral discharge parameters shown in Fig. 11 showthat the loop voltage drops, (ft + £-/2) rises and fie increases after the gyrotrons are

ALIKAEV et al.

542

TABLE I. LH EXPERIMENT

00 12 cm 3)

(kW)

WT
PLH
(kW) (W)(ms)

OH 170

4 .6

140

-

1.3

9

LH

6 .6

40

160

2.5

12

I
(kA)

17

i

0

100

300 ttms)

FIG. 11. Oscillograms of main discharge parameters under ECRH. ECRHpulse duration is shown bydashed lines.

switched on. The power losses recorded by the pyroelectric detector and the carbonline intensities rise, in this case. A comparison of the X-ray spectra obtained for the
Ohmic discharge and during the ECR pulse shows that the electron temperature rises
from 1.5 keV to 4 keV and the effective charge rises from Zeff = 1.3 to Zeff = 3.
The Te(r) distributions obtained from the X-ray spectra and n<.(r) are shown in
Fig. 12. The power absorbed in the plasma, PECR, and the energy confinement timeof the electrons, T§f\ were found to characterize this heating. According to
Fig. 11, a substantial rise in (ft + IJT) takes place during ECRH. As the calculations based on Te(r) and Zeff(r), assuming Spitzer conductivity with neoclassicalcorrections, have shown, the current density profile does not change when the ECR

543

IAEA-CN-47/F-II-4

r(cm)

10

r (cm)

F/G. 72. Spatial distributions of electron temperature and density in Ohmic regime and during
ECRH pulse.

TABLE II. ECR EXPERIMENTI(kA)(1012 cm" 3 )

(kW)

(kW)

WT
(kJ)

(ms)

(ms)

OH

160

4

135

—

1.1

8

7

ECR

160

6

50

260

3.4

11

12

PECR

fie

power is launched; hence.the change in internal inductance A^ = 0. In this condition the time derivative (d/dt)(ft + fyl) immediately after switching the ECR poweron yields the power absorbed in the plasma, PECR. The increase in thermal energy
in the electrons 50 ms after switching ECRH on is 2.3 kJ. The energy content isfound from an integration of n^r) and Te(r) (Fig. 12).
The electron energy confinement time under ECRH is found for the quasistationary state from the energy content W-fCR and PECR:
WECR

ECR

_
"ECR +

URTR I(5)

where UECR is the loop voltage during the ECR pulse. The values of W| CR , PECR and
7f£R are shown in Table II. In the table we see that T§£R > 7^". This is partly due
to the increase in electron density during ECRH. The plasma parameters U, ne, /3r

544

ALIKAEV et al.

and rEe obtained from a numerical simulation are in good agreement with theexperimental data. The calculations were based on the T-l 1 transport coefficients [9];the difference between the OH and ECRH power deposition profiles was taken into
account.
Thus, comparing LH and ECRH results, we note that the power absorption
coefficients for LH and EC power are approximately the same and equal to 0.7-0.8.
A rise in the electron energy confinement time is observed under LHCD and ECRH
which is due to the increase in fL,. The values of T^ and T|£ R are close to the Ohmicvalue 7^eM, in this case.

REFERENCES
[1]
[2]
[3]
[4]
[5]
[6]
[7]
[8]
[9]

ALIKAEV, V.V., et al., in Plasma Physics and Controlled Nuclear Fusion Research 1982 (Proc.
9th Int. Conf. Baltimore, 1982), Vol. 1, IAEA, Vienna (1983) 227.
ALIKAEV, V.V., et al., Fiz. Plazmy 11 (1985) 53 (in Russian).
TANAKA, S., et al., in Plasma Physics and Controlled Nuclear Fusion Research 1984 (Proc.
10th Int. Conf. London, 1984), Vol. 1, IAEA, Vienna (1985) 623.
TOI, K., et al., Phys. Rev. Lett. 52 (1984) 2144.
LLOYD, B., et al., in Non-inductive Current Drive (Proc. IAEA Tech. Comm. Mtg Culham,
1983), Vol. 1, UKAEA, Culham Laboratory, Abingdon (1983) 250.
MOTLEY, R., et al., ibid., Vol. 2, p. 299.
FISCH, N.J., Phys. Rev. Lett. 41 (1978) 873.
IVANOV, N.V., Vopr. Atomn. Nauki Tekh., Ser. Termoyadernyj Sintez 3 (1985) 22 (inRussian).
MEREZHKIN, V.G., MUKHOVATOV, V.S., Pis'ma Zh. Ehksp. Teor. Fiz. 33 (1981) 463 (inRussian).

IAEA-CN-47/F-II-5

ELECTRON CYCLOTRON HEATING OF
LOWER HYBRID CURRENT DRIVEN PLASMAIN THE JFT-2M TOKAMAKT. YAMAMOTO, A. FUNAHASHI, K. HOSHINO, S. KASAI,
T. KAWAKAMI, H. KAWASHIMA, T. MATOBA, T. MATSUDA,
H. MATSUMOTO, Y. MIURA, M. MORI, K. ODAJIMA,
H. OGAWA, T. OGAWA, H. OHTSUKA, S. SENGOKU,
T. SHOJI, N. SUZUKI, H. TAMAI, Y. UESUGI, T. YAMAUCHI,
M. HASEGAWA1, S. TAKADA2, K. HASEGAWA, A. HONDA,
I. ISHIBORI, T. KASHIMURA, Y. KASHIWA, M. KAZAWA,
K. KIKUCHI, Y. MATSUZAKI, K. OHUCHI, H. OKANO,
T. SHIBATA, T. SHIBUYA, T. SHIINA, K. SUZUKI, T. TANI,
K. YOKOYAMANaka Fusion Energy Research Establishment,
Department of Thermonuclear Fusion Research,
Japan Atomic Energy Research Institute,
Tokai-mura, Naka-gun, Ibaraki-ken, Japan
Abstract
ELECTRON CYCLOTRON HEATING OF LOWER HYBRID CURRENT DRIVEN PLASMA IN
THE JFT-2M TOKAMAK.
The second harmonic electron cyclotron heating of a lower hybrid current driven plasma has been
investigated in the JFT-2M tokamak to evaluate its potential for optimizing RF current drive. It isobserved that the extraordinary mode effectively couples to electrons at largely down shifted frequencies. From measuring the soft X-ray radiation parallel and perpendicular to the toroidal magnetic field,
the existence of selective perpendicular heating of electrons with parallel velocities satisfying therelativistic cyclotron resonance condition is concluded. Initial results on 200 MHz fast wave experiments
are also reported. It is observed that, in a discharge combined with electron cyclotron heating, the fast
waves launched by a phased two loop antenna array couple to energetic electrons.

1.

INTRODUCTION

Current drive by lower hybrid (LH) waves has been achieved in a number oftokamaks. These experiments have shown that the LH waves produce a current carrying fast electron tail. Recently, electron cyclotron heating (ECH) in the presence ofa mildly relativistic electron tail has received a lot of interest, because of the needfor supplementary heating and current profile control in high phase velocity RF current driven plasmas and because of possible improvements, using ECH, of the current drive by LH waves (LHCD) and electron cyclotron (EC) waves [1, 2].
1
2

On leave from Mitsubishi Electric Corp., Japan.
On leave from Mitsubishi Electric Computer System Corp., Tokyo, Japan.

545

546

YAMAMOTO et al.

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IAEA-CN-47/F-II-5

547

On the other hand, in a high temperature and density fusion plasma, LH waves
can only drive a skin current since for a plasma with large f^./f£e the lower bound of
the allowed parallel refractive index N, is restricted to relatively high values and thewave power is then absorbed near the plasma surface. Fast waves (FW) in the frequency range of fci < f0 < fLH(0) (fLH(0) is the maximum LH resonance frequency
in the plasma) can, however, penetrate into the plasma core, even if the plasma is
hot and dense. The current drive by fast waves (FWCD) is based on the samemechanism as LHCD, i.e. the current carrying electrons are accelerated throughLandau damping; this mechanism is expected to be applicable to fusion plasmas.

2.

EXPERIMENTAL SET-UP

The JFT-2M tokamak has a D-shaped vacuum vessel with a major radius of
Ro = 1.31 m and a 0.415 m x 0.595 m minor cross-section [3]. The ECH system
consists of two gyrotrons and two oversized waveguide transmission lines [4]. The
RF power is generated by a Varian gyrotron capable of producing 200 kW at 60 GHz
for up to 0.1 s. The extraordinary mode is radiated at an injection angle of 0 = 80°
to the direction of the toroidal magnetic field from a conical horn antenna propagating
the T E n mode. The horn antennas are placed outside the torus in the midplane. The
LH waves are launched by a four waveguide array, which is fed by four 0.75 GHz
klystrons. The phase difference between adjacent waveguides is set at 90°, and thus
the Brambilla RF spectrum is concentrated between N, = 1 and 5. The fast wave
system consists of a 200 MHz RF source capable of 100 kW for up to 0.1 s feeding
to a phased two loop antenna array [5]. The RF power is divided into the two lines
by a 3 dB hybrid coupler. The phase difference between the antennas, A$ FW , isadjusted by a line stretcher. The impedance of each line is matched by using X/8
double stub tuners.

3.

EXPERIMENTAL RESULTS

3.1. Selective electron cyclotron heating
Experiments with a radiated ECH power of PECH = 240 kW and a pulse widthof td = 50 ms have been carried out to investigate the EC wave interaction withmildly relativistic electrons [6]. Figure l(a) shows the time evolution of the relevant
plasma parameters during a typical shot with a central toroidal magnetic field at
which no cyclotron resonance layer lies in the plasma column (B, = 1.4 T, corresponding to fo/(2fcO) = 0.76). For this discharge, the target plasma has a
D-shaped cross-section (a = 0.34 m, b = 0.46 m) and the plasma current Ip kept
constant by feedback control. With application of ECH power during the LH pulse,
noticeable increases in A = /3p + f/2-1, in the stored plasma energy, WDIA, and in

548

YAMAMOTO et al.

the EC emissions, IECE, are observed, with no increase in the line averaged electron
density ne, as is shown in Fig. 1. The value of A is evaluated from a pqloidal magnetic field fitting code, and WDIA is estimated from the diamagnetic measurement,
assuming isotropic pressure of the plasma.
The Bt dependence of the change in the relevant plasma parameters is shownin Figs l(b) and (c). Figure l(b) shows the absorbed ECH power, Pab, and theincrease, AWDIA, due to ECH. The value of Pab is estimated from dAWDIA/dt at theinitiation of the ECH pulse. The experimental conditions are the same as in Fig. l(a),except for the toroidal magnetic field. It is found that Pab and AWDIA have a peakaround Bt = 1.07 T (corresponding to fo/(2fcO) = 1.0) and increase with Bt for
B, > 1.25 T. Figure l(c) shows the changes in A, /3°IA and |3pl + £; as a functionof the toroidal magnetic field, where (i°1A is the perpendicular component estimatedfrom the diamagnetic measurement, and |3p, + ^ is obtained from the relation
AA = A (|8 , + fj )/2. Most of AA is likely to result from the enhanced Pp, since
the change in lv during the ECH pulse duration of 50 ms, which is much shorter thanthe current diffusion time, tdiff > 1 s, can be neglected. For B, = 1.07 T, A/?™* isfound to be approximately equal to A/3p,, while AjS^ is about six times as large as
Aj8pn for Bt = 1.4 T. This result demonstrates selective perpendicular heating of thefast electrons for Bt = 1.4 T. Therefore, taking account of anisotropic plasma pressure, we obtain Pab = 37 kW and a resulting absorption efficiency of r)ab = Pab/
PECH = 0.15 atB, = 1.4 T.
Figures 2(a), (b) and (c) show the time evolution of the soft X-ray photon counts
integrated over different energy ranges for three viewing angles, 0SX = 0°, 90° and
180°, respectively, at Bt = 1.34 T. The X-ray radiation parallel (0SX = 0°) and perpendicular (0SX = 90°) to the magnetic field is measured simultaneously by two(c)

0.4 0.5 0.6 0.7 0.8 0.9 0.4 0.5 0.6 0.7 0.8 0.9 0.4 0.5 0.6
TIME

0.7 0.8

0.9(s)FIG. 2. Time evolution of soft X-ray photon counts integrated over different energy ranges for threedifferent viewing angles: (a) 0 K = 0°, (b) 6SX = 90°, (c) 6SX = 180°. B, = 1.34 T, Ip = 100 kA,
ne = 2 x 1018 nr3, PLH = 100 kW, PECH = 100 kW.

IAEA-CN-47/F-II-S

549

intrinsic germanium detectors with pulse height analyser systems. The forwardbackward (0SX = 180°) radiation is also measured by changing the direction of theplasma current. It is seen that ECH produces remarkable enhancement of the perpendicular radiation, appreciable increase in the backward radiation and a little decreasein the forward radiation. This result indicates selective EC perpendicular heating of
the energetic electrons but no enhanced absorption efficiency of the LH power, whichis consistent with the magnetic measurement mentioned above. The noticeable
increase in the photon counts in the backward radiation as well as the decrease in the
low energy component in the forward radiation during the ECH pulse suggests thegeneration of an inverse electric field, which decelerates the current carrying electrons and produces a backward stream of fast electrons. It is noted that an additional
current drive generates the inductive electric field. From the energy spectra obtained
from the data shown in Figs 2(a) to (c) we can infer enhancement of the fast electrons
having an energy of ~ 100 keV, which satisfies the relativistic EC resonance condition at the plasma centre.
Meanwhile, the Fisch and Boozer formula [7] predicts that the selective ECH
of fast electrons yields an additional current drive due to an asymmetric plasma
resistivity effect with high efficiency. From the calculation under the experimental
conditions as shown in Fig. l(a) [6], 90 kA of additional driven current, IECH, concentrated around the plasma centre and 28 kW of absorbed ECH power in a single
path are expected. Specifically, the current drive efficiency is rjCD = IECH/Pab = 3.2(A- W"1)- No appreciable additional current drive has, however, been observed
from the measurement of the loop voltage. The reason for this could be that the pulse
width of ECH has no sufficient duration for the inductive electric field to diffuse to
the plasma surface.
3.2. Current drive by fast waves
In this section, initial experimental results using a 200 MHz fast wave arepresented. Figure 3(a) shows the time behaviour of the relevant plasma parameters
for two typical discharge shots at ne = 1.5 X 1019 m 3 and Bt = 1.32 T. The solidline indicates the result of FW combined with ECH, and the dashed line represents
the result of independent application of ECH and FW pulses in a shot. The radiated
ECH and FW powers are PECH = 100 kW and PFW = 40 kW, respectively. The
value of A&PW is set at 135°, and the RF spectrum of the fast waves is thus calculated to have a mild peak around N, = 3 in the direction of the current carrying
electron drift. The working gas is deuterium, and the plasma current is kept constantat Ip = 200 kA. The present plasma parameters yield 1.1-5 for the accessible N, at
the plasma centre and r/a = 0.9 for the radial position of the LH resonance layer at
200 MHz. The combination of ECH and FW pulses is seen to make an appreciable
increase in the EC emissions with frequencies of 50.4 GHz and 78.4 GHz, which correspond to fr/(2fc0) = 0.68 and 1.06, respectively. Note that the emission at
50.4 GHz comes from non-thermal electrons since 50.4 GHz is not the electron

550

YAMAMOTO et al.

(b)

(a)
1.5

OH + ECH

_—
0.5

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)

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Te«=57keV

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0.8
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1

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t

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20
40
60
80
100
PHOTON ENERGY (keV ]FIG. 3(a). Time evolution of relevant plasma parameters with and without (solid and dashed lines) FW
power of 40 kW during ECH pulse. B, = 1.32 T, lp = 200 kA, PECH = 90 kW; (b) energy spectra ofsoft X-rays without and with (upper and bottom) FW power during ECH pulse. Experimental conditions
are the same as in Fig. 3(a).

cyclotron frequency in the plasma. The emission in the extraordinary mode wasmeasured in the horizontal direction with a calibrated heterodyne receiver. A little
increase in the loop voltage is also seen during the FW pulse, in both shots. At thesame time, an increase in the line intensity of metal impurities was observed.
We have also measured the soft X-ray radiation under the same conditions as
those of Fig. 3(a). The measured energy spectra of the forward radiation with andwithout FW pulse during the ECH pulse are shown in Fig. 3(b). The spectrum is time
averaged over the duration of the FW pulse and is superposed for two identical discharges. From this figure, we conclude that energetic electrons exist in the plasma
and the application of the FW pulse leads to a noticeable increase in the electron tailin the range of 40 to 100 keV.
The observed results indicate an appreciable interaction of fast waves and energetic electrons, leading to current drive at higher density. No indication of FWCD
has, however, been obtained from measuring the loop voltage. The reason for this
could be that the RF power and the shape of the RF spectrum are not sufficient toovercome the increase in loop voltage due to enhanced impurity flux. From the calculation [8], a driven current less than 10 kA is predicted in the present experimental
conditions assuming 1 % of the fraction of the energetic electrons with a temperatureof 40 keV. This would, by itself, result in a relative drop of the loop voltage of only
a few per cent.

IAEA-CN-47/F-H-5
4.

551

SUMMARY
The experimental results can be summarized as follows:

(1)

(2)

(3)

An absorption efficiency of i?ab = 0.15 of the ECH power is obtained at alargely down shifted cyclotron frequency of fo/(2fcO) = 0.76, when ECH isapplied into an LH current driven plasma.
The measurement of the soft X-ray radiation parallel and perpendicular to themagnetic field demonstrates selective electron cyclotron perpendicular heatingof the fast electrons, generated by the LH waves, with the parallel velocity satisfying the relativistic resonance condition.
It is observed from the measurements of the electron cyclotron emission and thesoft X-ray radiation that the 200 MHz fast waves launched from the phased twoloop antennas interact with the energetic electrons — in contrast to the LH
waves, at high densities.
ACKNOWLEDGEMENTS

The authors are grateful to Drs M. Tanaka, K. Tomabechi, and S. Mori for continuous encouragement, and to Dr. J.-M. Noterdaeme for fruitful discussions.
REFERENCES
[1]
[2]
[3]
[4]

[5]

[6]
[7]
[8]

FIDONE, I., et al., in Heating in Toroidal Plasmas (Proc. 4th Int. Symp. Rome, 1984),
Vol. 2, Int. School of Plasma Physics, Varenna (1984) 872.
KRIVENSKI, V., et al., Nucl. Fusion 25 (1985) 127.
SHOJI, T., et al., in Controlled Fusion and Plasma Physics (Proc. 11th Europ. Conf. Aachen,1983), Vol. 1, European Physical Society (1983) 55.
HOSHINO, K., et al., Design Concepts and Performance Tests of the 60 GHz Electron Cyclotron
Heating (ECH) System for the JFT-2M Tokamak, Rep. JAERI-M-85-169, Japan Atomic Energy
Research Institute, Tokyo (1985).
UESUGI, Y., et al., Coupling to Fast Waves via a Phased Loop Antenna Array for Fast Wave
Current Drive in the JFT-2M Tokamak, Rep. JAERI-M-86-103, Japan Atomic Energy ResearchInstitute, Tokyo (1986).
YAMAMOTO, T., et al., in Controlled Fusion and Plasma Heating (Proc. 13th Europ. Conf.
Schliersee, 1986), Vol. 10 C, Part II, European Physical Society (1986) 437.
FISCH, N.J., BOOZER, A.H., Phys. Rev. Lett. 45 (1980) 720.
YAMAMOTO, T., et al., Numerical Study of the Electron Heating and Current Drive by theFast Waves, Rep. JAERI-M-86-115, Japan Atomic Energy Research Institute, Tokyo (1986).

DISCUSSION
M. PORKOLAB: It appears to me that in Fig. 3(a), showing combined
ECH/FW pulses, the loop voltage increases when the fast wave is applied. This is552

YAMAMOTO et al.

perhaps due, as you say, to impurity injection. If so, the increased DC electric fieldin the plasma could enhance the electron tail and hence the ECH emission, rather than
the fast wave. In view of this, I am not certain that you have really demonstrated FW
current drive.
T. YAMAMOTO: It can be seen in Fig. 3(a) that the ECE does not increase,even when the loop voltage increases during the ECH pulse alone. In addition, no
increase in ECE accompanied by an increase in the loop voltage has been observed
when the FW power is applied to the plasma with n :£ 1 x 1019 nr3. These results
indicate that the increase in the ECE stems from FW heating and not from theincrease in the loop voltage.
No FW current drive has been observed from the measurement of the loop voltage. We have observed appreciable interaction between fast waves and fast electrons, which would lead to FW current drive.
T. CHO: During ECH, it is reasonable to expect the perpendicular soft X-rays
to increase, since the lobe of the X-ray inclines in the electron accelerated direction
owing to the relativistic effect. However, because of this relativistic effect, the X-ray
emission parallel to the magnetic field should be increased by LH current drive alone
in the energy range of the quasi-linear Landau-damped resonant particles. How
would you interpret the only slight change in the X-ray emission parallel to the magnetic field during slow-wave LHCD with the parallel refractive index Ns = 2?
T. YAMAMOTO: In the present discharge, the target plasma already includes
many runaway electrons and most of the soft X-ray radiation comes from these.
When LH power is applied to the plasma, the LH current drive causes a drop in theloop voltage which reduces the production rate of the runaway electrons, while the
LH waves accelerate the resonant electrons. As a result, the amount of runaway electrons does not change. Thus, the LH pulse does not contribute to an appreciableenhancement of the soft X-ray radiation. In the case of a discharge which includes
a few fast electrons, a remarkable increase in soft X-ray radiation has been observed.

IAEA-CN-47/F-II-6

ELECTRON CYCLOTRON AND
LOWER HYBRID CURRENT DRIVE EXPERIMENTS
IN THE WT-2 AND WT-3 TOKAMAKSS. TANAKA, Y. TERUMICHI, T. MAEKAWA,
M. NAKAMURA1, T. CHO2, A. ANDO, K. OGURA,
H. TANAKA, M. IIDA, S. IDE, K. OHO, S. OZAKI,R. ITATANI3, H. ABE3, Y. YASAKA3, M. FUKAO3,
K. ISHII3, A. MOHRI4, S. KUBO4
Department of Physics, Faculty of Science,
Kyoto University, Kyoto,Japan
Abstract
ELECTRON CYCLOTRON AND LOWER HYBRID CURRENT DRIVE EXPERIMENTS IN THE
WT-2 AND WT-3 TOKAMAKS.
An electron cyclotron (EC) driven current is generated in two types of toroidal plasmas: anohmically heated (OH) plasma and a lower hybrid (LH) driven current sustained plasma. In the former
case, by injecting microwave P EC near the EC frequency into an OH plasma, after OH power has been
shut off in the WT-2 tokamak, the plasma current is sustained and ramped up by the EC wave only,
without OH power. Here, the EC driven current is generated by EC heating of suprathermal electron
beam in the OH plasma. In the latter case, when P EC is injected into the LH driven current sustained
plasma, the plasma current Ip and its ramp-up rate increase, and the increased Ip is kept after PEC isshut off. Here, the EC driven current is generated by EC heating of the mildly relativistic electrons in
the LH driven plasma. On the other hand, the current Ip and its ramp-up rate decrease when the bulk
electrons are EC heated. This selective EC heating of the bulk or the mildly relativistic electrons is
performed by changing the radial position of EC resonance layer in the plasma column. Further, theelectron beam (EB) is injected by using the plasma anode method. EB injection into a currentless ECR
plasma forms a toroidal current, which is nine times the injected beam current.

1.

INTRODUCTION

Recently, experiments on non-inductive current drive (CD) have shown that theplasma current is started up, ramped up and sustained by the LH wave in many
tokamaks [1]. Meanwhile, other non-inductive CD methods have been suggested
theoretically. Fisch et al. [2] have predicted that the plasma current can be produced
by quasi-linear cyclotron damping of the EC wave. This current drive, ECCD, isachieved by creating an asymmetric electron distribution function along the field,
Osaka Institute of Technology, Osaka, Japan.
Plasma Research Center, University of Tsukuba,
Tsukul Ibaraki, Japan.
3
Faculty of Engineering, Kyoto University, Kyi
Kyoto, Japan.
Institute of Plasma Physics, Nagoya University, Nagoya, Japan.
2

553

554

TANAKA et al.

whose mechanism is quite different from LHCD. Here, we report on experiments inthe WT-2 tokamak, in which the plasma is sustained by the EC driven current only,
without OH power. Furthermore, the plasma current is produced by EC heating of
the mildly relativistic electron beam present in the LH driven current sustained
plasma. In addition, an experiment on plasma current drive by injection of an electron
beam is presented.
The WT-2 tokamak has an aluminium shell and an iron core, major and minor
radii of RQ = 40 and a = 9 cm, respectively, a toroidal field of BT < 15 kG, anda plasma current of Ip < 30 kA [3,4]. A microwave from a gyrotron(W/2TT = 35.6 GHz, PEC < 120 kW, pulse width t < 10 ms) is fed from the low
field side through circular waveguides to a Vlasov antenna with parabola reflector
placed along the major radius. By rotating the antenna around the guide axis, alinearly polarized electromagnetic wave (EMW), propagating parallel or antiparallelto the field BT with an angle of ±48° to BT, is emitted as the extraordinary(X) mode. To achieve LHCD, RF from a magnetron (0.915 GHz, PLH < 100 kW,
t < 20 ms) is launched into the plasma with a phased array of four waveguides (innerdimension, 13 X 185 mm2).
The operation of the new toroidal machine WT-3 (R = 65 cm, a = 22 cm andBT = 1.7 T) was started in June 1986 for a study of RF heating and current drive.
The RF sources are 200 kW, 56 GHz for ECH and 350 kW, 2 GHz for LHH and
500 kW, 12-30 MHz for ICH.

2.

ELECTRON CYCLOTRON CURRENT DRIVE AFTER OH PLASMA [5]In Figs l(a) to (g), the temporal evolution of plasma parameters with andwithout PEC is plotted. First, a low density slide-away discharge (dotted curves)
with bulk electron density % = 2 x 1012 cm" 3 and temperature Te = 70 eV, is
produced by an OH discharge with EC resonance preionization. When PEC isinjected after the primary voltage V, of OH transformer has been short-circuited,
the plasma parameters change drastically, and the constant plasma current Ip
continues to flow with a loop voltage of VL = 0 as long as PEC is applied (fullcurves). The data show that the plasma is sustained by the EC driven current only
since no OH current flows for VL = 0. It is noticeable that in this ECCD plasmastrong hard X-ray IHX appears and both the soft X-ray I s x (0.85 keV) and the second
electron cyclotron emission L. (68 GHz) increase to more than ten times the values
in the OH plasma. The result suggests that a suprathermal electron beam carrying theEC driven current is generated. After PEC has been turned off, VL changes abruptly
to a positive value, and Ip falls off in a short time. Here, IHX increases strongly,
suggesting that the suprathermal electrons are accelerated by the positive VL. On theother hand, the pyroelectric signal Ipyro and the light intensity of impurity line
IL (O II - O V) decrease gradually, suggesting that the temperature of bulk
electrons is nearly the same as that in OH plasma.

555

IAEA-CN-47/F-II-6

10
20
Time (ms)1020Time (ms)F/G. /. Temporal evolution of (a) loop voltage VL, (b) plasma current Ip, (c) electron density he,(d) hard X-ray emission IHX (20 keV), (e) soft X-ray Isx (0.85 keV), (f) electron cyclotron emission lf(68 GHz), and (g) pyroelectric signal lpyro. Full curves are with ECHpower, Pec (parallel injection)and dotted curves without PEC. p = 2 x 10 torr in H2, additional gas comes from puffing during
injection of PEC, BT = 12 kG.

40

PECH
•(c)/

V

(kW)

~> ECH L-

"

: /

OFF:

0

50
100
150
Photon Energy(keV)

8
12
Time(ms)

16

FIG. 2. (a) X-ray energy spectra emitted from plasma as shown in Fig. I, during: (1) time t = 0-5 ms(startup stage), (2) t = 5-10 ms (OHstage) and (3) t = 70-75 mi (ECCD stage); (b) (1) loop voltage
VL and (2) ramp-up rate as a function ofPEC; (c) EC driven current ramp-up, flat-top and decrease for(1) PEC = 70 kW, (2) 50 kW, (3) 30 kW.

556

TANAKA et al.

The X-ray energy spectra emitted from ECCD as well as OH plasmas aremeasured, and plotted in Fig. 2(a). The slope of the X-ray spectrum is about 15 keV,and the energy extends to 150 keV in the ECCD plasma. Thus, it is concluded thatin the ECCD plasma the electrons are composed of bulk electrons withTe ~ 70 eV, fie ~ 0.5 x 10 12 cm" 3and suprathermal electrons withTe ~ 15 keV, n,, ~ 1010 cm" 3 , which are carrying Ip ~ 3 kA. In addition, the factthat SX-rays up to 100 keV are emitted from the initial OH plasma suggests thatweakly suprathermal electrons are present and are necessary for the formation of
ECCD plasma.
As PEc increases, the current Ip due to ECCD changes from a decreasing
curve to flat-top and then to ramp-up as is shown in Figs 2(b) and (c).
Correspondingly, the voltage VL changes from a positive value to a negative one.
This behaviour shows that Ip is fully sustained by ECW. These flat-top discharges
due to ECCD, without OH power, are also obtained in WT-3, where Ip = 10 kA
for ne ~ 1012 cm" 3 , BT = 1.6 T, and PEC = 70 kW.
The effect of ECH is examined for various BT and ECW injection methods.
The flat-top or ramp-up discharge is obtained when the ECR layer is located on the
inner side of the torus. Inefficient ECCD under strong BT may correspond to thetheoretical prediction that the EC heated electrons become trapped on the outer sideof the torus. The ECCD efficiency with ECW, propagating parallel to the electron
drift, is better than that with ECW, propagating antiparallel or perpendicular to B T ,and ramp-up discharges are only formed in the former case. Radial profiles of I s x
have rather broad peaks near the plasma centre, and fie at peaks is somewhat
smaller than co. It is concluded that plasma current is generated by EC heating of thesuprathermal electron beam formed in the initial OH plasma. Such a unidirectional
electron beam may eliminate the cancelling effect of EC driven current on the
opposite side.
The above mentioned ECCD mechanism is also demonstrated in the microwave
discharge at ECR in WT-2, where a weak current Ip is generated [6]. Here, thetoroidal drift of electrons circulating in one direction is cancelled by superposing a
vertical field B v , and an electron beam is formed. The EC driven current isgenerated in the direction determined by B v , independent of the propagation
direction of ECW.

3.

ELECTRON CYCLOTRON CURRENT DRIVE IN RF TOKAMAK [7]In contrast to the OH plasma, in the LHCD plasma, high energy electrons in
the mildly relativistic range, which carry the current Ip, are produced. Highly efficient ECCD is expected in the LHCD plasma, since ECW is strongly absorbed bysuch high energy electrons and their collisionality decreases further. As shown in thedotted curves in Figs 3(a) to (k), when PLH is injected into an initial ECR plasma,
the current Ip starts up rapidly and attains Ip = 4 kA at the end of PLH and the557

IAEA-CN-47/F-II-6

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10
20
30
Time (ms)

(k)
10
20
30
Time (ms)FIG. 3. Temporal evolution of (a) plasma current Ip, (b) loop voltage VL, (c) electron density ne,(d) electron cyclotron emission 1^, (e) and (f) soft X-ray emission Isx, (g) pyroelectric detector signal
Ipyro, and (h) to(k) impurity line radiations O II, III, IV, V, respectively. Full curves are with PEC anddotted curves without PEC. The ECR layer is set at r = rECR = +8.2 to +6.3 cm, changing during
ECHpulse according to BT change in time (BT0 = 15.3-14.7kG at r - 0 cm), PEC = 80 kW, andP
LH ~ 48 kW. p = 6 x 10's ton in H2; additional gas comes from puffing during P^.

!0JP

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50 100 150 200
Photon Energy (keV)

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.

1.2

FIG. 4. Radial profile of (a), (b) soft and (c) hard X-ray emissions at the end of ECH pulse (t = 5 ms)
in Fig. 3. Solid (dotted) curves are with (without) ECH; (d) soft and hard X-ray spectra with (solid) andwithout (dotted) ECH pulse, (e) ECR condition.

558

TANAKA et al.

tokamak plasma is formed by RF power alone, without Ohmic heating power(so-called RF tokamak [8]). When PEC is applied to the RF tokamak plasma theramp-up rate of current, AIp/At, increases and Ip attains a value higher than thatwithout PEC (solid curve) after termination of the PEc pulse. This increment of Ip isascribed to the EC driven current and the voltage VL, the direction of which is
opposed to that of electron drift, increases in proportion to AIp/At. The electron
density ne remains constant. The X-ray emission I sx (0.85 keV), IHX and nonthermal ECE 1^ increase drastically during ECH, while Ipyr0, I s x (0.2 keV) and IL(O II-O V) do not change or even decrease slightly. The results suggest that highenergy electrons are generated by ECH although there is no effect on the bulk
electrons.
Measurements show that this EC driven current is generated and the emissions
I,, and I sx (0.85 keV) increase strongly when BT is strong and the ECR layer(w = 0e) is on the outer side of the torus (rECR 5: 5 cm). In Fig. 4(a) the radial profile of I sx (0.2 keV) is plotted, which is very similar to the density profile of thebulk electrons. The radial profiles of I s x (0.85 keV) and IHX (20 keV), representing
the profile of high energy electrons, show peaks on the high field side of the ECR
layer. When ECH is applied to the RF tokamak, both I sx (0.85 keV) and IHX(20 keV) increase drastically near the peaks of their profiles without PEC, while
there is no change of I sx (0.2 keV). Here, the ECR layer is on the outer side ofplasma (rECR = 0.3 a), and the radial position of the peaks of I sx (0.85 keV) and
IHX is not at the ECR layer, but is shifted slightly towards the high field side, where
Oce = 1.07 co.
In the RF tokamak, high energy electrons are present, and strong X-rays up to
300 keV are emitted. The energy spectrum is composed of the steep slope part(—15 keV) in the low energy range and the gentle slope part (~45 keV) in the highenergy range as is shown in Fig. 4(d) [9]. When PEC is injected, the X-ray emission
increases by a factor of three in the high energy range of 20-200 keV, while there
is no change below 20 keV. Considering the relativistic effect and the Doppler shift,
the ECR condition is given by
1 - (N,v,)/c = ( O » {1 - (v? + v 2 ± )/c 2 }*As the electron density increases, the relativistic effect dominates the Doppler shift
and resonance occurs at fle/co > 1, even if N|/V| > 0. A calculation shows that in
the field region of 0e/a> = 1.07 ± 0.05 the energy of the resonant electrons is20-150 keV for N, = 0.23 ± 0.1 and 2-50 keV for N, = - (0.23 + 0.1), where
v2 = 2v2| is assumed (Fig. 4(e)). It is concluded that mildly relativistic electrons
generated by LHW and drifting along one direction of BT are accelerated efficiently
in the perpendicular direction by ECW, then these electrons become more collisionless and carry the EC driven current efficiently. The collision time of electrons of
50 keV is about 60 ms (fie = 2 x 10!2 cm" 3 , Zeff = 2) and much longer than the
ECH pulse. Thus, in Fig. 4(d), the enhancement of X-ray emission due to PEC in

559

IAEA-CN-47/F-II-6

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£50
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(e)
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20
30
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20
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FIG. 5. Temporal evolution of plasma parameters as in Fig. 3. Here, ECR layer is set atr= rECR = ~2-0 to -4.8 cm (BT0 = 12.1-11.2 kG).

the high energy range is maintained after PEC has been switched off. Correspondingly, the increased current Ip due to ECCD does not decay after PEC has beenswitched off in Fig. 3(a).
The change of Ip due to ECCD in RF tokamak is measured for various BT.
When BT is weak and the ECR layer is located on the inner side (rECR = - 0 . 3 a),the effects of ECH are quite different from those observed in Fig. 3. During the PEC
pulse, Ip continues to decrease, AIp/At becomes negative, and Ip is smaller than thedotted values after termination of PEC in Fig. 5(a). The loop voltage VL is zero orslightly positive. The electron density n,. increases. The increment of IM (68 GHz)and I s x (0.85 keV) is small, but I s x (0.2 keV) and Ipyr0 increase strongly during PEC.
The line intensities of low excitation energy (O II, O III) do not change or evendecrease slightly compared with the dotted curves, in spite of the increase of n,,, but
curves of high energy (O IV, O V) increase strongly. It is roughly estimated that thebulk electron temperature increases from 20 to 70 eV by comparing the line intensities with those of no ECH. This behaviour suggests that the heating of bulk electrons
occurs at Ue/u * 1 and no high energy electrons are produced. The dependence of
Ip due to ECCD on BT is similar for various methods of PEC injection: parallel,
antiparallel, perpendicular and top injection although the absolute value of AIp/At isdifferent. The reason why the LH driven current Ip decreases during bulk electron
heating is unknown although it may be ascribed to an enhancement of return current.

560
4.

TANAKA et al.

EFFICIENCY OF ECCD

kA/ms)In the ECR plasma, the dependence of EC driven current Ip on PEC andrij. is investigated, and the relation, Ip = ^PEC/I^R, is demonstrated experimentally. Then, the figure of merit of ECCD is obtained as 17 = 1.2 x 10 ~3(1019 m~2 kA-kW" 1 ). Furthermore, by using the experimental data of the flat-top
discharge due to ECCD after OH plasma in Fig. 2(c), the efficiency is calculated to
be r\ » 1 x 10 ~3. These values are small, compared with r) « 6 x 10" 2 for
LHCD in WT-2 [10]. The experimental value of 77 obtained from the flat-top
discharge due to ECCD is calculated to be 17 = 0.8 X 10" 2 , while 17 ~ 1 x 10"'for LHCD (Ip « 45 kA, PLH = 100 kW, ne = 2.6 x 1012 cm' 3 ) in WT-3.
According to the quasi-linear theory of ECCD, the efficiency is given byr) = 0.306 (j/p) (Te/10 keV)/ln A. The expression for normalized efficiency (j/p) is
given by Fisch et al. [2] and Cordey et al. [11]. Substituting the experimental valuesof resonant electrons, Te supra = 15 keV, and the bulk electrons, Te = 70 eV, weobtain 77 = 2 x 10~\ which is two orders larger than for the experimental valuesof WT-2.
Next, the efficiency of ECCD is examined for RF tokamak. In Fig. 6 is plotted
the ramp-up rate AIp/At as a function of PRF (PLH or PEC). For the RF tokamak
discharge (PEC = 0), as PLH increases the ramp-up rate increases at first, thenattains a constant value (curve 1). When a constant PEC is superposed, the ramp-up
rate increases with PLH (curve 2) and exceeds that with PEC = 0. When the RF
tokamak is sustained by a constant PLH, the current ramp-up rate increases with PEC
and attains a constant value (curve 3). Unfortunately, the present pulse of PEC is too
short for Ip to attain a quasi-stationary state. If it is assumed that the change of the
ramp-up rate with respect to PRF, d(AIp/At)/dPRF, is proportional to the efficiencyof RFCD, i7RF, then both rjEC with a constant PLH and TJLH with a constant PEC

(3).

0.2

(2) - ^ .

f

(1)-^
< O.V - - "
•
a.
A
<
4 8
W /

e

c

20

40

pRF
FIG. 6. Ramp-up rate as a junction ofPRF
and (3) Pw = 60 kW; BT = 15 kG.

60
(

kW)

in the cases of(l) PEC = 0, (2) PEC = 62 kW

IAEA-CN-47/F-II-6

561

are nearly equal to rjLH without PEC in the unsaturated region. It is suggested
theoretically that
~ (3/4) {1 +(v/2v l l ) 2 }- 1 ~ 1
which is consistent with the experimental results. As has been noted, when PEC issuperposed on the RF tokamak, the ramp-up rate exceeds that for PHL alone. Suchan enhancement of LHCD efficiency combined with ECH is consistent with thetheoretical predictions by Fidone et al. [12].

5.

CURRENT DRIVE BY INJECTION OF ELECTRON BEAM [13]By the plasma anode method, an electron beam (EB) is injected into a currentless ECR plasma in WT-2. The peak current of EB is Ik = 0.46 kA, and the pulse
width is 5 ms. The energy of EB decreases from 2 to 0.4 keV during the pulse. The
plasma current Ip reaches 4 kA at 0.7 ms after EB injection. Here, the current
amplification ratio Ip/Ik is about 9 at this peak. When EB is injected, the plasma
density increases considerably, n<. = (2.5 — 60) x 1012 cm" 3 , and the Ipyro radiation rises rapidly. This abrupt increase in ne and Ipyro is attributed to the inflow of
impurities produced at the cathode during EB injection. However, it should be notedthat Ip continues to flow during EB injection although the slowing down time of EB
is very short, e.g., TS ~ 0.1 jts in the case where n, = 6 x 1013 cm" 3 , Zeff = 5
and an EB energy of about 0.5 keV is assumed. As Bv is varied, the peak value of
Ip changes, and its maximum value is Ip = 4 kA at Bv = 1.6 x 10" 3 T for
BT = 1.13 T.

6.

CONCLUSIONSEC driven current is generated in two types of plasmas. First, such a current
is generated by ECH of suprathermal electrons in the OH plasma, and the current
Ip is sustained or ramped-up when PEC is injected into the plasma after OH powerhas been switched off. This EC driven current is generated in the ECR plasma.
Secondly, EC driven current is generated by ECH of the mildly relativistic electron
beam present in LH driven current sustained plasma, and the ramp-up rate of Ip
increases when PEC is injected into RF tokamak. Such an enhancement of LHCD
efficiency combined with ECH is quite important for a new, useful and efficientmethod of RFCD.
When an electron beam of moderate energy, 0.4 to 2 keV, is injected into acurrentless ECR plasma with density 2.5 x 1012 cm" 3 at BT = 1.1 T, a toroidal
current of 4 kA, which is nine times the injected beam current, is formed.

562

TANAKA et al.

REFERENCES
[1]
[2]
[3]
[4]
[5]
[6]
[7]
[8]
[9]
[10]
[11]
[12]
[13]

Non-inductive Current Drive in Tokamaks (Proc. IAEA Tech. Comm. Meeting, Culham, 1983),
Vols 1 and 2, Culham Laboratory, Abingdon (1983).
FISCH, N.J., BOOZER, A.H., Phys. Rev. Lett. 45 (1980) 720.
TANAKA, S., et al., in Plasma Physics and Controlled Nuclear Fusion Research 1982 (Proc.
9th Int. Conf. Baltimore, 1982), Vol. 1, IAEA, Vienna (1983) 199.
TANAKA, S., et al., in Plasma Physics and Controlled Nuclear Fusion Research 1984 (Proc.
10th Int. Conf. London, 1984), Vol. 1, IAEA, Vienna (1985) 623.
ANDO, A., et al., Phys. Rev. Lett. 56 (1986) 2180; J. Phys. Soc. Jpn. 56 (1986).
SHIMOZUMA, T., et al., J. Phys. Soc. Jpn. 54 (1985) 1360.
ANDO, A., et al., Nucl. Fusion 26 (1986) 107.
KUBO, S., et al., Phys. Rev. Lett. 50 (1983) 1994.
OGURA, K., et al., J. Phys. Soc. Jpn. 55 (1986) 13.
NAKAMURA, M., et al., J. Phys. Soc. Jpn. 53 (1984) 3399.
CORDEY, J.G., et al., Plasma Phys. 24 (1982) 7.
FIDONE, I., GIRUZZI, G., GRANATA, G., MEYER, R.L., Phys. Fluids 27 (1984) 2486.
TANAKA, H., et al., Jpn. J. Appl. Phys. 24 (1985) 644.

DISCUSSION
A.C. RIVIERE: In the case of electron cyclotron current drive, did you makean experiment with the antenna in the opposite toroidal direction? If so, what was thedifference in current between the two antenna orientations?
S. TANAKA: Yes, we did perform such an experiment. When the antenna wasin the opposite toroidal direction, we could not obtain a flat-top or ramp-up
discharge, although a weak EC driven current did appear.

IAEA-CN-47/F-II-7

IMPROVEMENT OF
CURRENT DRIVE AND CONFINEMENTBY COMBINATION OF
LHCD AND NBI HEATING IN JT-60
JT-60 TEAM
(Presented by T. Imai)
T. Abe, H. Aikawa, H. Akaoka, H. Akasaka, M. Akiba, N. Akino, T. Akiyama,
T. Ando, K. Annoh, N. Aoyagi, K. Arakawa, M. Araki, K. Arimoto, M. Azumi,
S. Chiba, M. Dairaku, N. Ebiswa, T. Fujii, T. Fukuda,
H. Fumkawa,
K. Hamamatsu, K. Hayashi, M. Hara, K. Haraguchi, H. Hiratsuka, T. Hirayama,
S. Hiroki, K. Hiruta, M. Honda, H. Horiike, R. Hosoda, N. Hosogane, Y. Iida,
T. Iijima, K. Ikeda, Y. Ikeda, T. Imai, T. Inoue, N. Isaji, M. Isaka, S. Ishida,
N. Itige, T. Ito, Y. Ito, A. Kaminaga, M. Kawai, Y. Kawamata, K. Kawasaki,
K. Kikuchi, M. Kikuchi, H. Kimura, T. Kimura, H. Kishimoto, K. Kitahara,
S. Kitamura, A. Kitsunezaki, K. Kiyono, N. Kobayashi, K. Kodama, Y. Koide,
T. Koike, M. Komata, I. Kondo, S. Konoshima, H. Kubo, S. Kunieda,
S. Kurakata, K. Kurihara, M. Kuriyama, T. Kuroda, M. Kusaka, Y. Kusama,
S. Maehara, K. Maeno, S. Mase, S. Matsuda, M. Matsukawa, T. Matsukawa,
M. Matsuoka, N. Miya, K. Miyati, Y. Miyo, K. Mizuhashi, M. Mizuno, R. Murai,
Y. Murakami,
M. Muto,M. Nagami,
A. Nagashima,
K. Nagashima,
T. Nagashima, S. Nagaya, H. Nakamura, Y. Nakamura, M. Nemoto, Y. Neyatani,
S. Niikura, H. Ninomiya, T. Nishitani, H. Nomata, K. Obara, N. Ogiwara,
Y. Ohara, K. Ohasa, T. Ohga, H. Ohhara, M. Ohkubo, T. Ohshima, K. Ohta,
M. Ohta, M. Ohtaka, Y. Ohuchi, A. Oikawa, H. Okumura, Y. Okumura,
K. Omori, S. Omori, Y. Omori, T. Ozeki, M. Saigusa, K. Sakamoto, A. Sakasai,
S. Sakata, M. Satou, M. Sawahata, M. Seimiya, M. Seki, S. Seki, K. Shibanuma,
R. Shimada, K. Shimizu, M. Shimizu, Y. Shimomura, S. Shinozaki, H. Shirai,
H. Shirakata, M. Shitomi, K. Suganuma, T. Sugie, T. Sugiyama, H. Sunaoshi,
K. Suzuki,
M. Suzuki,
M. Suzuki,
N. Suzuki,
S. Suzuki,
Y. Suzuki,
H. Takahashi1, M. Takahashi, S. Takahashi, T. Takahashi, M. Takasaki,
M. Takatsu, A. Takeshita, H. Takeuchi, S. Tamura, S. Tanaka, K. Tani,
M. Terakado, T. Terakado, K. Tobita, T. Tokutake, T. Totsuka, N. Toyoshima,
H. Tsuda, T. Tsugita, S. Tsuji, Y. Tsukahara, M. Tsuneoka, K. Uehara,
M. Umehara, Y. Uramoto, H. Usami, K. Ushigusa, K. Usui, J. Yagyu,
K. Yamada, M. Yamamoto, O. Yamashita, Y. Yamashita, K. Yano, T. Yasukawa,
K. Yokokura,
H. Yokomizo,
H. Yoshida,
M. Yoshikawa,Y. Yoshinari.
R. Yoshino, I. Yonekawa, K. Watanabe

Naka Fusion Energy Research Establishment,
Japan Atomic Energy Research Institute,Naka-machi, Naka-gun, Ibaraki-ken,
Japan

1

Princeton Plasma Physics Laboratory, Princeton University, NJ, USA.

563

564

JT-60 TEAM
Abstract

IMPROVEMENT OF CURRENT DRIVE AND CONFINEMENT BY COMBINATION OF LHCD
AND NBI HEATING IN JT-60.
First results of the lower hybrid current drive (LHCD) experiment on JT-60 are presented, inwhich maximum RF driven currents of 1.7 MA with an RF power of 1.2 MW at 2 GHz are obtainedfor 2 s. High current drive efficiency, JJ CD = ne [1019 m~3]R[m]IRF[MA]/PLH[MW] = 1-1.7 forOhmic plasma is reliably attained. An enhancement of )j CD is obtained with the combination of LHCD
and NBI heating and a value of ?jCD = 2-2.8 is reached. Improvement in confinement because of thesynergetic effect of combined LHCD and NBI is also observed. The energy confinement time during
NBI heating is not degraded, compared with that in Ohmic discharges.

1.

INTRODUCTION

Improvement in current drive efficiency and reduction of confinement degradation during auxiliary heating are crucial problems of the next large tokamak aiming
at a steady state tokamak reactor. Recent progress in current drive, heating andsuppression of MHD activities making use of RF power makes us hope that we shallbe able to overcome these problems. As a result of successful heating and current
drive experiments in the Lower Hybrid Range of Frequencies (LHRF) on the JAERI
tokamaks (JFT-2, JFT-2M) [1-3], a scaling-up programme with high power LHRF
heating from 24 MW source power has started in JT-60. The major objectives of the
JT-60 LHRF system are the production of a reactor grade plasma with a combinationof 20 MW Neutral Beam Injection (NBI) and 6 MW Ion Cyclotron Range ofFrequencies (ICRF) heating, and current drive in low and medium density plasmas,
in order to ascertain the physics feasibility of stationary or quasi-stationary operationof the tokamak reactor.
In this paper, the first results of lower hybrid current drive in Ohmic and NBI
heated plasmas in JT-60 will be described, in which enhancement of current drive
efficiency by combination of LHCD and NBI heating and an improvement ofconfinement are obtained.

2.

LHRF SYSTEM

Since the performance of JT-60 and the NBI heating system are described incompanion papers [4, 5], only a brief description of the LHRF system will be given
here. The LHRF heating system produces 24 MW RF power at about 2 GHz(1.7-2.23 GHz) with three units of the eight klystron amplifier subsystems, where
the 1 MW klystrons developed in JAERI are used. The line-up of the amplifier chain
is shown in Fig. 1 (a). One of three 4 X 8 phased array launchers, which are installedin JT-60 at present, is designed for current drive and has additional passive
waveguides [6]. The results obtained here are from this launcher, whose N| spectra,

565

IAEA-CN-47/F-II-7

S.G.

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from Brambilla theory (waveguide width b = 16 mm).

from Brambilla theory, are shown in Fig. l(b). The peak of the Nj spectrum is 1.7,
with a phase difference of adjacent waveguides of A<j> = 90°, and an efficient
coupling to the high energy electrons of > 100 keV is expected. The radial positionof the launcher front is usually at r = 945-950 mm.

3.

CURRENT DRIVE IN OHMIC PLASMA

The current drive effect in Ohmic plasma is reliably obtained in JT-60, when
the lower hybrid wave is coupled unidirectionally to the low density plasma. A
typical time evolution of loop voltage V,, plasma current Ip, Ohmic primary

566

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567

IAEA-CN-47/F-II-7
2.0

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(IO l 9 m 3 )FIG. 3. Density scan of RF driven current normalized by RFpower, where Igp = /current IF, electron line average density ft,, and A (= j3p+<S/2-l) in a discharge
with RF power PLH = 1 MW is shown in Fig. 2(a)-(e), where /3P and £ are thepoloidal beta and the internal inductance, respectively. Two RF pulses from 3 to
3.5 s and from 4 to 8.0 s are injected during one shot. The V, drops sharply tozero or negative values at the RF pulses, and IF stays constant. The RF driven currents (IRF) as estimated from the drop of the loop voltage (AV,) at the second
pulse are: IRF = IplAVj/V] = 1 MA. The current drive efficiency as definedby r?CD = fle [1019 irr3]R[m]IRF[MA]/PLH[MW] is given by IJCD = 1. from IRF
and Fig. 2.
The change in A of Fig. 2(d) suggests a reorganization of the current distribution j(r) during LHCD, since ^ is dominant in A because of the low density. A at theend of RF pulse is lower than that at t = 3.0 s. Therefore, it is indicated that j(r)is broadened by LHCD. The effect of j(r) flattening due to LHCD will be discussed
later, in Section 5, in connection with MHD activities and confinement.
Recharging effects of the OH primary coil are clearly seen in a discharge withsmaller plasma current and lower density as shown in Fig. 2(f, g). IF > 0 is clearlyobserved during the RF pulse. Current sustainment without Ohmic supply is also
demonstrated in Fig. 2(h). RF currents of 0.7 MA are only sustained by RF powerof 0.76 MW, during 2 s. The maximum RF current obtained so far is 1.7 MA, as
will be shown in Fig. 5(a).
The density dependence of IRF ( = LjAVj/V,) is shown in Fig. 3. IRF isroughly proportional to 1/n,,, as usual, since our present density regime is lower

568

JT-60TEAM

than the density limit. It may be stated that an efficient current drive of
ijCD = 1-1.7 in an Ohmic target plasma is obtained although the estimation of the
RF currents (IRF) contains some error bar due to the change in ^ (roughly 20%).
These values are appreciably larger than previous results obtained in other small
and medium size tokamaks. The theoretical current drive efficiency is given
by 7?™ = K^fg [7], where p = 1/N? (1 - N?/Ni)/log (N2/N,), N, and N2 being
accessibility limit of N| and the upper limit of the N| spectrum of radiated power;
f is a directivity factor, g is the portion of the accessible wave, and K is a constantfactor if Zeff is constant. The factors /t, f and g are not very different among various
tokamaks if the NB spectrum is optimized for current drive. Therefore, the reason
for our improved efficiency lies elsewhere. The effect of electron temperature does
not explicitly appear in r;TH. However, since we need some secondary mechanism tofill up the spectral gap between the wave phase velocity and the thermal velocity inorder to explain the effective current drive results obtained previously in many
tokamaks, this secondary effect necessarily brings about some reduction in
efficiency. Therefore, the electron temperature must play an important role in determining the efficiency, because the higher the electron temperature, the smaller the
spectral gap if we optimize fi. From this conclusion, we can achieve a more efficientcurrent drive in higher temperature plasmas, which can be realized by NBI heating.

4.

IMPROVEMENT OF VcD WITH NBI HEATING

As was discussed in Section 3, higher current drive efficiency is expected in
NBI heated plasma. Figure 4 shows the time evolution of the relevant plasma
parameters in a typical LHCD + NBI shot, where the RF and NBI powers are 0.88
and 8.8 MW, respectively, and Ip = 1 MA. The IRF current is estimated in thesame manner as that in the case of an Ohmic target plasma. Here, the increment of
the loop voltage after the RF pulse, while the NBI power is still on, is used as AV,.
An interesting feature shown in Fig. 4 is that the loop voltage drop ratio (AV,/V,)
does not change much before and after NBI is turned on, while tie increases by morethan a factor of 2.5. As is seen from the figure, IAV,|/V, = 1.25, because of
LHCD before NBI, and 0.75 during the NBI pulse and, hence IRF = 1.25 and0.75 MA, in the respective cases. Using ft,. = 0.35 X 1019 m" 3 (OH + LHCD),
0.9 x 1019 m" 3 (LHCD + NBI) and R = 3.1 m, ijCD = 1.5 in an Ohmic plasmaand 2.4 in an NBI heated plasma are obtained. The estimated RF current, IRF, forvarious cases is plotted in Fig. 5(a), where the currents in an Ohmic target plasma
are also plotted. As is seen from the figure, rjCD, for NBI + LHCD, is 2-2.8, while
for OH + LHCD, it is 1-1.7. The efficiency is almost doubled by NBI. This efficiency enhancement is also supported by measurements of the hard X-ray intensity,
IHX, and the electron cyclotron emission intensity, IECE(3fce). The linear correlation
between IRF and the hard X-ray increment during LHCD AIHX is obtained from
Fig. 5(b). Figure 5(c) shows the transition of AIHX from OH + LHCD to569

IAEA-CN-47/F-II-7
2
:

i

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PNBI = 8.8 MW

f \ = 0.88MW NBI V

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LHCD + NBI heating.

LHCD + NBI in the same shot. Since AIHX <x IRF and IRP OC 1/n., AIHX oc 1/n,. if
RF power and efficiency are constant. But the data in Fig. 5(c) indicate that AIHX isnot simply proportional to 1/fl,,, when LHCD is combined with NBI heating. The
relative enhancement of AIHX due to NBI heating is seen when the data before andafter NBI in the same shot are compared. A similar effect is observed in AIECE.
The above-given result is very important and promising for a future applicationof RF current drive and makes steady state tokamak operation a hopeful device.

5.

IMPROVEMENT OF CONFINEMENT

As was described in Section 2, the j(r) in the LHCD plasma seems to be broader
than that in the OH plasma. Therefore, improved confinement due to current profile
control by LHCD is expected. Actually, sawtooth oscillation and/or m = 1 mode arestabilized during LHCD, as is also observed in other, smaller tokamaks. The effectshould be applied to NBI heated plasmas.
Figure 6 shows the time evolution of A and other relevant parameters of the
OH + NBI shot (Fig. 6(a)) and the LHCD + NBI shot (Fig. 6(b)), where
Ip = 1.0 MA and BT = 4 T with NBI power of PNBI = 3.4 and 4.4 MW, respectively. The total stored energies estimated from the A for both shots are570

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P

OH

FIG. 6. Time evolution of V,, A and ncl in case of (a) OH + NBI with Pm, = 3.4MW and(b) LHCD + NBI with Pw = 7 MW, PNBI = 4.4 MW. (c) LHCD + NBI with PNBI = 11 MW.
Energy confinement time with LHCD + NBI is not degraded from Ohmic values (d).

572

JT-60 TEAMW* [OH + NBI] = 0.26 MJ and W* [LHCD + NBI] = 0.4 MJ, assuming A^
due to NBI — 0. A substantial enhancement of the stored energy by the combinationof LHCD and NBI heating is found from these values. The confinement time for theincremental stored energy during NBI heating increases by 20-30% in the LHCD
plasma. A remarkable increase in stored energy is obtained with the higher NBI
power of 11 MW in the LHCD plasma, and W* reaches 1.1 MJ, from Fig. 6(c).
Surprisingly, the global energy confinement time is approximately equal to that in the
Ohmic discharge and is 2.5 times larger than that from Kaye-Goldston scaling. The
confinement is not degraded by high power NBI (PNBI > POH) m LHCD plasma,even if hydrogen is dominant. The particle confinement is also improved, sincedensity profile peaking is observed in the LHCD + NBI discharge [5]. These profile
reorganizations of j(r), n^r) and possibly Te(r) in the LHCD plasma seem to play
an important role in confinement and make a new H-mode regime using RF current
drive a very promising device.

6.

CONCLUSIONSUp to 1.2 MW of LHRF power is launched unidirectionally in toroidal direction
into Ohmic and NBI heated plasmas in JT-60. Our results may be summarized asfollows:(1) Maximum RF currents of 1.7 MA are obtained.
(2) Efficient RF current drive (T;CD = 1-1.7) is obtained in Ohmic plasmas.
(3) A substantially improved current drive efficiency (T;CD = 2-2.8) is observed inplasmas with high power NBI heating, due to the increase in electron
temperature.
(4) The degraded energy confinement in NBI heated plasmas is improved by
LHCD.
Although more detailed studies should be done in the future, the resultspresented in this paper are very important and promising with respect to a future large
tokamak. In particular, items (3) and (4) seem to be highly relevant to a steady state
tokamak reactor.

ACKNOWLEDGEMENTS
The authors wish to thank the members of JAERI who have contributed to the
JT-60 project. They also wish to express their gratitude to Drs S. Mori, Y. Iso, andK. Tomabechi for continued encouragement and support.

IAEA-CN-47/F-II-7

573

REFERENCES
[1]
[2]
[3]
[4]
[5]
[6]
[7]

NAGASHIMA, T., FUJISAWA, N., in Heating in Toroidal Plasmas (Proc. 1st Joint VarennaGrenoble Int. Symp. Grenoble, 1978), Vol. 2 (1978) 281.
IMAI, T., et al., in Radiofrequency Plasma Heating (Proc. 4th Top. Conf. Austin, TX, 1981),
American Institute of Physics, New York (1981).
YAMAMOTO, T., et al., Phys. Rev. Lett. 45 (1980) 716.
YOSHIKAWA, M., et al., Paper IAEA-CN-47/A-I-1, these Proceedings, Vol. 1.
JT-60 TEAM (presented by NAGAMI, M.), Paper IAEA-CN-47/A-II-2, these Proceedings,
Vol. 1.
NAGASHIMA, T., et al., JT-60 radiofrequency heating system: Description and R&D results,
to be published in Nucl. Eng. Des./Fusion.
FISCH, N.J., Phys. Rev. Lett. 41 (1978) 873.

DISCUSSION
F.X. SOLDNER: The confinement times are derived from A, the sum ofpoloidal equilibrium beta and internal inductance. Can you rule out an increase in theinternal inductance during combined LHCD + NBI in the case of lower NBI power
for which you have shown A curves for comparable cases of NBI alone and for
NBI + LHCD?
Further, what kind of enhancement of the energy confinement is left if you
include the injected LH power, i.e. if you scale plasma energy versus total power
input?
T. IMAI: We cannot completely rule out a difference in A A between caseswith NBI + LHCD and NBI alone, for medium NBI power, as seen in Fig. 6(a)and (b), since no kinetic measurements were made because of the low density. In
order to avoid this ambiguity, we conducted a higher power experiment and obtained
the result shown in Fig. 6(c), where an improvement in confinement can be clearly
seen. In addition, we did not observe any evidence of an abrupt increase in internal
inductance. If it occurred, a substantial increase in loop voltage should also have been
observed.
As to your second point, in our TE calculation we disregarded both theabsorbed LH power and the increase in stored energy due to the LH power AWgH,
since we have not yet been able to determine AW£H- However, our estimate of TE
is valid if PLH -4 PNBi, as in our present experiments. If you compare theincremental energy confinement time, rfd = AW$B/PNB, you also find an enhancement of rl d in the LHCD plasma.
R.J. GOLDSTON: In contrast to the H-mode, one imagines that most of theextra stored energy in this regime lies in the centre of the discharge. The electron
temperature profile must be very impressive. Do you have any data on this that you
could show us?

574

JT-60 TEAM

T. IMAI: Unfortunately, our density regime in this series of experiments isstill low, since the available LH power is not yet high enough to drive a current ina high density plasma. So it is very difficult, for the time being, to produce atemperature profile by Thomson scattering.
K. LACKNER: I would like to recall that the TEXTOR results presented by
Dr. Vandenplas1 for fairly low density also seemed not to show any deterioration of
the additional heating, compared to Ohmic confinement times. Your results probably
fall into the same category. In your plot, comparing TE with NBI to NBI/LHCD, theformer cases appear to lie at higher densities. Was the change in particle confinement
time associated with the addition of LHCD?
T. IMAI: Yes, we observed an improvement in particle confinement for NBI
heated plasma with LHCD. This point was dealt with in the paper of Dr. Nagami2.

12Paper IAEA-CN-47/F-I-4, these Proceedings, Vol. 1.
Paper IAEA-CN-47/A-II-2, these Proceedings, Vol. 1.

HAEA-CN-47/F-III-2

CURRENT DRIVE, CONFINEMENT AND
INSTABILITY CONTROL EXPERIMENTS
WITH ECRH ON THE CLEO TOKAMAK
D.C. ROBINSON, N.R. AINSWORTH, P.R. COLLINS,
A.N. DELLIS, T. EDLINGTON, P. HAYNES, S. IYENGAR1,
I. JOHNSON, B. LLOYD, M. O'BRIEN, B.J. PARHAM,
A.C. RIVIERE, D.F.H. START, T.N. TODD
Euratom-UKAEA Fusion Association,
Culham Laboratory
Abingdon, Oxfordshire,
United Kingdom
Abstract
CURRENT DRIVE, CONFINEMENT AND INSTABILITY CONTROL EXPERIMENTS WITH
ECRH ON THE CLEO TOKAMAK.
Electron cyclotron resonance heating has been carried out on CLEO (R/a = 7, R = 0.9 m) using
the second harmonic extraordinary mode at both 28 GHz and 60 GHz to investigate current drive, confinement and instability control. With the 28 GHz source, at low collisionality (cc* - 0.01) and witha non-thermal population of electrons the confinement is enhanced relative to that in ohmic plasmas,
whereas with the 60 GHz source and ve* ~ 0.1 the confinement degrades with little non-thermal electron production. No evidence for current drive has so far been obtained at 28 GHz, but at 60 GHz adifferent loop voltage response was obtained when either the antenna direction or the plasma current
direction relative to the antenna pattern was changed, indicating a driven current of about 5 kA for
150 kW of absorbed power. Low mode number MHD activity has been controlled in a number ofexperiments, allowing an increase in the disruptive density limit and control of sawtooth activity.

INTRODUCTION
A series of experiments with 28 and 60 GHz ECRH has been
carried out on CLEO using a variety of different waveguidetechnologies and modes, to study confinement, current drive, MHD
control, preionisation and p-limits in tokamak, stellarator andhybrid configurations.
Current drive experiments have been carried out using
second harmonic extraordinary mode radiation launched from thelow field side at both 28 [1] and 60 GHz.
Considerable
development work has been aimed at producing high quality
antenna patterns at high power for the current drive experiments
at 60 GHz and recent results are detailed below.
Commonwealth Universities scholar.

575

576

ROBINSON et al.

Low-m MUD mode control has been achieved in a variety ofexperiments, including stabilisation of the 2,1 tearing mode toraise the disruptive density l i m i t and control of the 1,1 mode
affecting the sawtooth period and amplitude.

ENERGY CONFINEMENTFor CLEO tokamak the plasmas associated with 2o> ECRH at
28 and 60 GHz l i e in a large aspect ratio, low density regime
where conventional ohmic heating scalings (e.g- neo-Alcator)
predict significantly lower confinement time than those of
auxiliary heating, while the resulting data f a l l between thesepredictions and are therefore not easily categorised as simply
L-mode or ohmic. Energy confinement was generally better in thebi-Maxwellian (thermal plus 13 keV t a i l ) 28 GHz ECRH plasmas [2]than in their OH targets, even at absorbed power to i n i t i a lohmic heating power ratios ( p a h S /PoH^ °* > *°* ^ Orl< o n o t h e r
machines [3] has also correlated confinement enhancement with
distorted electron distributions. In the more thermal 60 GHz
ECRH cases, w i t h similar P . /P n H r a t i o s , the confinement
degrades and the data indicate a scaling of WTQT ~ Iapah witha < 0.8 and with the resonance in the plasma centre.
For 28 GHz, 2w heating,the f a i r l y weak sensitivity of theenergy content to resonance position was accompanied by clear
in-out asymmetry in electrical conductivity, consistent with astrong trapped particle population when heating on the low f i e l dside.
Surface barrier diode array data show that thesi owing-down time of this non-thermal trapped particle group isessentially classical [ 4 ] , while the particles would have toslow down much more rapidly i f they were to account for theobserved anomalous electron energy loss. Lowering the plasma
density raised the stored energy, anisotropy and the populationof the 13 keV t a i l .
At 60 GHz the higher densities and lower temperatures
made the ECRH plasmas much less collisionless, resulting inessentially Maxwellian isotropic distributions, while thedensity dependence of Tr or W. . became weakly positive up to
the cut-off and the sensitivity to resonance position became
more c r i t i c a l . At both frequencies the stored energy increase
rose with plasma current, saturating near q~2 with theIIAEA-CN-47/F-III-2

577

TABLE I. PARAMETERS OF OHMIC HEATING TARGETPLASMA AND OF THE RESULTANT ECRH PLASMA
Ip = 10 kA, B* = 1.07 T

Ohmic heating

ECRH

Te(0)

400 eV

1250 eVne(0)18

8 x 10 nT

/3p

0.7

2.7

14 kW

155 kW

2.3 ms

0.9 ms

POH/RF3

6 x 10 1 8 m- 3

RF-dominated 28 GHz case and at higher q with the 60 GHz case,where the stored energy of the target plasma became significant.
The effect on this current scaling of adding vacuum rotationaltransform using the i=3 stellarator windings was investigated at
28 GHz, with the conclusion that an improvement was realised
only when a significant relative contribution was made by thevacuum transform.
The parameters of the ohmic heating target plasma and of
the resultant ECRH plasma at 60 GHz are summarised in Table I
and typical waveforms are shown in Fig 1. The ohmic plasmas
feature sawteeth and Z f f < 2. RF probes indicate that about
85% of the incident RF power is absorbed.
The confinement
degradation seems to be stronger than the Merezhkin scaling [5]but similar to the L-mode.
With gettering, good vacuum conditions and constant gas
injection rates, ECRH in CLEO invariably produces a f a l l in line
averaged density, reaching a factor > 2 when the i n i t i a l density
is a l i t t l e above the nominal cut-off.
However, measurementswith a two-channel interferometer show that part of this f a l l isdue to a broadening of the density p r o f i l e . Ha measurements showthat the particle confinement decreases by a factor ~ 2.
In CLEO the stored energy (p + S../2), SXR flux, etc. ingeneral do not relax during the RF pulse (usually 20-50ms induration) even when the resonance is displaced off axis.
Fundamental heating has been directly compared with second
harmonic heating at the same toroidal field using 28 and 60 GHz.

ROBINSON et al.

578(V>l)lN3dbinO VWSVld(A) 39V±1OA dOOl

A1ISN3Q

IIAEA-CN-47/F-III-2

579

580

ROBINSON et al.

For an 11 kA plasma current at n ~ 3.6xlO 18 m"3 and B ~ 1.04 T
the central electron temperature rise was p r a c t i c a l l y the same:
450 eV (ohmic) * 1 keV for 28 GHz and 450 eV * 1.15 keV for 60
GHz. However, the range of parameters over which the heating i s
e f f e c t i v e is greater with second harmonic heating.
CURRENT DRIVE

At 28 GHz the wave was launched in the median plane atabout 80° to the magnetic field parallel or anti-parallel to I .
The directed launch was obtained by mixing TEQ1 and TE n modes
and checked by launching the gyrotron output through the entire
waveguide system into free space, with f a r - f i e l d antenna
patterns recorded on heat sensitive paper.
The principal
antenna lobe was directed with and against the plasma current,
but no discernible differences in the current or loop voltage
behaviour appeared, despite a theoretically predicted current of~ 10 kA (for the far-field launch angles).
60 GHz ECRCD experiments utilised a Vlasov antenna,
converting a TE02 mode to a single crescent-shaped launchpattern, angled at about 70° to the magnetic f i e l d . This
theoretical pattern was verified in detail using a real-time

30°

20°

10°

0°

10°

1

I

I

I

I

Infra-red image of antenna pattern(60GHz,180kW)

30mm

I
/—'
l_^— waveguide
FIG. 2. Far-field infra-red image of antenna pattern for the Vlasov antenna.

581

IIAEA-CN-47/F-ni-210Pr, = 185 kW
Ol p =10kA C.W

0-8

>

06

O
>

8 0A

02

096

100

1-04

108

1-12

FIG. 3. Loop voltage during ECRHfor clockwise and anticlockwise plasma current
as a Junction of toroidal field.

temperature-analysing infra-red camera to look at the result of
launching the gyrotron output onto large absorbent tiles, as
shown in Fig 2. Calorimetric measurements indicate that 185 kW
is radiated into the torus with this directivity, from one
200 kW gyrotron 32 metres, 3 bends and 2 convertors away.
Current drive experiments with this configuration werecarried out at n ~ 3.5 x 1 0 1 8 m" 3 (see Fig. 1) and withessentially Maxwell ian central temperatures of 1.3 keV at I =10 kA and 1.5 keV at I = 1 5 kA,as deduced from SXR spectra.
Reversing the servo-controlled plasma current or the antenna
direction produced a clear and consistent difference in loop
voltage. Figure 3 shows the difference for a toroidal field
scan with both current directions. The theoretical assumptionthat the ECRH current is essentially the same for each plasma
current direction allows its magnitude to be unfolded from the582

ROBINSON et al.
-|100

16 •

Theory

75
12

10

X

50

Experiment

8

25

100

102

1-04

106

108

1-10

FIG. 4. Comparison of experimental and theoretical values for deduced RF driven current as a functionof toroidal field, and percentage absorption as a function of B^, for Ip = JOkA and P^ = 185 kW.
For percentage absorption, the open and closed circles correspond to clockwise and counter-clockwise
plasma current, respectively.

loop voltage difference, with the result shown in Fig. 4. The
deduced RF current increases with plasma current from 2.5 kA at
5 kA to 5 kA at 15 kA.
A calibrated 'horn receiver' detecting the magnitude of
60 GHz radiation in the torus has been used to estimate theunabsorbed gyrotron power. The absorbed power is typically
80-90% and only a weak function of B , I and n , indicating theimportance of wall reflections for absorption, as theoretically
predicted. The wave driven current, however, varies morestrongly with B and ID>as shown for example in Fig. 4.
Theoretical calculations of the RF driven current in theseexperiments have been made using a ray tracing code which
includes a current drive model allowing for trapping effects andthe weakly relativistic resonance condition [6], The radiation
pattern shown in Fig. 2 was modelled by a bundle of 270 rays.

IIAEA-CN-47/F-III-2

583

Plasma profiles of the form n = n(l-r 2 /a 2 ) 2 and T = T(l_ r 2/ a 2)3 (n^^ a n d j ^ typically 0.8 x 1 0 " m-3 and 1.3 keV,
respectively) and a hot-core shift of up to 4 cm were assumed, in
accordance with the experimental data.
The calculations
including wall reflections predict greater than 80% absorption,
in reasonable agreement with the 'horn receiver1 measurements asshown in Fig. 4. With the resonance in the hot core the driven
current profile has a half-width of about 3 cm, with anefficiency of typically 60 A/kW.
This leads to predictedcurrents of approximately three times the observed values in the
hot core (Fig. 4). As the heating and driven current profiles
are somewhat broader than the high-q displaced ohmic profiles,
sawteeth suppression is to be expected. In general, this isobserved at 15 kA, but, at 10 kA with gas puffing, sawtooth
activity has been observed several msec after the RF is applied,with a period ~ tr and with individual sawteeth showing strong
saturation. SXR profile broadening is observed with ECRU at 5,Tne
10 and 15 kA. For these pi asmas, p / p - r r o v o n ~ °« 3 «calculations show that positioning the resonance away from themagnetic axis leads to broader driven current profiles but onlya modest fall in absorption.
The observed loop voltage response to co- and counterdirections is in satisfactory agreement with the predictions of
ID heat and current diffusion calculations using deposition
profiles calculated from the ray tracing code.
The interpretation of these results as current drive needs
to be treated with caution. Preferential loss of fast electrons
carrying toroidal current in one direction could give rise tosome of the observed voltage difference, as observed on previous
stellarator experiments on CLEO. With a symmetric antenna at 10
kA (but only weakly at 15 kA) the loop voltage response to the
ECRU varies with plasma current direction, indicating apreferential loss current in the opposite direction to the net
RF currents seen with the Vlasov antenna. The line averaged
density and SXR flux profiles are similar for each current
direction during the RF but not the H« emission. In addition,
the profiles are sensitive to whether gas puffing is applied tocompensate the apparent fall in line density.
The bootstrap current is in the same sense as the total
current, and for p > 2, v* ~ 0.1, and with the inferred profiles

584

ROBINSON et al.

it should be ~ 6 kA with a hollow distribution.
In theexperiment the bootstrap current must be much smaller, as theaverage decrease in loop voltage corresponds closely to the risein temperature if the resistivity is neoclassical.

MODE CONTROL
Experiments on CLEO using 60 GHz 2u near-axis ECRH havesucceeded in raising the disruptive limit for 9 kA discharges at
1.07 T by ~ 70%, attaining an I/N value of 1.0 x lO-l^ A-m (where
N = :ra 2 n). Adding vacuum rotational transform to perturb the
relative locations of the current density gradients and the q=2
surface has been similarly successful, although the density
limit remained disruptive and could only be improved at high q[7].
Four saddle coils have been added to CLEO, capable ofoperation at up to 6 kiloampere-turns per winding in four
configurations producing predominantly (m,n): 1,1; 2,1; 1,0(3,2) or 2,0 (2,2) (the principal resonant modes of the n=0
options being shown in brackets). Following the successful
stabilisation of ohmic heating sawteeth by 1,1 magnetic
perturbations in TOSCA [8], experiments have been conducted on
CLEO to investigate sawteeth stabilisation using the 1,1
perturbations in the presence of high power 2 W ECRH on axis at
28 GHz since this produces strong sawteeth at high central
temperatures.
Preliminary results, however, indicate onlyinterference of the sawtooth process and not fullstabilisation.
CONCLUSIONSECRH current drive experiments at 60 GHz have shown anasymmetric current component of up to * 5 kA for 185 kW injected
power. For 28 GHz, second harmonic heating with v* ~ 0.01 and a
non-Maxwellian electron distribution function, the energy
confinement time is better than the ohmic target plasma, whileat 60 GHz with v* ~ 0 . 1 , near-Maxwel1ian d i s t r i b u t i o n s are
obtained but with some confinement degradation. Low-m MHD mode
control has been demonstrated in a variety of ways,

IIAEA-CN-47/F-III-2

585

significantly raising the disruptive density limit and achievinga measure of sawteeth control, both by using localised ECRH andexternally imposed helical fields.
ACKNOWLEDGEMENT
The authors would like to thank the CLEO operating team fortheir valuable support in these studies.
REFERENCES[1]

EDLINGTON, T., et al., in Controlled Fusion and Plasma Heating (Proc. 13th Europ. Conf.
Schliersee, 1986), Part II, European Physical Society (1986) 219.
[2] ROBINSON, D.C., et al., in Plasma Physics and Controlled Nuclear Fusion Research 1984
(Proc. 10th Int. Conf. London, 1984), Vol. 1, IAEA, Vienna (1985) 205.
[3] SOLDNER, F., et al., in Controlled Fusion and Plasma Physics (Proc. 12th Europ. Conf.
Budapest, 1985), Part II, European Physical Society (1985) 244.
[4] ROBINSON, D.C., "Small scale turbulence and anomalous transport in magnetized plasmas",in Magnetic Reconnection (Proc. Workshop Corsica, 1986).
[5] MUKHOVATOV, V., JETP Lett. 33 (1981) 463.
[6] FERREIRA, A., O'BRIEN, M.R., START, D.F.H., Plasma Phys. Controll. Fusion 26 (1984)
1565.
[7] TODD, T.N., et al., in Controlled Fusion and Plasma Physics (Proc. 12th Europ. Conf.
Budapest, 1985), Suppl., European Physical Society (1985) 76.
[8] McGUIRE, K., ROBINSON, D.C., in Controlled Fusion and Plasma Physics (Proc. 9th Europ.
Conf. Oxford, 1979), Vol. 1, UKAEA, Culham Lab., Abingdon (1979) D2.7.

DISCUSSION
M.C. ZARNSTORFF: How much of your jSp is due to fast electrons? And
what is your anisotropy?
D.C. ROBINSON: According to diamagnetic loop and 0P 4- iJ2 measurements, there is no significant anisotropy in these 60 GHz experiments. The SiLi
detector measurements of the electron distribution function show only a slight distortion from Maxwellian at energies of a few times thermal (1.5 keV); thus we believe
there is no significant f5p contribution from the fast electrons.
M.C. ZARNSTORFF: After the ECH pulse, /3p + {12 dropped to far below
the initial Ohmic value. What is the reason for this?
D.C. ROBINSON: In this particular case, the charge came from the toroidal
field, which led to some pick-up in the diagnostic; thus the low /3p + ij2 value iserroneous is this case. With fi^ held constant, /3p + \I2 returns to the previous
Ohmic value after the ECRH pulse.

586

ROBINSON et al.

M.A. DUBOIS: On TFR, we have also suppressed sawteeth with ECRH, but
our explanation is very different. You say the temperature profile broadens; what
diagnostic do you use to measure this?
D.C. ROBINSON: In these experiments the sawtooth is suppressed by current
drive and heating. Theoretical calculations predict current and temperature broadening. Broadening of the temperature profile is deduced from soft X-ray measurementswith surface barrier diodes and, when combined with density profile measurements,
with SiLi detectors.

IAEA-CN-47/F-III-3

CONFINEMENT IN ELECTRON CYCLOTRONHEATING EXPERIMENTS ON DOUBLET III
R. PRATER, S. EJIMA, R.W. HARVEY,
A.J. LIEBER, K. MATSUDA, C.P. MOELLER,
and the DOUBLET HI PHYSICS GROUPGA Technologies Inc.,San Diego, California,
United States of America

Abstract
CONFINEMENT IN ELECTRON CYCLOTRON HEATING EXPERIMENTS ON DOUBLET III.
Electron cyclotron heating (ECH) experiments have been performed on the Doublet III tokamak
in the expanded boundary divertor configuration. Effective ECH is found even at densities twice thecutoff density, with no decrease in the global energy confinement time when the cutoff is exceeded.
Analysis of the behaviour of the sawteeth and of the soft x-ray emission indicates that for the overdense
discharges ECH takes place near the plasma edge. A simplified transport analysis indicates that the electron thermal transport coefficient decreases by a factor of 2 when the heating is near the edge, while
it increases by a factor of 2.5 when the heating is centrally localized. The transport analysis is confirmedby the behaviour of the propagation of the heat pulse when ECH is initiated. The edge heating results
in a large improvement in energy confinement over the inner two thirds of the plasma.

INTRODUCTION
Improvement of energy confinement in tokamaks remains a central
objective of the fusion program. Attainment of a mode of improved confinement (the H-mode) has been associated with changes in the plasmaedge in previous experiments using neutral injection in tokamaks [1]. Experiments using Electron Cyclotron Heating (ECH) on the Doublet III
tokamak have suggested means of improving confinement through direct
heating of the plasma edge.
In results presented here, Doublet III was run in the expandedboundary divertor configuration with elongation of 1.5-1.7. The working
gas was deuterium. The ECH experiments [2] were carried out using a
60 GHz source with power delivered to the plasma in the range 0.5 to 0.7
MW, for pulses of typical duration 250 msec. The frequency corresponds
to a gyromagnetic resonance at afieldof 2.14 T. The power was launchedfrom the vessel inner wall at the plasma midplane, using oblique launch(30° from the radial at the inner wall) of the extraordinary mode with
better than 90% mode purity [3]. The measured launching pattern ofthe elliptically polarized TEoi mode has a divergence of 12° FWHM in
the vertical plane and 16° in the horizontal plane.
587

588

PRATER et al.

Ray tracing calculations have been performed for this mode and geometry, using the TORAY code [4] for experimentally determined density
profiles. In these calculations, the ray trajectory is determined using thecold plasma dispersion relation, and the absorption is integrated along
the ray. It is found that the single-pass damping of a group of rays
representing the distribution pattern of the antenna is strong up to aline-averaged density of n^ = 5.6xlO19 m~3, at which density the absorption drops rapidly with increasing density, due to refraction of therays away from the plasma center and to the decrease of the absorptivity
with density. This critical density constitutes an effective cutoff density.
The maximum local density at which the dispersion relation permits
propagation of any polarization or angle is 8.8 XlO19 m~ 3 at 60 GHz.

EXPERIMENTAL RESULTS
In order to experimentally test the anticipated cutoff density, theplasma density was scanned over a wide range, 1 < fie < 11 XlO19 m~ 3 ,with ECH power in the range of 0.5-0.7 MW and with the gyroresonancenear the plasma axis. The central electron temperature determined from
Thomson scattering, Fig. la, shows the largest increase at low density
near 2XlO19 m~ 3 , where AT e « 2 keV. As the density increases, AT edecreases to about 0.3 keV near the critical density, but at densities
above the critical density ("overdense") the temperature increase remains constant at that value.
Similarly, the total plasma energy W, Fig. lb, increases by about
20 kJ for all densities above 2x 1019 m~3, with no change as the criticaldensity is exceeded. The plasma energy is determined from MHD equilibrium analysis,using signals from magnetic flux loops and probes. The
error in this analysis is estimated to be about 0.1 in f3p, or about 15%
in the ECH case. The values for W generated from measurements ofplasma diamagnetism agree with those from MHD analysis within 20%.
The energy confinement time T^, determined simply from the expression TE = W / ( P O H + PECH)> also shows no degradation as the critical density is exceeded. This behavior is shown in Fig. Id. In thiscalculation of TE the conservative assumption is made that all powerincident on the plasma is absorbed, and therefore the plotted T& value is
its lower bound. At the highest densities, the drop in loop voltage during the ECH phase generates a decrease in the ohmic input power whichis comparable to the ECH power. For these discharges the incremental
confinement time rinc = AW/AP is divergent.

589

IAEA-CN-47/F-III-3
120(b)

RESONANCEINACCESSIBLE

RESONANCEINACCESSIBLE
80

ECH •
•

•
OHMIC

40

OHMIC
0

2

4

6

8

10

12

0

2

4

6

8

10

12

8

10 12

(c)
OHMIC
>

2

UJ

•

•

r

ECH *'

68

9

rrf 3 )

610 12
fie (10

19

3

m" )

FIG. 1. (a) Central electron temperature from Thomson scattering, (b) total plasma energy from MHD
equilibrium analysis, (c) loop voltage on the plasma, and (d) gross energy confinement time, as a Junction of line averaged density. The hollow circles are taken from ohmic times and the filled circles arefrom ECH times. The toroidal field is 2.1 T and the plasma current is 500 kA.

The improvement in confinement time over the ohmic confinement
time for the overdense heating case may not be significant, since theohmic phase of the discharge is rather unusual. In particular, it is characterized by the presence of a volume of very high density (a "Marfe"[5]) at the inner wall of the plasma chamber. This high density region
can be clearly identified in the radiated power signals from the bolometer, which examines 21 chords in a plane of constant toroidal azimuth
from the outside midplane. The Marfe appears to have a crucial role inthe development of the ohmic discharge and in the attainment of densities nearly a factor two beyond the usual experimental density limit for
Doublet III of ne(1019 m~3) < 27 x B T /Rq (in MKS units), but it also

590

PRATER et al.

seems to cause a reduction in confinement similar in magnitude to thatof a transition from divertor to limiter configuration [6]. The D a radiation from the top of the plasma is also similar in magnitude tothat found near a limiter in normal limiter discharges. When the ECH
is applied, the Marfe moves rapidly downwards into the divertor region
where it dissipates [7], and within 6 msec from the start of the ECH thedischarge has evolved into a normal well-behaved divertor configuration,
except that it has unusually high density.
An immediate question arises regarding the overdense heating case:does the ECH power get absorbed near the plasma center where the gyroresonance lies, but where the dispersion relation forbids propagation,or is the power absorbed at some other location where the resonancecondition is met but the density is lower?
One way to address this question is to examine the behavior of thesawteeth when ECH is applied. At densities low enough that refraction
is not important, the ECH power can be concentrated within a few cmof the magnetic resonance. When the gyroresonance is located near theplasma center the ECH power density can be much larger than the ohmic
power density, with the result that the power applied to the volume
within the q=l surface increases and that the gradient of the electron
temperature evaluated at the q=l surface tends to increase. Both ofthese effects may be expected to decrease the sawtooth period r st [8].
Similarly, placing the ECH power outside the q= 1 surface by shifting thelocation of the resonance reduces the power input within that surface

20
(a)

50

3£n e <4x10 1 9 m~ 3

B T - 2.1 T

_<b>

17

£ 30 -

o 14

/ECH

1

/
20
10

0

i

19

2.1

B T (T)

2.3

2-5

i

i

OHMIC

/
>

- •—"fie

Q.
I

i

i

i

i

i

i

10

1
1S m
«-a
"e ( 1 °O9"

FIG. 2. Sawtooth period as a function of (a) toroidal field, for IP = 350 kA and 3.0 < ne
< 4.0 x 10'9 m ~3, and {b) density, for IP = 500 kA and BT = 2.1 T. ECH power is 0.5-0.7 MW.

591

IAEA-CN-47/F-III-3

5
m

780

820

FIG. 3. SXR emission temporal response along three chords. The ECH is initiated at 800 msec.
The toroidal field is 2.2 T, the plasma current is 350 kA, and the density is ne = 5.5 x 1019 m~3.

since the loop voltage drops when the ECH is added. It also tends
to flatten the electron temperature profile near the q=l surface. Thisshould result in an increase in the sawtooth period. These effects areobserved in ECH experiments in moderate density plasmas when theresonance is scanned across the plasma, Fig. 2a. Placing the resonancewithin the q=l surface results in a decrease in r st , and placing theresonance outside the q=l surface results in an increase in r st .
When the resonance is maintained near the plasma center and thedensity is scanned, r6t displays the expected decrease from the ohmic
phase to the ECH phase consistent with central heating when the density
lies below the critical density, as shown in Fig. 2b. But above the criticaldensity the sawtooth period exhibits an increase of a factor five when
ECH is added. This suggests that when the critical density is exceeded,
the radius of power deposition shifts from the plasma center to muchfurther out on the temperature profile.
Evidence of a more direct nature comes from the soft x-ray (SXR)
emission behavior of the plasma when ECH is initiated. Propagationof the ECH-generated heat pulse through the plasma can be followedby plotting the delay in the first appearance of increase in the SXR
emission on the various chords. This method has also been used in ECH
experiments on T-10 [9].

592

PRATER et al.
(a)

UNDERDENSE

to- 3.5x10

19

m~

(b)

20 r

3

OVERDENSE

ne - 1.0X1020 n1-3
p - 500 KA

I
I!

I
MODEL
ECH POWER
INPUT

10 -

I
I

6 -

x

OH^-v.

-RESONANCE
}

>-*

0.2

0.4

0.6

f

1.0

0.2

0.4

0.6

r/a

0.8

10

0.0

I
1

1

I

i

iI
rn

0.8

j

1.0

r/a

FIG. 4. SXR delay time Td (defined in Fig. 2) plotted against relative minor radius: (a) underdense
discharge (ne = 3.5 x W'9 m~3) from Fig. 2, and (b) overdense discharge (ne = 1.0 X 1020 m~3,
ne(0) = 1.6 x 1020 m~3), with BT = 2.1 T and IP = 500 kA. The dashed lines show the results froma transport model for various values of xe- Also shown is the assumed location (shaded area) of the
input ECH power used in the model.

An example of SXR data from a discharge with density well below
the critical density (fie ss 3.0 X1019 m~3) and with centrally located resonance is shown in Fig. 3. The 32 SXR chords, which lie in a plane ofconstant azimuth, are parameterized by the relative minor radius r/aof the flux surfaces to which they are tangent. The increase in SXRemission is prompt from chords with r/a < 0.15 but delayed for chords
away from center, and the delay increases with distance. Outside ofabout r/a w 0.7 the signal-to-noise ratio becomes too small for this typeof analysis.
The direction and magnitude of the propagation speed of the heatpulse can be determined from a plot of the SXR delay Td as a function ofminor radius. Figure 4a shows ra for the same low density discharge as
Fig. 3, and Fig. 4b shows data from an overdense discharge. It is clear
from Fig. 4 that in the underdense case the heat propagates radiallyoutward from the center, while for the overdense case the heat propagates in an inward direction, at least for r/a < 0.7. It may be concluded

IAEA-CN-47/F-III-3

593

from this that for the underdense case the ECH power is being deposited
within a radius of 0.15 while for the overdense case the ECH power isdeposited somewhere outside of r/a=0.7.

Modeling
The ONETWO transport code [10] has been used to model transport in ECH discharges in Doublet III. The inputs to the code include
the directly measured quantities, such as the toroidal field, the plasma
current, the plasma geometry from the MHD equilibrium analysis, theelectron temperature profile, the density profile, and the radiated power
profile. The electron temperature profile is measured using second harmonic electron cyclotron emission, and the density profile is determinedby unfolding data from a three-channel interferometer, using the actual
flux surface geometry to determine chordal path lengths. The radiatedpower profile is measured by the 21-channel bolometer. For these sawtoothing discharges, the central q(0) is input as 1. The impurity density
is assumed to be oxygen, and it is adjusted in an iterative manner toobtain q(0)=l given the loop voltage V/. Further inputs are the ECH
power profile as calculated by TORAY and the assumption that the ion
thermal conductivity is neoclassical [11]. The time-dependent electron
and ion heat equations and the current density diffusion equation using
Spitzer conductivity are solved, keeping the density profiles constant.
Code outputs are Xe(r)» Ti(r), V*, plasma kinetic energy W, and global
energy confinement time TE, which are compared with the experimental
data.
Analysis of discharge #44055, from which data in Fig. 2 was taken,
showed that a good fit to T e (r), V^ and W was obtained with Xe
nearly constant across the profile, for the ECH power deposited within
r/a=0.15. The value of Xt so determined during the ECH steady statephase is 0.8 m 2 /sec, about 2.5 times its value during ohmic heating inthe same discharge, which accounts for the observed decrease in grossenergy confinement time from 100 msec before ECH to 54 msec during
ECH. The change in the global confinement time when ECH is added
represents the transition from the ohmic scaling to the scaling for discharges dominated by auxiliary heating, which is obtained whether theauxiliary heating is ECH or neutral beam injection [12].
An alternative way to determine Xo is to follow the propagation of
the heat pulse at the onset of ECH. The ONETWO code calculates the594

PRATER et al.

SXR emission along the chords corresponding to the SXR diode arraygeometry, and the delay between the application of ECH and the startof the rise of the calculated emission can be compared with experiment.
This delay rj is plotted for two values of Xe as the dashed lines in Fig. 4a.
It is clear from the figure that rj is sensitive to the value of Xe used in
the calculation, and that the value for Xe of 2.5 times that of the ohmic
heating phase is a good fit to the data. This implies that the magnitudeof the thermal conductivity as determined by heat pulse propagation isclose to that determined by analysis of the final steady state, and thatthe increase in Xe from ohmic to ECH occurs throughout the plasma on
the time-scale of the heat pulse propagation.
The overdense ECH discharges have also been analyzed using theONETWO code, although T e profiles from electron cyclotron emission
are not available, due to cutoff of the 2wce radiation. In order to proceedwith analysis, the thermal conductivity is assumed to be spatially constant and the T e profile is calculated (assuming a boundary temperatureTeb). Following the conclusion from Fig. 4b that the heat propagates inward, the ECH power is modeled as being deposited near r/a=0.8. The
requirement that the central temperature and the total plasma energy
match the experiment places a constraint on the Te(r) shape, with theprincipal uncertainty coming from the ion transport (mainly through theassumed value of Xi) and from the uncertainty in Teb- If the calculated
electron temperature profile is parameterized for convenience of discussion as T e (r) = (Te0 - T eb )(l - r 2 /a 2 ) Q T + T eb , and assuming T eb =0.2
keV, this constraint places the limits 2.0 < ax < 2.5 on the profile parameter, which means that the T e profile remains rather peaked during
the ECH. Lower T eb produces fits with smaller ax, but even T eb = 0
implies that ax > 1.5, in order for the plasma energy to match experiment. The conclusion that the electron temperature profile remains
rather peaked is also consistent with the determination of the internal
inductance, £i « 1.5, which doesn't change much during the ECH.
The value of the average Xe determined from this simplified treatment is about 1 m2/sec during ohmic heating, and during ECH it decreases a factor two to 0.5 m2/sec. Although Fig. 4b clearly shows propagation of the heat pulse toward the plasma center, comparison with themodel is considerably complicated by the process of the dissipation of
the Marfe, which takes about 6 msec. In addition, for the model the determination of Td for locations near the plasma center is ambiguous, due
to smearing out of the leading edge of the heat pulse. Nevertheless, it
may be concluded from Fig. 4b that the thermal diffusivity determinedfrom heat propagation is not larger than that during the ohmic phase.

IAEA-CN-47/F-III-3

595

DISCUSSION AND SUMMARY
In the overdense heating case, the central energy confinement time
improves dramatically. The behavior of both the sawtooth period andthe SXR emission imply that the ECH power is not being depositednear the resonance at the plasma center, in violation of the dispersion
relation, but rather that the plasma edge is being heated. The drop in
the loop voltage from 2 V prior to ECH to 0.8 V during ECH, Fig. lc,
indicates that the ohmic power decreases at the plasma center; however,the classical skin time is longer than the ECH pulse length, and theplasma modeling for this case indicates that a central voltage drop to
1.3 V takes place at the plasma center during ECH. But the measured
central density is constant and the central electron temperature increases
from about 0.5 keV to 0.8 keV. These observations imply that the centralenergy confinement time increases from 70 msec before ECH to 190 msec
during ECH.
A similar conclusion may be reached concerning the plasma within
r/a=0.7: the confinement time for this volume of plasma increases by afactor 2 to 2.5, with the caveat that the electron temperature profile isonly inferred from the measurement of central temperature and of total
plasma energy. The conclusion that confinement strongly improves is
not very sensitive to the details of the electron temperature profile, but
depends mainly on the absence of ECH power within r/a=0.7 [7].
Shifting the location of the resonance by changing Bx for underdense plasma produces a similar effect. Over the range 1.8 < Bx < 2.5 T,
for which the resonance moves from r/a = —0.5 to +0.5, the global energy confinement is constant [2]. (This result has also been seen in T-10
ECH experiments [13].) This implies improvement in the central energy
confinement as the heating is moved off axis. Neutral beam heating
experiments on TFTR [14] and ASDEX [15] have also shown similar
results when the depth of the neutral beam penetration is reduced byoperating with high plasma density or reduced beam energy. In thosecases, a very long energy confinement time near the axis was observed.
The high density heating case reported here with power deposition at
r/a > 0.7 may correspond to an extreme case of these results.
The physical bases behind these observations are not well understood. It is possible that a heat diffusivity which is dependent on thegradient in T e may account for this behavior. A heat diffusivity which
is proportional to (VT e ) 2 has been derived [16], and this strong dependence on the profile could result in the observed independence of theincrease in stored energy on the profile of the heating/ At the sametime, it is observed that the transition to H-mode is closely associated596

PRATER et al.

with changes in the plasma edge. It may be that direct edge heating
is an important ingredient in the mix of conditions that produces the
H-mode. Experiments aimed at edge heating in DIII-D at low density
using second harmonic ECH are planned to try to clarify this.

ACKNOWLEDGMENT
This work was supported jointly by the U.S. Department of Energy
and by the Japanese Atomic Energy Research Institute under Contract
Nos. DE-AT03-84ER51044 and DE-AT03-80SF11512. We would liketo thank T. Ohkawa and D. Overskei for their encouragement during
this work. ECH engineering support from D. Remsen and D. Bell andoperational support from R. Callis and his group are also gratefully
acknowledged.
REFERENCES[1] WAGNER, F., et al., Phys. Rev. Letters 49 (1983) 1408.
[2] PRATER, R., BURRELL, K.H., EJIMA, S., JAHNS, G.L., LIN, S.H.,
MOELLER, C.P., Summary of Electron Cyclotron Heating Physics Results from the Doublet III Tokamak, GA Technologies Report GAA17939 (1985).
[3] MOELLER, C.P., PRATER, R., LIN, S.H., The 60 GHz Transmission
System for the Doublet HI ECH Experiment, GA Technologies Report
GA-A16979 (1983).
[4] KRITZ, A.H., HSUAN, H., GOLDFINGER, R.C., BATCHELOR,
D.B., Proc. 3rd Joint Varenna-Grenoble International Symposium on
Heating in Toroidal Plasmas, Grenoble (Euratom Report EUR7979 EN),v. H (1982) 707.
[5] LIPSHULTZ, B., LABOMBARD, B., MARMAR, E.S., PICKRELL,
M.M., TERRY, J.L., et al., Nud. Fusion 24 (1984) 977.
[6] OHYABU, N., BURRELL, K.H., DeBOO, J., EJIMA, S., GROEBNER, R., et al., Nucl. Fusion 25 (1985) 49.
[7] EJLMA, S., PRATER, R., Interpretation of Electron Cyclotron Heating Results in Overdense Plasma in Doublet III, GA Technologies Report
GA-A18535 (1986), submitted to Nuclear Fusion.
[8] McGUIRE, K., ROBINSON, D.C., Nucl. Fusion 19 (1979) 505.
[9] ALIKAEV, V.V., RAZUMOVA, K.A., in Course and Workshop on
Applications of RF Waves to Tokamak Plasmas (International School of
Plasma Physics, Italy EUR 10333 EN), v. I (1985) 377.
[10] PFEIFFER, W., et al., GA Technologies Report GA-A16178 (1980).

IAEA-CN-47/F-III-3

597[11] CHANG, C.S., HINTON, F.L., Phys. Fluids 25 (1982) 1493.
[12] BURRELL, K.H., PRATER, R., EJIMA, S., et al., in Plasma
Physics and Controlled Nuclear Fusion Research (Proc. 10th Int. Conf.
London, 1984), IAEA, Vienna, v. I (1985) 131.
[13] ALIKAEV, V.V., ARSENTIEV, Yu.L, BAGDASAROV, A.A., BEREZOVSKIJ, E.L., BORSHEGOVSKIJ, A.A., et al., in Plasma Physics
and Controlled Nuclear Fusion Research (Proc. 10th Int. Conf. London, 1984), IAEA, Vienna, v. I (1985) 419.
[14] MURAKAMI, M., ARUNASALEM, V., BELL, J.D., BELL, M.G.,
BITTER, M., et al., Plasma Physics and Controlled Fusion 28 (1986)17.
[15] WAGNER, F., GRUBER, O., LACKNER, K., MURMANN, H.D.,
SPETH, E., et al., Phys. Rev. Letters 56 (1986) 2187.
[16] GUZDAR, P.N., LIU, C.S., DONG, J.Q., LEE, Y.C., Bull. Am.
Phys. Soc. 31 (1986) 1100.
DISCUSSION
R.J. GOLDSTON: There is an inductive term in the power balance of the form
V4 Ip(d/dt)Lj. L, can be measured from MHD fitting, or from a combination of
MHD fitting and diamagnetism in a high density isotropic plasma. Could you comment on the size of this inductive term?
R. PRATER: This term is quite small since very little change is found in theinternal inductance during ECH, even with off-axis heating.

IAEA-CN-47/F-III-4

ELECTRON CYCLOTRON HEATING ON TFR
H.P.L. DE ESCH, J.A. HOEKZEMA, W.J. SCHRADER,
E. WESTERHOF, W. KOOIJMAN, R.W. POLMAN, T.J. SCHEP
Association Euratom-FOM,
FOM Instituut voor Plasmafysica,
Rijnhuizen, Nieuwegein, Netherlands
EQUIPE TFRAssociation Euratom-CEA sur la fusion,
CEA, Centre d'etudes nucle'aires de Cadarache,Saint-Paul-lez-Durance, France
Presented by M.A. Dubois

Abstract
ELECTRON CYCLOTRON HEATING ON TFR.
Results are presented of the joint Euratom-CEA, Euratom-FOM electron cyclotron heating
experiments on TFR (March 1985 to June 1986). Up to 500 kW of 60 GHz power was launched into
TFR; with central heating, a central electron temperature up to 5 keV was obtained.

1.

INTRODUCTION

The ECRH facility on TFR comprised three 60 GHz gyrotrons, each delivering over 200 kW for 100 ms in the TEQ2-mode.
In the three transmission lines conversion to TE-| -| was realizedwith an efficiency of approximately 70$ (0-mode into TFR overgyrotron output) [1]. The open oversized waveguide antennae
were located in the same poloidal cross-section, in the horizontal plane and at + and -15°, and were aimed at the toroidal
axis [1]. The full 3 dB beam width on axis was about 3 cm.
Facing two of the antennae was a concave corrugated reflector(10° oblique), converting transmitted power into the extraordinary mode at 20° off the meridian plane. The third, upper,
antenna had a 16° angle with the meridian plane.
Experiments began in March 1985 with one gyrotron at
100 kW, 30 ms with a smooth, concave, roof-top-shaped reflector(angles ± 10°) on the inboard torus wall [2]. Full power(500 kW into TFR for 100 ms) experiments were done between
September 1985 and June 1986 [3,4,5]. TFR was operated withcarbon limiters. The parameters in the ohmic regime varied over
the range:
B(0) = 1.8 to 2.7 T; n e (0) = (1 to 5)*1019 m~3 ; I p = no to 160 kA

599

600

DE ESCH et al.

Under iron-free plasma conditions, two ohmic regimes
were obtained: At low plasma current ('L.C : I p < 105 kA) no
MHD activity was observed and T E > 8 ms, while at high current('H.C: I p > 105 kA) strong saturated MHD activity coincidedwith ig 1 5 ms. When iron was present (due to a calorimeter inthe shadow of the limiter), the iron-dominated regime O L D . 1 )showed no such clearcut differences with I p , and MHD, whenpresent, tended to be erratic or explosive.

2.

HEATING

When the resonance is located near the magnetic axis,
for average ohmic plasma parameters 60-70? of the power should
theoretically be absorbed in the first pass. Because of T e
increase during ECRH, this figure should rise over 90?. Measurements of the transmitted power, using small holes in the
mirror, indicate that at least the expected absorption takes
place. For the -rE-calculation, we have assumed that the total
injected power is absorbed. This is probably slightly pessimistic since over 10? of the power is injected in the X-mode.
When the resonance is set off-centre, a large fraction of thepower is transmitted through the plasma. Especially at low
densities, the reflected, X-mode power is mostly absorbed bysuprathermal electrons, for which confinement can be completely
different from that of the bulk. Experimentally it is foundthat the increase of plasma energy (not counting a possible
suprathermal population) has a maximum around central heating
and falls abruptly when the resonance is displaced off-centre
by more than a/3. This is illustrated in Fig. 1, which is valid
irrespective of the regime (although in the I.D. regime theoff-centre displacement was limited to +_ a/3 due to disruptions[4]).
For central heating, the degradation of confinement with
ECRH power in the different regimes is shown in Fig. 2 forintermediate density. It is clear that in the L.C. regime theconfinement is substantially degraded even with a small amountof additional heating; with ECRH, the difference between theregimes disappears. At lower density, the degradation of energy
confinement is much less (apart from the disappearance of the
L.C. regime): confinement may even improve, while at higher
densities the degradation is worse. This is due for a largepart to density behaviour: at high density, a strong pump-out
is observed in the centre, while at low density an increase ofdensity is seen. Combining discharges at different densities in
the ohmic phase to obtain the Tg-value at constant density for
different power levels, we find only a limited reduction of Tg(Fig. 2 ) .

601

IAEA-CN-47/F-III-4

800

BOO -

400

°-

200 -

0.0

-0.2

R

0.2

ECRH-Ro

FIG. 1. Plasma kinetic energy (Te from Thomson scattering) as a Junction of resonance layer position/ = 90 kA, nJO) - 2 x 1019 m~3, PECRH = 250 kW, Pa = 110 kW (without ECRH).

10

i

i

i

i

i

4

-

6 -

= \

]'r—

—B———.—Q

^ ^

—

2 -

O
i

0

1 2

i

i

i

1

3

4

5

6

7
'-i

*"

P

ERCH

FIG. 2. Energy confinement time as a function of centrally deposited ECRH power in three different
regimes, at ne - (1-1.5) x 1019 m~3:
• L.C., Ip = lOOkA
v H.C., lp = 115 kA
o ID.
• ID. at constant central density: ne(0) = 2.8 X 1019 m~3.

602

DE ESCH et al.

Transport simulations using the MAKOKOT code were
presented [4,5] with the conclusions that with ECRH, X e ~ s c a l l n 8
is proportional to rTe'2/(n.q ) and that the pump-out can be
explained by a combination of the rTg'Wn.q.) dependence of the
diffusion coefficient, the decrease of the neoclassical pinch
velocity, and enhanced recycling. Besides, the energy confinement time dependence with injected power is well described bythis scaling. Simulations with the ICARUS code have also been
performed, and a very good agreement is obtained both at highand low density in the I.D. regime forsee Fig. 3- This Xe~P r °fil e i-s ver>y close to twice the value
proposed by TANG [6].
At low density where diffraction of the ECRH beam in theplasma is weak, a very high central temperature (Te up to 5 keV
as measured by Thomson scattering and ECE) is obtained in anarrow central spike, too narrow for the ECE spatial resolution. Using discharges with the resonance on the magnetic axis
but at slightly different horizontal positions, a fine central
Te-profile was obtained from the central channel of Thomson
scattering: Fig. 1.
The width of the central Te-spike (3 cm) is in agreementwith the antenna pattern. The height of the spike indicatesthat confinement inside q=1 is degraded but, despite the verylarge VT e , not dramatically. The inequality R'VTe S 13 keV isfulfilled outside q=1 , as proposed in [7], but much larger
gradients can be obtained inside q=1. This is a strong incentive to stabilize sawteeth on larger machines where the energy
content inside q=1 is considerable.
In our experiment, although gradients inside q=1 canreach very high values, sawteeth are not very conspicuous: evenwhen T e (0) is larger than 3 keV, the sawtooth drop 6T e (0) is
less than 700 eV: the temperature is not flattened completely
inside q=1. When heating is applied outside q=1, sawteeth are
not affected. When heating is applied in the centre, sawteeth
are also unaffected, except at high density, in which case theperiod decreases. More interesting is sawtooth behaviour when
the ECRH layer is at the radius of the surface q=1: in thoseconditions, an increase of the sawtooth period is observed,
corresponding to a larger delay between m=1 growth and internal
disruption: the crash is triggered for a larger island size
than in ohmic regimes. At high density, sawteeth can disappear,
replaced by a saturated m=1 mode. This behaviour can be attributed to a local modification of the T e (r)- (and hence j(r)-)
profile. It is a direct effect on the sawtooth trigger and nota mere modification of the regeneration slope as in otherexperiments.

603

IAEA-CN-47/F-III-4
2.0

0.0
0.00

0.05

0.10

0.15
>•

r

0.20
(m)

FIG. 3. Electron temperature profiles in the Ohmic regime and witkfitll ECRH power at high density(ne = 2.2 x lO19 m~3) in the I.D. regime and resonance located at the centre:—ECE
Thomson scatteringprofiles from transport code simulation (ICARUS).

0.00

FIG. 4. Thomson scattering Te profile (for a low-density discharge (ne = 0.8 x 1019 m 3) at fullcentral ECRH (o) and with ohmic heating only (+).

604

DE ESCH et al.

An estimate of the ECRH deposition profile can beobtained from a measurement of the immediate change of thelocal T e after ECRH switch-on. The soft X-ray emission profile
was used to determine this profile and confirms the localization, in case of central heating, mainly within a radius of
2 cm. Analysis of heat pulse propagation experiments is not yet
completed and will be published later on.
3-

IMPURITY BEHAVIOUR

With central heating, no deleterious effect in impurity
production is observed during ECRH. A slight enhancement ofmetal production is suspected from preliminary numericalsimulations, but lower than 50$. Highly ionized ion lines (e.g.
Fe XXIII) appear during ECRH, indicating increased ionization
degree as a consequence of the strong electron heating. Whenthe ECRH layer is far off-centre, B<j> - 1 .8 T or B<t> - 2.4 T(2.14 T = central heating) strong increase of line radiances isobserved implying increased central ion densities by a factorof - 2 for light impurities and - 2 to 3 for metal impurities,
probably as a result of transmitted power.
4.

STABILIZATIONIn the H.C. regime, where in the ohmic regime saturated
m=2 modes are present, application of microwave power has beenshown able to suppress the q=2 tearing mode in two different
situations:- when the resonance is applied on the magnetic axis: energy
increase leads to an outward shift and, contact with theinner llmiter being reduced, the q=2 mode disappears as q aincreases. In some cases where low-power ECRH was applied in
comb-like pulses, stabilization was obtained without a marked
outward shift, but coincided with a better horizontal position regulation induced by ECRH: here also, interaction withthe limiter was reduced.
- when the power is applied just outside the q=2 surface [2,3].
the mode is also suppressed. The absorbed power is very small
and does not directly explain an outward shift. Two explanations are then possible: 1) local temperature gradientmodification leads to a j(r) change and, hence, to tearingmode stabilization. The subsequent confinement improvement
allows ohmic heating to increase the particle and energy
content, leading to outward shift; 2) the non-absorbed power
falls on the vacuum vessel, causing enhanced degassing and,
consequently, a density and energy increase, which leads toan outward shift and subsequent stabilization. A carefulanalysis shows that m=2 activity starts to decrease as soon

IAEA-CN-47/F-III-4

605

as q a rises above a critical value (- 3.3) and rises again
when q a drops below a sometimes different value (3.1 to 3.3).
If the current is too high, the change of q a induced by ECRH
is too small to cross this value, and MHD activity persists.
So, although stabilization has been routinely achieved with
ECRH, and even with a feedback loop C3], we are led toconclude that at least when the mode-converting mirror was inuse, the first mechanism is occulted by the second, which isof less physical interest. A better prospect will be openwith larger machines at higher densities and temperatures.

5.

SUPRATHERMAL GENERATION AND PELLET ABLATION

Creation of suprathermal electrons has been demonstrated
at low densities. As previously mentioned, they outlast the
ECRH pulse [2], showing the good quality of confinement inside
q=1. In low-density regimes, even a short ECRH pulse cantrigger the discharge to a suprathermal 2 w c e emission regime.
For n e (0) below 1.5x1O 1 9 m " 3 , the ECE is completely dominatedby suprathermals; between 1.5 and 2.5x10 l 9 m~3 ECE profiles
show an asymmetry which is most likely due to suprathermal
trapped electrons. Above, the ECE and Thomson-scattering
T e -profiles are similar.
A natural step was to try current-drive experiments with
ECRH. The result [3] is negative: all the effects observed
between 2.14 and 2.40 T are explainable by the resistivity
decrease due to bulk heating. At higher fields, no effect isobserved. As has been explained previously [5] this negative
result is likely to be due to a depletion effect of the electron distribution function when the resonance is applied far
from the bulk of the maxwellian. This depletion effect atintermediate energies, seen on the SiLi diagnostic and confirmed
in numerical simulations, is accompanied by population
increase at higher energies, mainly in the perpendicular
direction, which is responsible for the suprathermal 2u) c eemission. A direct evidence for this very energetic tail is the
behaviour of deuterium pellets injected into ECRH-heatedplasmas: they are sublimated differently than in normal ablation cases, and the radius at which they are sublimated is
directly linked to the ECRH layer.

ACKNOWLEDGEMENTThe Netherlands' contribution to this work was performed
under the Euratom-FOM association agreement with financial
support from ZWO and Euratom.

606

DE ESCH et al.
REFERENCES

[1]
[2]
[3]
[4]
[5]
[6]
[7]

VAN DER GEER, C.A.J., et al., and CRENN, J.P., et al., in Fusion Technology (Proc. 14th
Symp. Avignon, 1986), Pergamon Press, Oxford (1987).
FOM-ECRH TEAM and TFR GROUP, in Controlled Fusion and Plasma Physics (Proc. 12th
Europ. Conf. Budapest, 1985), Vol. 9F-II, European Physical Society (1985) 60.
TFR-GROUP, HOEKZEMA, J.A., et al., in Electron Cyclotron Emission and Electron Cyclotron Resonance Heating (Proc. 5th Int. Workshop San Diego, CA, 1985) 154.
TFR-GROUP and FOM-ECRH-TEAM, in Controlled Fusion and Plasma Heating (Proc. 13th
Europ. Conf. Schliersee, 1986), Vol. 2, European Physical Society (1986) 207.
JACQUET, L, PAIN, M., ibid., p. 215.
TANG, W.M., Microinstabilities Based Confinement Model for Enhanced Transport, Rep.
PPPL-2311, Princeton Plasma Physics Laboratory (1986).
TFR GROUP, Plasma Phys. Controll. Fusion 28 (1986) 85.

DISCUSSION
A. GIBSON: Is there any threshold power for the sawtooth stabilization which
you describe, and are you able to compare the power required for stabilization withany theoretical expectation?
M.A. DUBOIS: A strong effect was observed even when one gyrotron only wasused with ECRH on q = 1 (power ~ 150 kW). We did not try to use a lower power
level. I do not expect any threshold, but rather a linear relationship at low power,
namely, the increase in the sawtooth period is approximately equal to the ECRH
power.
A.C. RIVIERE: In connection with the observation of a narrow high peak in
Te for central resonance, how far did you have to move the resonance off axis sothat this steep narrow peak in Te disappeared? Further, how important were thesawtooth oscillations for these measurements?
M.A. DUBOIS: The peak was maximum and narrowest for central heating, but
a really visible peak existed as long as ECRH was applied inside q = 1.
With regard to your second question, the maximum amplitude we observed wasless than 600-700 eV, and usually it was much less. A considerable peak subsisted
after the sawtooth crash, showing that the flattening of the temperature profile wasfar from being complete.

IAEA-CN-47/F-III-5

INVESTIGATION OF PLASMA FORMATIONREGIMES FOR ICRF HEATING EXPERIMENTS
IN THE TUMAN-3 TOKAMAKWITH MAGNETIC COMPRESSION
L.G. ASKINASI, N.E. BOGDANOVA, V.E. GOLANT,
S.G. GONCHAROV, M.P. GRYAZNEVICH,
V.V. D'YACHENKO, A. IZVOZCHIKOV, E.R. ITS,
S.V. KRIKUNOV, S.V. LEBEDEV, B.M. LIPIN, I.P. PAVLOV,
G.T. RAZDOBARIN, V.V. ROZHDESTVENSKIJ,
N.V. SAKHAROV, M.A. KHALILOV, F.V. CHERNYSHEV,
K.G. SHAKHOVETS, O.N. SHCHERBIN1N
A.F. Ioffe Physico-Technical Institute,Leningrad
S.E. BENDER, Yu.A. KOSTSOV, R.N. LITUNOVSKIJ,
O.A. MINYAEV, E.N. RUMYANTSEV, P.P. TEPLOV
D.V. Efremov Scientific Research Instituteof Electrophysical Apparatus,
Leningrad
Union of Soviet Socialist Republics
Presented by O.S. Pavlichenko

Abstract
INVESTIGATION OF PLASMA FORMATION REGIMES FOR ICRF HEATING EXPERIMENTS
IN THE TUMAN-3 TOKAMAK WITH MAGNETIC COMPRESSION.
The effect of fast current increase on the plasma parameters is studied. A remarkable increasein the peripheral density and a growth of MHD activity are observed. During the current rise period
for a low magnetic field, Thomson scattering measurements yield hollow electron temperature distributions, which are rapidly transformed into a peaked distribution just after the current maximum. The
method of plasma formation studied leads to stable discharges with a low safety factor, q(a) = 2.
Experimental data on ICR heating with P < 0.5 MW in a plasma formed by a fast current rise arereported. A simultaneous increase in the ion temperature by a factor of 1.4 and in the density by a factorof 1.7 are recorded during the RF pulse.

607

608

1.

ASKINASI et al.

INTRODUCTION

The first experiments on fast current increase after magnetic compression werecarried out in the TUMAN-3 tokamak in 1983 [1]. The experiments were mainly
devoted to an evaluation of the time needed for the current to penetrate into theplasma core, but the conclusion that anomalously quick skin current dissipation takes
place could not be considered to be very reliable without any detailed measurementsof the electron temperature. In contrast to our data, later TFTR [2] and JET [3]results have shown that the rate of current density redistribution corresponds to theclassical Spitzer resistivity. The experiments [1] have also demonstrated the influence
fast current input has on the main plasma characteristics. For example, the ion
temperature was 1.5 times higher than the temperature measured in a quasi-stationary
Ohmic discharge with the same values of Ip, BT, and n [4].
The new set of measurements described in this paper aimed at studying themechanism of additional current penetration and the time evolution of the electron
temperature profiles during current rise. A plasma column that was separated fromthe wall was formed by a rapid shift along the major radius (R-compression) [5],whereas, in previous experiments, minor radius compression had been used. The
plasma obtained in this way is supposed to be used as an initial plasma in ICRF
heating experiments.
2.

EXPERIMENT

The main diagnostics used for recording the plasma parameters are as follows:a microwave interferometer (X = 2 mm) 'looking' through the plasma along ten
vertical chords, a Thomson scattering apparatus including ruby laser and polychromator with four space points, five energy channel neutral particle analysers
capable of recording charge exchange fluxes within +16 cm from the equatorial
plane. In addition, measurements of soft X-rays, vacuum spectral line intensities,
radiation losses and several electromagnetic parameters were performed.
The time behaviour of the plasma parameters in a discharge with fast current
increase is shown in Fig. 1. At the initial stage, the plasma is formed near the outerchamber wall. According to the magnetic flux measurements, the discharge axis is
at R = 61cm (Fig. l(b)). The low field side limiter is placed at the radius
R = 75 cm, the high field side limiter at the radius R = 31 cm. The plasma current
Ip = 43 kA (Fig. l(a)), the average density n = 0.6 X 10l3 c u r 3 (Fig. l(d)), theion temperature Ts(0) = 100 eV (Fig. l(e)). At 25 ms, a specially programmed
equilibrium control system moves the plasma column to the radius R = 53 cm,
during 1.5 ms. At 25.5 ms, switching on an additional capacitor battery provides aplasma current rise up to 120 kA during 4 ms. Then, the current decreases at a ratethat corresponds approximately to the decay of the toroidal field (Fig. l(c)). The
evolution of the plasma parameters was studied for two values of the toroidal field:
BT(0) = 5 kG (Fig. 1) and BT(0) = 7 kG (Fig. 2).

609

IAEA-CN-47/F-m-5

BT = 5 kG

so
55
50

Tt(el/)

100

/0

40

50

t(ms)

FIG. I. Time behaviour of different plasma parameters in discharge with fast shift and current rise at
BT(0) = 5 kG.

The interferometer data yield the chord distributions of the plasma density
at different times (Fig. 3). A comparison between the positions of the n,i(R)maxima and the values of the major radius R obtained from magnetic measurements
demonstrates sufficient accuracy of the electromagnetic diagnostics, even for asignificant column shift. This enables the use of electromagnetic diagnostics in theequilibrium control system.

3.

RESULTSThe current rise disturbs the plasma parameters substantially. During the timeinterval from 25 to 32 ms, the total number of particles, N, increases by a factor of
1.7 in discharges with BT(0) = 5 kG and by a factor of 1.4 in discharges with

610

ASKINASI et al.

B T = 7 kG

ss

H(cm)

so
55

a
s
t
2

nel(/0"cm'*)

e
4

z
200
ISO
{00

50
30

(0

40

SO

SO

tfms)

FIG. 2. Time behaviour of different plasma parameters in discharge with fast shift and current rise at
BT(0) = 7kG.

\ \

35

«5

SS

BT = 5 kG

65

75 a (cm)

SS

*S

55

65

75

FIG. 3. Chord distributions of density during current rise at Bj(0) = 5 kG and B-rfO) = 7 kG.
Curve 1: 25 ms, curve 2: 32 ms, curve 3: 41 ms, curve 4: 32 ms of discharge with major radius
compression without current rise.

611

IAEA-CN-47/F-III-5

lo10

B T = 7 kG

B T = 5 kG
1.6

1.6

1,2

a

0,i

0,6

0,4

OA

20

r(cm)

10

IS

20

p(cm)

FIG. 4. Radial distributions of density, obtained by Abel inversion of profiles in Fig. 3. Curve a:
25 ms, curve b: 32 ms, curve c: 32 ms of discharge with major radius compression without current rise.

BT(0) = 7 kG (Fig. 3). For the sake of comparison, curve 4 (Fig. 3) presents thechord distribution of the density in the plasma after major radius compression without
current rise. N is only 15% higher than at the initial stage. It should be noted thatthe increase in the total number of particles is mainly due to an increase in the
peripheral density (see the radial profile in Fig. 4). The ion temperature also
increases during the current rise as is shown by curves e in Figs 1 and 2. Particularly
fast ion heating was observed in the case of low magnetic field.
Of great interest is the time evolution of the electron temperature profiles,
which can yield the localization of the heat source in the plasma and, therefore,
provides indirect information on the current distribution. Figure 5 shows electron
temperature profiles measured in a discharge with BT(0) = 5 kG at different times.
The most striking feature of the profiles obtained on the current front at 1 and 3 ms
after the start is the noticeable rise in Te at medium radius, whereas the central
temperature is virtually the same as at the initial stage. Then, the Te profile isflattened (t = 6 ms and t = 11 ms) and later on becomes-peaked (t = 16 ms). The
evolution of Te(r) at BT(0) = 7 kG does not show any hollow distributions (Fig. 6).
The fact of a weaker plasma disturbance, in this case, is also confirmed by a lower
loop voltage: Up (B = 7 kG) = 2.5-3 V, Up (B = 5 kG) = 3.0-4.0 V.

4.

DISCUSSION

The experiments have demonstrated the possibility of rapid current input intoa tokamak (dl/dt = 25 MA-s" 1 ). The current rise time (4 ms) is substantiallyshorter than the classical time of current penetration into the central part of the plasma

ASKINASI et al.

612

B T = 5 kG

3

10

IS

10

rfcm)

FIG. 5. Evolution of electron temperature profile during current rise at BT = 5 kG.

B T = 7 kG

too
5

10

<5

20

r/cm)

FIG. 6. Evolution of electron temperature profile during current rise at BT = 7 kG.

IAEA-CN-47/F-III-5

IS

613

10

FIG. 7. Radial profile of ion temperature during current rise at BT = 5 kG.

column. In spite of strong MHD activity during the front of current rise, it is possible
to obtain a stable plasma with a low safety factor. At maximum current,
Ip = 127 kA, major radius R = 53 cm, toroidal field BT(0) = 5 kG, the value of q
at the limiter was equal to 1.7, without toroidal correction. With the correction
included, we find q(a) = 2.
The data show that a narrow electron temperature profile with cold peripheryis necessary for a current rise to take place. Profiles of this kind may be obtainedby minor radius compression [1] or by major radius compression such as in the recent
experiments. This conclusion is confirmed by the fact that attempts to increase thecurrent in the plasma after a smaller shift or without any shift have ended with disruptions in virtually all cases. A more detailed analysis is needed to understand themechanism of current density profile redistribution. An MHD stability analysis of thecurrent density profiles obtained from aneo(r) as well as from transport simulation isin progress now.
As was mentioned above, discharges at a stronger longitudinal field,
BT(0) = 7 kG, are disturbed by current rise to a lesser extent than at
BT(0) = 5 kG. Only 10% of the discharges ended with a large disruption near thecurrent maximum, whereas at BT(0) = 5 kG disruptions occurred in 50% of thecases. The radial distributions of i^ (Fig. 4) and Te (Figs 5 and 6) show that the
disturbances are substantially higher at lower toroidal fields. Large disturbances ofne(r) and Te(r) and higher amplitudes of MHD oscillations at BT(0) = 5 kG may be
due to a flattening of the current density distribution.
A fast increase in the ion temperature following plasma shift and current rise
was observed in part of the discharges (see Figs l(e) and 7). The ion temperature

614

ASKUVASI et al.

up

without Rf

100

10

20

3D

so

eg tfmsj

FIG. 8. Time behaviour of different plasma parameters during ICRF heating (P < 200 kW) in
discharge with fast shift and current rise.

increased by a factor of two, during 3 to 4 ms. Transport simulation and experiment [1] show that classical energy transfer from electrons to ions and neoclassicalion heat conductivity cannot provide such fast heating, typical times of the Tj
increase being, in these cases, 10-15 ms. Some turbulent mechanism seems to be
responsible for fast ion heating during the current rise.

5.

ION CYCLOTRON HEATING

Experimental results on ion cyclotron heating of a plasma in the quasi-stationary
Ohmic heating regime are presented in [5]. At an RF power of P < 200 kW
launched into the plasma by a low field side loop antenna, maximum increase in theion temperature, AT, = 80 eV, was recorded. Pulse duration and power level werelimited by disruptions due to an increase in density during RF heating.

IAEA-CN-47/F-III-5

615

Preliminary experiments on ICH of the plasma formed by shift and fast current
rise were performed. The plasma was more stable; disruptions during RF heatingwere practically absent, in contrast to the previous set of experiments. A maximum
density of % = (3-3.5) x 1013 cm" 3 at a plasma current of Ip = 120 kA wasreached (Fig. 8). At an RF power of P < 200 kW, an ion temperature increase of
AT; = 60-80 eV was observed. The value of AT; is consistent with previous
results. Now, a special coating of antenna and electrostatic shield is being designed
in order to decrease the impurity and hydrogen fluxes originating during the
RF pulse.

REFERENCES
[1]
[2]
[3]
[4]
[5]

BENDER, S.E., et al., in Controlled Fusion and Plasma Physics (Proc. 11th Europ. Conf.
Aachen, 1983), Vol. 1, European Physical Society (1983) 111.
FREDRICKSON, E., et al., in Controlled Fusion and Plasma Heating (Proc. 13th Europ. Conf.
Schliersee, 1986), Vol. 1, European Physical Society (1986) 148.
TANGA, A., et al., ibid., p. 264.
BULYGINSKY, D.G., et al., in Heating in Toroidal Plasmas (Proc. 4th Int. Symp. Rome,
1984), Vol. 2, Int. School of Plasma Physics, Varenna (1984) 961.
ASKINASI, L.G., et al., in Controlled Fusion and Plasma Physics (Proc. 12th Europ. Conf.
Budapest, 1985), Vol. 9F, Part I, European Physical Society (1985) 363.

DISCUSSION
J.-M. NOTERDAEME: You mentioned that there was need to coat the ICRH
antenna. What type of coating are you using?
O.S. PAVLICHENKO: To the best of my knowledge it is a titanium nitride
coating.

IAEA-CN-47/F-III-6

ALFVEN WAVE HEATING AND
CURRENT PROFILE MODIFICATION IN TCA
G. BESSON, G.A. COLLINS1, B.P. DUVAL,C. HOLLENSTEIN, A.A. HOWLING, B. JOYE, J.B. LISTER,
J.-M. MORET, F. RYTER, C.W. SIMM, H. WEISEN
Centre de recherches en physique des plasmas,
Association Euratom-Conf&ie'ration Suisse,
Ecole polytechnique fedeYale de Lausanne,
Lausanne, Switzerland
Abstract
ALFVEN WAVE HEATING AND CURRENT PROFILE MODIFICATION IN TCA.
The increase in electron temperature observed during Alfven wave heating under optimal conditions, near the threshold of the (n, m) = (2, 0) continuum, leads to a large increase in the sawtoothramp rate (~ 3 W/cm3) and to a reduction in the loop voltage for several confinement times. The
increase in electron density during the pulse moves the resonance surfaces outwards, which probably
reduces the heating efficiency. Then, both the temperature increase and the voltage drop disappear. As
the rf power is increased, this process becomes more rapid, and the reduction in the loop voltage lasts
for only a few milliseconds. The ion temperature increase from 300 to 600 eV can mostly be attributed
to the increased electron density, although in some conditions it can precede the density increase, implying some direct ion heating. Besides bulk plasma heating there is evidence of changes in the current
profile which are triggered by the appearance of the resonant surfaces. Both increases and decreasesof the internal inductance have been produced with frequency programming.

1.

INTRODUCTION

The paper describes some of the recent results obtained
during Alfven Wave Heating (AWH) experiments on the TCA tokamak(R,a = 0.61, 0.18 m, B$ = 1.5 T, I p < 170 kA) using the fullset of eight unshielded antenna groups described in detail elsewhere [1,2]. One set of carbon limiters was used for the experiments described. The quasi-cylindrical condition for the Shear
Alfven Wave resonance is given by:w 2 (r) = (n+m/q(r))2 • ( B ^ / M O P W R 2 ) • (1-u)2/(oci2)
in which (n,m) are the toroidal and poloidal mode numbers, respectively. For the parameters of TCA this resonance condition stipulates a frequency in the range 1-5 MHz, and the work reported
here was carried out at 2.3 - 2.7 MHz with n = 2.
It has been theoretically predicted that a substantial
fraction of the input rf power should appear as an inwardly
propagating Kinetic Alfven Wave [3]. We have experimentally
Australian Atomic Energy Research Establishment, Lucas Heights, Australia.

617

618

BESSON et al.
1.0-

r/a
Resonance
Curves

1.ST.D2.qa =3.2^_____
^

—
_ .

0.5-

(a)

l(2rV

n-

[(2.0)((2.1)A DAW

Antenna Loading !mM

100-

DAW
(2.2)

DAW
(2.0) \

•rf..

01 - 6~ip (arb)A _1n"e amplitudenTT-0- 7)e phase

\

-nC1

2

(b)

~-

K.
\\

(c)

>v—
(d)

\

ne(10^m-3)

(e)

i

•

FIG. 1. The main features of the Alfv6n wave spectrum excited on TCA.

detected density fluctuations associated with this wave, during
the heating experiments, and they exhibit many of the features
predicted [4]. Figure 1 summarises the important features of the
AWH scheme as a function of the electron density: Fig. 1a shows
the calculated positions of the resonance layers for n = 2,m = 0, +1,2; Fig. 1b shows the individual antenna loading,varying from 20 - 80 mQ in the continua to 200 mQ for the Shear
Alfven Wave eigenmodes referred to as Discrete Alfven Waves(DAW); Fig. 1c shows the b^ component of the rf wavefieldmeasured behind the limiter radius, also showing the continua andthe eigenmodes; Fig. 1d shows the amplitude of the synchronous
density modulations along a fixed radial chord (r ~a/2) and
Fig. 1e the phase of the radially propagating Kinetic Alfven Waveat this position.
Since the first AWH experiments on TCA, it has been clearthat the excited spectrum strongly affects the macroscopic plasma
behaviour during the rf pulse. The fact that the resonance condition depends on the plasma density, which increases substantiallyduring the rf pulse, has enabled us to make very detailed spectral scans during the rf pulse. However, it has also limited thehigh power study to a power level of 570kW in the (n,m)=(-2,-1)

619

IAEA-CN-47/F-III-6

continuum [2]. Even at power levels below this, we observe apower limitation near the AW mode thresholds which are crossed asa result of the density increase.
2.

PLASMA HEATINGThe flow of energy into the Kinetic Alfven Wave was predicted to heat the electrons via Landau damping. Experimental
evidence for heating of the electrons is obtained on TCA from two
separate measurements: a reduction in the loop voltage and anincrease in the axial electron temperature sawtooth amplitude.
Figure 2 shows a summary of the sawtooth modulation results [5].
We obtained a rate of increase of stored energy on axis,
3/2«e«ne(0)dTe(0)/dt, up to 3 W/cm3, equal to 5-8 times thevolume averaged rf power, suggesting that the electron power
deposition profile must be peaked on axis. The increase incentral heating rate is maintained at ~4.5x<Prf>, even in thehigher density case, n^O) ~ 1 0 2 0 m~ 3 .
Figure 3 shows the evolution of the loop resistance during apulse in which the rf power increases to 200kW, P r f/P o h~70%.

(arb)'

Soft X-ray flux

Te(0)
(eV)

ne(0)=4.9*1013cm~3^ ^ - ^ ~100012.0) continuum
df- 1.6 ms(a)
Prf-150 kW
Wg(0h2.65W-cm'3(Prf)=0.37W-cm~3

900-

sooy^

\

OH

700-

nef0J=10.4»!0!3cm~3

SOO700-eoo:,(2,11 continuum\ ^ ^ \
^~^

(b)

p

rC 255 kW^ ^ • ^ \OH

3

Wet0h3.0 Wcrn
<Prf) = 0.64Wcm'3

ne(0)=3xl013crn~3\
BOO-

L—rA ^ ^ f" ^

700"-

0

0.5
sawtooth phase

1

FIG. 2. Te(0) sawtooth excursion for: (a) the start of the (n, m) = (2, 0) continuum,(b) the start of the (—2,-1) continuum, and (c) a low density target plasma.

620

BESSON et al.

Te (OllkeVl

40

50

60

70
80
90
time Cms]

100 110

FIG. 3. Increase of electron temperature and reduction of loop resistancewith a modest level of rf power.

The density at the start of the rf pulse is chosen so as tomaximise the drop in the loop voltage and corresponds to theonset of the toroidal ly coupled (n,m) = (2,0) continuum(rig = 2 x 10 1 9 m~ 3 , D 2 , 2.5 MHz). The 300 eV increase inT e (0), measured by single-point Thomson scattering triggered at
the middle of the sawtooth period, leads to a reduction of
0.45 V (25%) in the resistive loop voltage, calculated as
V L - Lpdlp/dt. The dashed line through the resistance curve
is proportional to T e (0)~ 3 / 2 . The initial voltage increase
observed corresponds to a transient inductive effect, due tod(p + li/2)/dt. After 17 msec the resistance returns to thetarget plasma value and even increases, consistent with themeasured increase in T e (0)~ 3 / 2 . The electron confinement time,T
Ee ~3.5 msec, is short compared with the timescale of thisevolution. The electron density rises throughout the rf pulse,
forcing the resonance layers towards the outside, as seen in
Fig. 1, possibly explaining the decrease in T e (0). The
appearance of the next continuum threshold (-2,-1) does notproduce a pronounced drop in the loop voltage. This may in part621

IAEA-CN-47/F-ffl-6

50

60

70

80
90
time Ims)

100

110

FIG. 4. Prolonged voltage reduction obtained during frequency pre-programming.

be due to the new surface sharing the rf power with the old
surface, and partly due to the centrally flat resonance frequency
profile of the (-2,-1) continuum.
1
Previously, it was assumed that the increase in P ra a
(Fig. 3) was responsible for the temperature drop towards the endof the rf pulse. The fractional radiated power has now been
reduced to an acceptable level, with Prad/( p oh +p rf)~50~ 60%
up to (Poh+Prf)=0.95MW. It is now difficult to assess whether
the increase in the radiated power is due to more than theincreased plasma density and the reduction in electron
temperature. Measurements of the plasma mass also indicates thatmost of the increase in plasma density is hydrogenic.
It is noteworthy that if both An^/ne' and A P T O T / P O H
are of the order of unity then a curious situation arises. A
scaling law leading to a progressive decrease in confinement timewith additional heating [6], when combined with this large
density increase, would predict a decrease in <T e +Tp. Since< T p already increases considerably, at least by the minimum
resulting from the density increase, <Te> must decrease
dramatically if such a scaling is to hold. The lack ofstationarity with a significant rf power input still prevents us
from accurately assessing the confinement time during the rf
pulse.
Since the application of the rf pulse does not lead to astationary state, the measured loop voltage itself is sensitiveto d(p+li/2)/dt, and fast varying poloidal currents can lead tomisinterpretation, especially when we suspect the presence of modifications of the current profile. We have prolonged the periodof reduced resistive loop voltage to beyond 20 msec (~6 x -egg)
by pre-programming the rf generator frequency at low power.

622

BESSON et al.

Figure 4 shows such a result, in which the spectral position,
given roughly by f 2 ne, has been partly stabilised by ramping
down the frequency at a rate of 17 kHz/msec to compensate for thedensity rise. The plasma current, vertical field current andplasma current position are quite stable. During this period theerror in the loop voltage due to these currents will be small,confirming the reduction, 22% in this case.
As the rf power is increased, the density ramps faster andreduces the time during which d(ngf2)/dt is small enough tostabilise the evolution of the spectrum around the optimal
region. This explains why the most convincing electron heatinq
results have been obtained at the 100-200 kW level whereas the rf
generator has delivered up to 570 kW to the plasma. Work iscurrently underway to increase the useful frequency sweep rangeof the rf generator to resolve this question.
An understanding of the increase in electron density is,therefore, still of extreme importance. The magnitude of theincrease is relatively insensitive to the excited mode structure,or to the number of active antennas and hence also to the antenna
voltages, but it remains a linear function of the total delivered
rf power. It also has an almost linear dependence on the plasma
current. Detailed analysis of its time evolution using azero-dimensional model, dN/dt = S-NA e ff, shows that the causeof the increase, a change in either S or t e ff, remains
unchanged when the plasma current is reduced. The recycling or
pump-out term, which defines the new steady state, increases as
the current is reduced, leading to a smaller density increase butwith a shorter rise time. The H a + D a emission generally
decreases at the start of the rf pulse at high current andremains below the "ohmic" value throughout, suggesting an
increase in particle confinement. This decrease disappears as theplasma current is reduced, and we even observe a small increasein H a + D a for q a > 6.
In the conditions described, we still observe a large
increase in the ion temperature during the rf pulse, typicallyfrom 300-600 eV with ~200kW of rf power. In some conditions,preceding a continuum threshold, we observe a fast increase in
Ti which indicates direct ion heating. However, even at thehighest powers, the ion temperature never exceeds the electron
temperature, suggesting that this direct ion heatinq must be
small.
To summarise, the newest results with naked antennas, carbon
limiters and active frequency control have provided further proofof electron heating using Alfven Waves, by prolonging the periodof reduced plasma loop voltage.
3.

CURRENT PROFILE CHANGESThe predicted local power deposition to the electrons
suggests that AWH might be able to modify the current profile, if
at all possible. Several experimental results from TCA suggestthat this is the case and they are listed below.

623

IAEA-CN-47/F-IH-6

1.25-

75
100
time I ms I
FIG. 5. Evolution of (Sdia and /3 + 1/2 during an rf pulse.

IQOkA--100kW
700mA0

50

WO

150

t Cms]
FIG. 6. Reduction in % following the (2, 0) threshold, deduced from /3 + 1/2.

624

BESSON et al.

0.2-

Aneig

with respect to the (n,m)=(2.0) threshold density
n o u
a
_

3.0

3.2

3.4

3.6 qla)

-0.2-

-0.4-

-0.6-

a Loop voltage
decreaseso MHO activity
destabilised

FIG. 7. Occurrence of the voltage drop and onset ofMirnov activity with respect to the (2, 0) thresholddensity, as the plasma current is varied.

The very rapid rise in P + lj/2 compared with the diamagnetic signal {Fig. 5) suggests a sudden increase in 1^/2 (current
peaking) when the rf pulse is first applied. For a given rf
power, this increase in 1^/2 depends on the position in the
spectrum, being generally larger when the resonance surfaces arein the outer region of the plasma.
The value of P + 1^2 can also decrease during the rf pulse,
also seen in Fig. 5. A case is shown in Fig. 6, clearly
indicating a reduction in l i by more than the value of (3 in theoriginal target plasma. The magnitude of the decrease isestimated assuming different ohmic scaling laws for the increased
density plasma, from which the several curves on the figure arederived. Values in the range -Al^~0.4-0.6 are obtained.
There is a disruption power limit at fixed points in the
spectrum, near the thresholds of the continua, when the increase
in m=2 activity at these points exceeds a critical value. The
increase in m=2 is attributed to a change in the current profilenear the q=2 surface.
There is a change in the details of the DAW spectrum excited, showing up as a hysteresis when the generator frequency isvaried L2J. This has been explained by the dependence on q(r) in
the resonance condition.
The origins of these changes are still not fully understood,
but it now seems that both the thresholds of the continua and thenext outermost surface must play a role.
Figure 7 shows the density difference between the occurrenceof the (n,m) = (2,0) CftW, and a) the occurrence of the drop inloop resistance and b) the increase in m=2 Mirnov activity. The
figure shows that, as the plasma current is varied, the loop
resistance drop always occurs shortly after the appearance of the(2,0) threshold, whereas the onset of the destabilization of the
MHD activity varies with respect to this threshold. Calculations
using assumed plasma current profiles show that the increase inIAEA-CN-47/F-III-6

625

MHD activity in fact coincides with the approach of the outerresonance surface (2,1) to the estimated position of the q=2 surface, illustrated in Fig. 1.
All these observations indicate that the radial positions of
the Shear Alfven Wave resonance surfaces influence the current
profile. Much of the discontinuous behaviour observed during the
AWH pulse, and documented in detail elsewhere, could be attributed to these profile changes.
ACKNOWLEDGEMENTS: We gratefully note the continued interest ofProfessor F. Troyon in our work, as well as the support of thewhole TCA team. This work was partly funded by the Fonds National
Suisse de la Recherche Scientifique.

REFERENCES[1] COLLINS, G.A., et al., Phys. Fluids 29 (1986) 2260.
[2] BESSON, G., et al., Plasma Phys. Controll. Fusion 28 (1986) 1291.
[3] HASEGAWA, A., CHEN, L., Phys. Rev. Lett. 35 (1975) 370.
[4] BEHN, R., COLLINS, G.A., LISTER, J.B., WEISEN, H., Plasma Phys. Controll. Fusion 29(1987) 75.
[5] JOYE, B., et al., Phys. Rev. Lett. 56 (1986) 2481.
[6] GOLDSTON, R.J., Plasma Phys. Controll. Fusion 26 (1984) 87.

DISCUSSION
M. ONO: What is the ion temperature behaviour during the regime of large
density increase?
J.B. LISTER: The ion temperature increases typically by 300-400 eV, as wehave always reported. The question is really whether this is due only to the increasein density, via an increase in Pei, or whether the RF delivers power directly to theions. At the 11th European Conference in 1983 we reported a value of Tj greaterthan the Ohrnic value, and at the 12th European Conference in 1985 we reported
values of dW;(0)/dt > Pei(0). The errors in these analyses are large, however, andwe consider the question of direct ion heating an open one.

IAEA-CN-47/F-IV-l

ANALYSIS OF ICRF HEATING
AND CURRENT DRIVE IN TOKAMAKS
A. FUKUYAMA, T. MORISHITA
Department of Electronics,
Okayama University,
Okayama
Y. KISHIMOTO, K. HAMAMATSU
Japan Atomic Energy Research Institute,Naka-machi, Naka-gun, Ibaraki-ken
S.-I. ITOH
Institute for Fusion Theory,
Hiroshima University,
Hiroshima
K. ITOH
Plasma Physics Laboratory,
Kyoto University,
Uji, Kyoto
Japan

Abstract
ANALYSIS OF ICRF HEATING AND CURRENT DRIVE IN TOKAMAKS.
To analyse ICRF heating and current drive processes in tokamak plasmas, a theoretical modeland computational codes are developed. Wave excitation, propagation and absorption, high energy tailformation, thermalization of the absorbed power, and current generation are analysed simultaneously,
and the temporal and spatial evolutions of bulk plasma parameters are calculated in a consistent mannerand in a realistic geometry. This wave thermalization transport analysis enables a comparison ofmodelling with experimental results. Through the kinetic treatment of the wave in the ICRF heatedplasma, the good accessibility, the controllability of the deposition profile, the sustainment of the highenergy particles in the core region, and the current drive ability in the high density regime areconfirmed. A study on confinement time and MHD stability is presented. It is found that localized,
intensive heating increases the central pressure and invalidates the stability criterion of the internal kink
mode.

1.

INTRODUCTION

Ion cyclotron range of frequency (ICRF) waves have been extensively used
as a reliable heating tool and have become a promising candidate for current drive
in tokamaks. Analysis of their properties is substantial for an explanation ofexperimental results and for designing future programmes. To analyse the heating
627

628

FUKUYAMA et al.

and current drive processes, we have developed a theoretical model and the computational codes associated with it. Three stages: (1) wave propagation and absorption,(2) relaxation in a velocity space, and (3) spatial transport are systematicallyanalysed according to different time-scales. For given launching conditions, wave
field structure, high energy particle generation, power deposition profile, time evolution of temperature profiles and driven current are evaluated in a consistent manner.
We examine the confinement characteristics (e.g. the confinement time andattainable temperature), the stability of the internal kink mode and the optimum
conditions for the current drive.

2.

WAVE EXCITATION AND PROPAGATION

We solve Maxwell's equations, using the kinetic conductivity of an
inhomogeneous and multicomponent plasma [1, 2]. The following linear processes
are described simultaneously: coupling to finite size antennas, wave propagation(reflection, refraction and tunnelling through evanescent layers), mode conversion to
ion Bernstein waves, cyclotron damping up to second harmonics, Landau damping
as well as transit time magnetic pumping (TTMP). Figure 1 shows the poloidal structures of the wave field (a) and the power deposition profile (b) for two ion hybrid
resonance heating in JET (D majority, H minority) [3]. We observe that good accessibility to the central region is obtained, and that 80% of the injected power isabsorbed near the plasma centre (r S 0.2 m). When a high energy tail componentof the minority ion exists, the deposition profile broadens as is shown in (c).

(a)

(c)

FIG. 1. Poloidal structure for two-ion hybrid resonance heating in JET (B, = 2.04 T,o>/2w = 29.8 MHz, ne0 = 2.94 x 1Q19 m'3,
To = 2.5 keV, nH/nD = 0.1, monopole antenna),(a) Wave electric field, Im Ey(r,8); (b) power deposition profile to protons; (c) power deposition
profile in the presence of a high energy tail component near plasma centre, r £ 0.3 m.

629

IAEA-CN-47/F-IV-l[kev][kW/m3]

1.13x10'

(a)
T

Ht

0

[kW/m3]P e / 2 :C0NPUTEDr(b). P e EXPERIMENT'
P . :COMPUTED WITH
f \ V '.SPATIAL TRANSPORT[m] 0.6

FIG. 2. (a) Radial profiles of nHl and TH, of hot component at power level of 0.66 MW.
n
H^nD = 0.04; other parameters are the same as in Fig. 1; (b) radial profiles of power absorbed byelectrons: calculated in the wave thermalization analysis (solid line), computed by coupling with 1-Dtransport code (dashed line), evaluated from temperature rise in JET (solid circles); (c) Bo dependenceof the absorbed power density. Arrows indicate positions of fundamental cyclotron resonance of proton.

3.

MODIFICATION OF THE VELOCITY DISTRIBUTION

The velocity distribution of the resonant particles in a heated region is strongly
modified, and this modification affects wave propagation and absorption. We
describe the time evolution of the local velocity distribution, taking account of thechange in the wave structure. The Fokker-Planck analysis [4], including collisionaland quasi-linear diffusion, is combined with the 1-D ICRF wave code [1]. We calculate tail formation and power partition through collisional relaxation [5]. In the caseof two ion hybrid resonance heating, a high energy tail of the minority ion velocity
distribution appears first near the cyclotron resonance. The density nH, and thetemperature THt of the hot component of the minority ions increase, and the widthof the heated region broadens in a collisional time-scale. Figure 2(a) shows the radial
profiles of nHt and THt, together with the minority density nH, for the JET heating
experiment. The power deposition profile in Fig. l(c) is computed on the basis of
the distribution of the hot component. The power transferred to electrons throughcollisions is enhanced near the plasma centre with the increase of the hot component.
The spatial profile of the power absorbed by electrons is shown in Fig. 2(b). Solidcircles denote the value derived from the temperature rise in JET [3]. The dashed line
indicates the initial temperature increment computed by coupling with the 1-D transport code, which is in good agreement with the experiment. Deposition control byfrequency change is illustrated in Fig. 2(c) for fixed RF power. As the cyclotron
resonance shifts from the plasma centre, the absorbed power density and the tailtemperature decrease and the power partition changes.

630

FUKUYAMA et al.

[m]

FIG. 3. (a) Temporal evolution of central temperatures, Te0 and T^, proton energy content, WH, andgrowth rate of internal kink mode, y. Bo, na, co are the same as in Fig. 1, nH/nD = 0.04 and
Pxr = 3 MW; (b) electron temperature profile at 0 ms, 100 ms and 1800 ms.

4.

SPATIAL TRANSPORT

To study the long time development of plasma profiles, the power depositionis linked to the 1-D transport code. For a given density profile, Te(r,t), TD(r,t) andpoloidal magnetic field Bp(r,t) are computed. An electron energy diffusion coefficient of the ALCATOR-INTOR form (xe = 1 X 1019 n^ 1 m 2 -s~') is employed.
As for the ion energy diffusion coefficient, we use Xi — 0.5 [m^s" 1 ] + XCH after
Doublet III results [6], where XCH is the neoclassical value calculated [7]. Neoclassical resistivity is employed and Zeff = 2.29. These coefficients are chosen to
reproduce the Ohmic plasma in JET. Figure 3(a) shows the time evolution of thecentral temperatures, T^ and TDQ, and the minority energy content WH, in the caseof 3 MW ICRF input with Ip = 2 MA. The Te profiles in the Ohmic phase, in theinitial phase of RF heating and in the saturation phase, are shown in Fig. 3(b). In theinitial phase, Te increases mainly in the central region (r :£ 0.3 m). The ratio ofcentral to average temperature is about 3.4 for t > 250 ms. The confinement timein the saturation phase is 625 ms, which is comparable to that of the Ohmic phase.

5.

MHD ACTIVITY

The MHD stability is affected by the localized energy deposition and by the fastions due to the ICRF heating. We study the stability of the m/n = 1/1 internal kink

IAEA-CN-47/F-IV-l631

mode (m/n: poloidal/toroidal mode numbers). We employ the simplified stability
criterion, 5WT > 0 [8], where
5WT = 3v Aq { 1 3 / 4 8 ( K + 4 ) - 0 P [(rj/r,)'- 4 -l]/(»—4)}

with
0P m - P r t / B j (r,)] l" (r/r,) 2 (dp/dr) drJo
q(r,) = 1

and

q(r2) = 2

The safety factor is fitted as q(r) ~ 1 - Aq [l-(r/r,)*] near the axis.
The time evolution of 5WT is given by the wave thermalization transport calculation. With the injected power of 3 MW, we find that j8p increases rapidly and thatthe mode becomes unstable with a delay time of 35 ms after the onset of RF heating(see Fig. 3(a)). The growth rate y is about 104 s" 1 . This pressure driven internalkink instability may be the origin of the rapid internal disruption which is notaccompanied by the precursor oscillations. This delay time seems to be shorter thanthe experimentally observed period of the giant sawtooth. If we include the finite
gyroradius stabilization effects of the tail ions [9], the critical /3P value increases,
and the delay time may be longer by a factor of two. We note that the critical j3p
value depends weakly on nH/ne. If we assume that Te(r) is limited by this instability,
the energy confinement time reduces to the order of 100 ms for the parametersof Fig. 3.

6.

CURRENT DRIVE IN HIGH DENSITY PLASMA

The current drive by using the phase shifted antenna array in the toroidal direction is examined [10]. The Fokker-Planck equation for electrons at each spatial point
is combined with the ICRF 1-D wave code [1]. The stationary distribution function
is given by

F(u) = C exp-

u du/(l + D w /D c )

L Jo

J

where u = v/ve (ve is the electron thermal velocity) and D w and D c are quasi-linear
and collisional diffusion coefficients, respectively [10]. The current is driven by
Landau damping and TTMP. Application to JET plasma is given, assuming thatseven antennas are installed on the low field side. Figure 4(a) shows the induced
current profile and the normalized efficiency t\ = (J(x)/Pe(x))(mvev0/nee) [11]632

FUKUYAMA et al.

[A/m
Ixio'

-1.2

F/G. 4. (a) Induced current profile and normalized efficiency. B, = 2.165 T, a/2ir = 43 MHz,nt0 = 8 X 1O19 m'3, To = lOkeV, nH/ne = O.I; (b) phasing and spacing dependence of global
efficiency Va for a/2ir = 33 MHz and ne0 = 1O20 m'3. A: hi = 0.126 m, B: Az = 0.188 m,
C: Az = 0.295 m, D: Az = 0.377 m, E: Az = 0.589 m.

(antenna spacing Az = 0.188 m, phase difference A<f> = 2TT/9). At the density ofiieo = 8 x 1019 m" 3 , the current is driven near the plasma centre. About 85% of
the incident power is absorbed directly by the electrons. The global efficiency
T/G 3 IRJJ/PRF is 0.025 A/W.
Phasing and spacing dependences are examined (Fig. 4(b)). It is found that Az
has an upper limit to avoid negative current, which is generated by the subpeaks of
the radiated RF spectrum. Once the spacing is properly chosen (such as the cases
A, B and C), the efficiency is insensitive to the number of antennas. The A<£
dependence of rjc is weak for ir/4 :S A</> :S 3x/4.

7.

SUMMARY AND DISCUSSION

We have studied an ICRF heated plasma in a tokamak. The wave thermalizationtransport analysis enables us to deal quantitatively with transport phenomena andstability of the ICRF sustained tokamak plasmas. Through the kinetic treatment of
the wave in the ICRF heated plasma, high antenna loading, good accessibility,
controllability of deposition profile, sustainment of high energy particles in the coreregion, and current drive ability in the high density regime are confirmed for present
day machines and future designs. The wave thermalization transport code simulates
the experimental observation of temperature rise and confinement time. It is foundthat the central ICRF heating is responsible for the internal kink instability, which
may limit global confinement. A future task is to investigate the transport coefficientof ICRF heated tokamaks by applying the analysis to large tokamak experiments.

IAEA-CN-47/F-IV-l

633

ACKNOWLEDGEMENTS
This work was supported by a grant in aid for fusion research and scientificresearch from the Ministry of Education, Science and Culture of Japan.

REFERENCES
[1]
[2]
[3]
[4]
[5]

[6]
[7]
[8]
[9]
[10]

[11]

FUKUYAMA, A., NISHIYAMA, S., ITOH, S.-I., ITOH, K., Nucl. Fusion 23 (1983) 1005.
FUKUYAMA, A., GOTO, A., ITOH, S.-I., ITOH, K., Jpn. J. Appl. Phys. 23 (1984) L613;
FUKUYAMA, A., ITOH, K., ITOH, S.-I., Computer Phys. Rep. 4 (1986) 137.
JACQUINOT, J., ANDERSON, R., ARBEZ, D., BARTLETT, D., BEAUMONT, B., et al.,Plasma Phys. Controll. Fusion 28 (1986) 1.
STIX, T.H., Nucl. Fusion 15 (1975) 737.
MORISHITA, T., FUKUYAMA, A., HAMAMATSU, K., ITOH, S.-I., ITOH, K., Analysisof the Minority Velocity Distribution and the Power Deposition Profile in ICRF Heating inTokamak, Rep. HIFT-124, Inst. Fusion Theory, Hiroshima (1986).
GROEBNER, R.J., PFEIFFER, W., BLAU, F.P., BURRELL, K.H., FAIRBANKS, E.S., etal., Nucl. Fusion 26 (1986) 543.
CHANG, C.S., HINTON, F.L., Phys. Fluids 25 (1982) 1493.
BUSSAC, M.N., PELLAT, R., EDERY, D., SOULE, J.L., Phys. Rev. Lett. 35 (1975) 1638.
ITOH, K., ITOH, S.-I., TUDA, T., TOKUDA, S., J. Phys. Soc. Jpn. 53 (1984) 1759.
KISHIMOTO, Y., HAMAMATSU, K., FUKUYAMA, A., ITOH, S.-I., ITOH, K., ICRF
Current Drive by Use of Phase Control of Antennas, Rep. HIFT-120, Inst. Fusion Theory,
Hiroshima (1986).
FISCH, N.J., Phys. Rev. Lett. 41 (1978) 873.

IAEA-CN-47/F-IV-2

DOUBLE HELIX (] x b) CURRENT DRIVE
IN THE RF TOKAMAK DEVICE RYTHMAC
M.J. DUTCH, A.L. MCCARTHY
School of Physical Sciences,
The Flinders University of South Australia,
Bedford Park, South Australia,
Australia

Abstract
DOUBLE HELIX <J x b> CURRENT DRIVE IN THE RF TOKAMAK DEVICE RYTHMAC.
Travelling wave fields in m = 1, n = 4 double helix structures surrounding a toroidal plasma
vessel are used to drive toroidal and poloidal plasma currents by means of the non-linear Hall effecttechnique <j x b> in high power experiments of short duration at 330 kHz. The evolution of the magnetic configuration has been studied using magnetic probe data to calculate and plot the poloidal flux
function. Contours of equal driven toroidal current extracted from the same probe data reveal negative
toroidal current at a small radius and at early times. This observation is consistent with the model
presented. The rapidly rising quasi-steady driven toroidal current distributed over the minor section creates a changing magnetic flux and thus a back electromotive force (emf) observable as the loop voltage.
This emf drives negative current along the path of minimum impedance. The configuration approaches
a steady state where this back emf and negative current die away without any evidence of disruptions.
A series of experiments with 40 ms of current drive is planned to study the equilibrium and stabilityof the configuration on this longer time-scale, beyond the start-up phase studied here. Extension of the<j x b> RF current drive technique to an m = + 1 and — 1 helical mesh antenna which drives purely
toroidal current with no poloidal current is presented. The data for this case are contrasted with thosefor the m = +1 double helix antenna and the m = — 1 double helix antenna applied separately, anda theoretical discussion of the drive by this helical mesh antenna is given.

1.

INTRODUCTION

To achieve steady operation of a tokamak reactor it isnecessary to develop continuous toroidal current drive. Manyexperimental groups have investigated RF techniques for suchcontinuous current drive and research in this field has beenreviewed recently by Jones [1],We present here data and discussion which gives a detailed
account of the start-up phase of our m =.1 double helix driven
RF tokamak RYTHMAC, shown schematically in figure la, whose
experimental operation and theoretical basis in terms of thenonlinear Hall effect <j x £> have been previously reported[2,3].
This configuration,which has been reported by Hotta et aL[4] todrive toroidal current, is shown here to drive toroidal andpoloidal current; the poloidal current, which interacts with the

635

636

DUTCH and MCCARTHYh.

FIG. 1.
a. Schematic diagram showing the m = 1 double helix structure used to drive toroidal and poloidal
currents.
b. Line generator waveforms at 2 us per division, showing phase difference and reproducibility.
c. Line generator waveforms, 10 us per division.
d. Rogowski belt measure of toroidal current, 10 us per division, 500 A per division. The spike at
5 y.s is spark gap noise at the time the line generators are triggered. Figures 1b and Id have two
shots superposed to show the reproducibility.
e. Loop voltage; the vertical scale is 10 V per division.
f(i), f(ii). Poloidal flux contours at times r, and t2 marked in d.
g(i), g(ii). Contours of equal toroidal current density at times t} and t2 marked in d. Heavy contours
indicate positive current, light contours negative current. Below are the current profiles in the
median plane,h. Contours of equal AB+ caused by driven poloidal current.

IAEA-CN-47/F-IV-2

637

toroidal field, will take part in the equilibrium and stability
and may need to be controlled independently of the toroidal
current.
We are led by our detailed study of the m = 1 double helix
current drive experiments to develop a new antenna configuration
which we call the m = + and -1 helical mesh antenna, shown
schematically in figure 2 in a cylindrical geometry large
aspect ratio approximation for clarity. We report observations
which demonstrate experimentally that this antenna drives
purely toroidal current without poloidal current as does the
Synchromak configuration but with an m = 0 antenna [5].
In the section entitled model and theory we give a resume*of the <j x b> technique of current drive for the m = 1 double
helix case.
For the m = + and -1 helical mesh antenna wedetermine the spatial and temporal dependence of the vacuum
magnetic fields. The motion of these fields gives insight into
the current drive of this device.

2.

MODEL AND THEORY

2.1. m = 1 Double Helix Antenna
To describe the mechanism of steady current drive in our
experiments in the simplest model, the ions are assumed
immobile and uniformly distributed, and the electrons aretreated as an inertialess, pressureless negatively charged fluid.
We ignore steady applied magnetic fields. " The generalised
Ohm's law is written:
1
E = r|J + — J x B(1)
The RF currents in the helical structure of fig. la induce
screening currents in the plasma which try to exclude the RF
fields of the structure.
But finite resistivity permits thesefields to penetrate.
The screening currents j follow helicalpaths parallel to the structure currents and because of resistivity have phase different from that of the RF magnetic fieldb of the structure.
There will thus be a nonlinear Hall effectterm consisting of a steady part and a part of frequency 2w.
We include the steady parts of these nonlinear Hall effectterms in the components of Ohm's law, using angular brackets todenote the time averages:
1

fy
E

$
J

ne

e= B*^KW^

638

DUTCH and MCCARTHY

FIG. 2.
a. Schematic diagram of them = +1 and -1 helical mesh antenna viewed horizontally. The arrows
on the windings show the direction of the RF structure currents at the instant when the current ismaximum in phase 1 and zero in phase 2. The symbols © and ® in the plasma vessel indicate
the structure of field 6. The nature of the screening currents J induced in the plasma is represented
by the dashed paths.
b. The same structure as in a, one quarter of a period later. Now the RF current is zero in phase 1
and maximum in phase 2.
c. Horizontal cut through the minor axis as seen from above, showing the vacuum travelling magnetic
field pattern produced by the helical mesh antenna.
d. Oscillogram data for m = + 1 double helix antenna.
d(i), e(i), f(i). Steady driven toroidal current 1^, recorded at 10 /is per division. The vertical scale is
450 A per division.
d(ii), e(ii), f(ii). Change in the toroidal magnetic field on the minor axis, kB^O, 0), caused by thedriven poloidal current recorded at 10 /is per division. The vertical scale is about 0.004 Tper
division.
e. Data for the m = —1 double helix antenna.
f.
Data for the m = +1 and —1 helical mesh antenna.

639

IAEA-CN-47/F-IV-2

From these equations we see that even in the absence of imposed
electric fields, such as the inductively coupled E§ of theTokamak,we would expect the nonlinear Hall effect term to balance
the resistive terms to drive not only toroidal current but alsopoloidal current.
We have presented a complete treatment of
this problem in helical coordinates, with some analytic andsome numerical solutions [3].

2.2. m = + and -1 Helical Mesh Antenna
We represent the structure currents shown in figures 2aand 2b in the^natural helical coordinate systems (r, x+i £+iand (r, x _ p 8_i) for the m = +1 and m = -1 conductors,
respectively [3].
6 ( r-r o )which can be written in cylindrical coordinates as:

i
J

ext

- Re

I

i(wt+9+kz) '-kr£+z'
k r
2 "

,

o I

e
+2

i(wt-6+kz) kr §+
e
k r

[

1«Cr-rr )o00

and we sum to get

•'ext

k

e
kr sinCwt+kz] since) 0 + cos(a>t+kz)cosCe) z
cr

o

r

o
(4)

I e measures the strength of the RF currents in the helicalconductors located at radius r 0 and of pitch length £=27r/ic;

ko = C k 2 * ^2) 2 .
r

Maxwell's equations in helical coordinates give the total
structure fieldb = Re iB i;."(kr)'le
B k
0) O

+ e

>r

i(tot+x+1).
- e

640

DUTCH and MCCARTHY

where B(0 measures the strength of this RF field at r=0 and is
proportional to I e .
Ii(kr) is a modified Bessel function oforder 1.
Expressing this in cylindrical coordinates and summing we
get:I'(kr)sin(wt+kz)cos(6) r - —=-b = 2Bsin(urt;+kz)sin(6) G0)+ I1(kr)cos(wt+kz)cos(0) z(5)from which we may note immediately that all three spatial components of the field progress as a wave on the structure in the axial
direction,as indicated by the sin(wt+kz) dependence, but do not
progress in the poloidal direction 9. Figure 2c shows apartial representation of b.
Immediately we see from the model of electrons being
entrained to move with the field linesf provided jjp- << w << —— ),while ions remain essentially at rest, that net current can be
driven in the axial direction,but no current will be driven in
the poloidal direction.
A full analysis using the interaction of the screening
currents with the structure fields will be presented elsewhere.
3.

EXPERIMENTS

3.1. m = 1 Double Helix antenna experiments
The experiments were performed in the RYTHMAC device which
has a pyrex glass toroidal vessel of major radius 25 cm andminor radius 5 cm. A DC toroidal field of 0.03 Tesla and a
vertical field of 0.008 Tesla was applied. The working gas
was Argon at 1 millitorr. No preionisation was used.
The antenna is fed by two RF sources which are modified
Weibel line generators, giving 16 periods of 330 kHz radio
frequency current at a power of about 3.5 Megawatt per line.
The duration of each shot is about 50 us.
An externally integrated Rogowski belt measures the total
toroidal current. Hall probes which compare the verticalfield produced by the driven current I<j> at the centre of thetorus with the DC vertical field give results in excellent

IAEA-CN-47/F-IV-2

641

agreement. Poloidal current as well as toroidal current isdriven by the double helix antenna structure[3].
In figure lb,c we display the RF line generator waveforms
and in figure Id we see the quasi steady driven toroidal current
as measured by the Rogowski belt. It is important to notethat two shots are overlapped in fig. lb,d>which demonstrates
a high degree of reproducibility in the data allowing us to
study the plasma configuration by analysing data from several
hundred shots. The driven toroidal current in fig. Id isapproximately 1.5 kA.
The magnetic configuration has been studied by measuring
the B r and B z components of the magnetic field on a grid ofpoints in the minor section;from this set of data thepoloidal flux function has been calculated.
Flux contours forthe configuration produced are shown in fig.If. The driven
toroidal current extracted from the magnetic probe data shows,in fig. lg, negative current at small radius at early times.
This is the result of the back emf caused by the flux of the
distributed driven toroidal current driving negative current
on the path of minimum impedance.
A single-turn loop is used to measure the back emf (seefigure le). For a vacuum shot with full RF current drive
applied the loop voltage V L is of course zero,unlike the loop
voltage with a vacuum shot for an inductively driven tokamak.
The loop voltage to be observed for a non-inductively driven
tokamak is a result of the changing flux created by the rapidly
rising driven toroidal current. For a non-inductively driven
tokamak the forward driving term is localised while the backf-d$R

emf V L = JE.dl = , is not. The spatial distribution ofthese terms need not be the same. On the shortest toroidal-d$R

p a t h s , E. in / E . d £ =ican exceed —~~<jQ br>of equation (2)and negative current will be driven in that part of theminor section. For inductively driven tokamaks, only the net.
-d$g
flux changes in the core are important and JE.dt = < has thesame sign on all toroidal paths within the plasma.
In this experiment the RF is maintained for a relatively
short time because the sources are line generators, and theloop voltage goes to zero as the driven current maximises; theloop voltage changes sign when the driven current falls. Thisback emf driven negative current dies away as the configuration
evolves.
If the RF drive were to be maintainedrthis transientphenomena would cause no problems, but during the start-up
phase the interaction of this current contributes to the forces
which will determine the equilibrium and stability.

642

DUTCH and McCARTHY

Of more concern for this m = 1 double helix current drive
configuration in determining equilibrium is the role played by
the driven poloidal current.
Evidence for the existence of
this current is given in fig. lh as a contour plot of equalAB^, the change in toroidal field created by the driven
poloidal current.
It is important to the discussion that themeasured magnitude of the AEU requires the poloidal current
density Jg to be approximately the same as the driven toroidal
current density JA,as expected from the theoretical analysis[3],It follows that the interaction of J0 with the imposed B^ is avery important term in determining equilibrium,giving approximately three times as large a force as the interaction of thetoroidal current with the appropriate vertical field forequilibrium in the absence of driven poloidal current. Unlike
the interaction of the negative toroidal current in the startup phase, the interaction of this driven poloidal current withthe toroidal field provides a term in the equilibrium equations
which is maintained continuously.
Accordingly we sought tocontrol this.

3.2. m = + and -1 Helical Mesh Antenna Experiment
For these experiments the RYTHMAC machine and plasma conditions were essentially the same as in 3.1,except that them = + and -1 helical mesh antenna shown schematically in fig. 2
was built from superposed m = +1 and m = -1 double helix
antennas.
By making different connections, the windings canbe operated as just an m = +1 or just an m = -1 antenna as wellas the helical mesh antenna.
The impedance presented to eachof the RF sources differs in the different configurations, but
we choose the operating conditions such that for all the data
presented the magnitude of the RF currents in the antenna
windings is approximately the same.
For the helical meshantenna,one line generator phase feeds one pair of the m = +1
coils in series with one pair of the m = -1 coils, while theother RF source delayed by one quarter of a period feeds theother pairs similarly.
The data presented compare the results for driven
toroidal and poloidal current from the three different antennaconfigurations; m = +1 double helix in fig. 2d, m = -1 double
helix in fig. 2e and the m = + and -1 helical mesh antenna in
fig. 2f.
The toroidal current is measured using a calibrated
Rogowski belt.
Poloidal current is monitored using an integrated wirewound probe placed at the centre of the minor section
to measure the change in the toroidal field ABx(0,0).
For all three cases the sign of the toroidal current, I<j,,is the same; that driven by the helical mesh antenna is slightly

IAEA-CN-47/F-IV-2

643

greater in magnitude. The AB<j)(O,O) caused by the poloidal
current driven by the m = -1 double helix antenna is opposite
in sign to that for the m = +1 case, indicating that thepoloidal currents flow in opposite directions for these two
configurations. The helical mesh antenna makes a much smaller
AB^fOjO), consistent with a small driven poloidal current
relative to that for the other two antennas.
4.

SUMMARY

We have made detailed studies of the start-up phase of them = 1 double helix <j x b> RF current drive configuration. In
approaching problems encountered there,we have built andsuccessfully tested an antenna for <j x b> RF current drive
which produces essentially pure toroidal current of magnitude
approximately 900 Amps for the conditions chosen, without
driving poloidal current.
ACKNOWLEDGEMENTS
The authors are grateful for the loan of the RF sourcesfrom Professor I.R. Jones. This work has been supported by
grants from the Australian Research Grants Scheme and theAustralian Institute of Nuclear Science and Engineering.

REFERENCES
[1]
[2]
[3]
[4]
[5]

JONES, I.R., Comments Plasma Phys. Controll. Fusion 10 3 (1986).
DUTCH, M.J., MCCARTHY, AX., Plasma Phys. Controll. Fusion 28 (1986) 695.
DUTCH, M.J., MCCARTHY, A.L., STORER, R . G . , Phys. Rev. Lett. 56 (1986) 1563.
HOTTA, E., SUZUKI, M., OHTA, H., HAYASHI, I., Jpn. J. Appl. Phys. 24 (1985) 110.
FUKUDA, M., MATSUURA, K., HIRANO, K., MOHRI, A., FUKAO, M., J. Phys. Soc. Jpn.
41 (1976) 1376.

IAEA-CN-47/F-IV-3

DEPENDENCE OF
POWER ABSORBED IN A PLASMAON ALFVEN HEATING ANTENNA PARAMETERS
A.G. KIROV, V.P. SIDOROV, A.G. ELFIMOV,S.N. LOZOVSKIJ, L.F. RUCHKO, K.G. KOMOSHVILI,
V.V. ONISHCHENKO, V.V. DOROKHOV, A.V. SUKACHEV,
V.D. SAMUSHIA, L.Ya. MALYKH, N.L. GULASARYAN,
D.A. VOJTENKO
I.N. Vekua Institute of Physics and Technology,
USSR State Committee for Utilization of Atomic Energy,
Sukhumi, Union of Soviet Socialist Republics
Abstract
DEPENDENCE OF POWER ABSORBED IN A PLASMA ON ALFVEN HEATING ANTENNAPARAMETERS.
The results of theoretical and experimental studies on plasma heating by Alfv6n waves arepresented for various types of antennas and their parameters. The relation between central andperipheral heating is analysed. Analytic expressions are given for slow Alfve'n wave amplitudes on theplasma edge. The new results of the numerical computations are analysed, and conditions for RF power
input into the central plasma region for devices of different sizes are formulated. For the R-O device,a high heating efficiency of the poloidal antenna is demonstrated, and the electric field amplitudes inside
the plasma column are measured.

1.

INTRODUCTION

Alfven heating efficiency and the possibility of controlling the plasma
parameters by an Alfven wave depend on the ratio between the energy input by theslow Alfven wave at the plasma edge and that by the kinetic Alfve'n wave within theconversion zone; these properties also depend on the degree of proximity of the'helical' resonance frequency to the lower boundary of the Alfve'n continuum, when
the RF power input occurs near the plasma centre and the active antenna impedanceis highest. These conditions depend strongly upon the antenna parameters (m, n).
These items have already been considered and even partially solved, boththeoretically and experimentally [1-6]. In this paper, we present our new results on
this subject.

2.

ANALYTIC CONSIDERATION

RF fields excited by an antenna with RF currents, ~ exp i(Qt - m<p — k2z),
in a plasma cylinder of radius a and length 2irR placed in a helical magnetic field
645

646

KIROV et al.

Bo were determined by means of a technique described in Refs [7, 8]. The assumptions that are made and the notations we apply are the same as in Refs [7, 8]. Except
for a certain region near the conversion point rA, where €| = N? (Nj| = kjC2/122,kn = kz (1 + m/nq)), n = kzR is the toroidal wavenumber, the amplitudes of the
RF electric field components in the plasma may be represented in the form
Sj = Sj + 8j, where 8/(r) is the 'smooth' part of the field connected with the fast(MHD) wave and determined from the solution of a second order MHD equation[9, 10], and £j(r) is the 'fast alternating' part of the field connected with the slow
Alfven wave of small wavelength \ A or damping depth <5S (5S, XA/2TT -^ A, A :£ ais the scale of plasma non-uniformity). In particular, in the 'cold' plasma region,
where vre < vph = fi/kz, rA < rph < r < a [10] (rph is the 'phase resonance'
point, where vre = vph), we find, by a WKB approach and assuming that the conditions M/lml <^ 1, €i(a)/|m|62(a) > \A/27ra, vei ^ 0 are fulfilled:- (me-2/e,)£,]r=a (f(a)/f(r)) (^(a)V(r))

(1)

where
f(r) =
r

= sin
Jfph

XA(r) =h =tg£

n + m/q= XA (rph2r+ O)/2TT8 S (rph-0)

for a 'hot' peripheral plasma (vTe > v ph ),<p(r) = exp

dr

Ss(r)

Assuming that |m|€2(a)/ei(a) ^ 1, ( r S ^ ~ £ v(a) and taking, for simplicity,q = const, we obtain from (1) that the ratio of slow and fast wave amplitudes at
r = a is of the order of647

IAEA-CN-47/F-IV-3

2m
no(rA)
no(a) m + nq(a)(2)(no(r) is the plasma density for a hot peripheral plasma, XA/2ira — 8s/a).
We see from Eq. (2) that, for sufficiently low plasma density at the boundary,
the amplitude Sr(a) (but not S^,) may reach high values. This effect is essentiallyconnected with the stationary current j z and disappears when jz(a) = 0.

3.

NUMERICAL COMPUTATIONS

The antenna active impedances were calculated in cylindrical geometry within
the kinetic approach, using the one-dimensional numerical code EPSI, analogous
to [3]. It is shown that the maximum impedance and the best RF power input localization occur at |Ic^a] = 2.5 for devices with various aspect ratios: INTOR ( = 4.2;
circular cross-section), TFTR («3.0) and T-10 ( = 5.1).
In Ref. [4], the RF power absorption maximum had been found, in the MHD
approach, at |kza| = 2.5 for m = - 1 . Here, this effect is found for modes with
m = + 1 , - 2 , - 3 , - 4 . Figure 1 shows an example of the active plasma impedance
versus the position of the local Alfve'n resonance point XA. The computations havebeen carried out for INTOR, with the plasma density distributed parabolically andthe electron temperature profile given by Te(x) = To {1 - [1 - (T,/T0)1/Y]x2}1',where 7 = 4/3, T t /T 0 = 0.025, q(a) = 2.19. Figure l(b) shows the impedance
curves for the m = +1 mode. We see that the direction of wave rotation does not

(b)

100

n = 10 12-,

/A\ -

J

50

'A

p = m e /m|

=1

•

\

K \

m

1.0
1
0 = m e /nii

FIG. 1. Plasma impedance versus conversion point position for various INTOR modes (a) for m < 0;(b) for m > 0.

648

KIROV et al.

(a)m = -1n = -10

400

,

200
0.03

0.0

0.5

1.0

0.0

FIG. 2. Plasma impedance versus conversion point position for m = —I, n = —10 mode at various
values (a) of plasma current, (b) of shear (y parameter).

(a)

(b)

0.0

FIG. 3. Radial distribution of electric field and absorbed powerm = —n = —2; (b) for m = —1, n = 3.

in R-O device:(a) forhave any substantial effect on the value of the absorption, which is due to small
gyrotropy in INTOR (fi/wci :£ 0.1). The effect of the plasma current on the plasma
impedance properties has been studied. The results shown- in Fig. 2(a) permit theconclusion that a current value of ~ 5 MA is preferable from the point of view ofabsorption level and its localization (here, y — 2/3).
As is seen from Fig. 2(b), a shear effect of the magnetic field only occurs inthat part of the plasma where high-Q kinetic resonances are observed. The parameter7 determines the temperature and the current (through Spitzer's conductivity law)
profiles (here, q(a) = 1.44).
The 'global' Alfve'n modes [4, 5, 10] are studied as a function of ll^al, and it
is shown that (1) below the Alfve'n continuum threshold, there may exist two andmore resonances of the 'global' mode and, (2) in certain conditions, the 'global'
mode may penetrate into the continuum for Ik^a] < 1 and leave it for ll^al > 2.5.

649

IAEA-CN-47/F-IV-3

Fields and absorbed power have been computed for the devices of the R-O type
= 2 X 10 13 crrr 3 ).
(a = 3.5 cm, R = 50 cm, Bo = 2.5 kG, T0(0) = 15 eV,
Figures 3(a,b) show the distributions of the electric fields and absorbed power
densities for the m = - 2 , n = 2 and m = — 1, n = 3 modes, respectively. The
conversion point was fixed at XA = 0.65. It is seen that, for the m = — 1, n = 3
mode, the slow Alfv6n wave contribution to the power absorbed at the plasmaboundary is small and the main portion of the power is absorbed in the central zone.
Calculations carried out for various conversion point positions and for frequencies
below the Alfve'n continuum threshold have shown that the wave amplitudes and thepower input at the periphery for the m = — 1, n = 3 mode are always smaller thanfor the m = - 2 , n = 2 mode.

4.

EXPERIMENTAL STUDIESThe R-O device [1] is an I = 3 stellarator with a quartz discharge chamber(R = 50 cm, b = 5 cm, a = 3.5-4 cm, Bo S 8 kG, t S 0.8). The helical andpoloidal antennas were fed by separate generators with frequencies chosen so as tosatisfy the Alfven resonance condition in radially close regions. For the helicalantenna with m = +2, n = +2, the frequency is f = 1.25 MHz, while for thepoloidal one with m = + 1 , n = + 3 , we have f = 2.2 MHz.

t (ms)

FIG. 4. Energy content (I), plasma density (2) and absorbed RF power due to helical (3) andpoloidal (4) antennas: {1,2 — only helical antenna switched on). H2, i0 = 0.4, Bo = 2.6 kG.

650

KIROV et al.

40<p> (kW)

FIG. 5. Temperature increase for helical (!) and poloidal (2) antennas
H2, ig = 0.4, Bo = 2.6 kG, <nf> = 7.5 x JO13 cm'3.

versus absorbed RF power.

r (cm)

FIG. 6. Radial distribution ofRF-field E component measured at instants of time indicated by arrows
in Fig. 4: 1 - f = 1.25 MHz, helical antenna; 2 - f = 2.15 MHz, poloidal antenna; H2, i0 = 0.4,
Bo = 2.6 kG.

The principal goal of the experiments was to study the dependence of therelative heating efficiencies and the impedances on m, n and on the types of antennas.
The discharge oscillograms are shown in Fig. 4. The special generator produced an
initial plasma with fio ~ (1-1.5) x 1013 cm" 3 and Te = 5 eV; then the helicalantenna was switched on and, later, the poloidal antenna with m = ± l , n = =F3
was turned on. From the plasma diamagnetism, density and absorbed power
measurements, the temperature increase (AT) and, hence, the heating efficiency(•q = (AT/P) n<)(1013)) are found. As is seen from Fig. 4, the heating efficiency ofthe poloidal antenna is many times higher than that of the helical antenna. This effectremained virtually unchanged over a wide range of absorbed powers (Fig. 5).

IAEA-CN-47/F-IV-3

651

Estimates of active impedances, Za = 2P/I 2 , for both types of antennas were made,and it was found that they depend weakly on the level of power input.
Figure 6 shows the distributions of the electric fields in the plasma (measured
by a double electrical probe), excited by helical and poloidal antennas. We see that,for the helical antenna, the electric field amplitudes at the plasma boundary (theseparatrix radius is r s = 3.5 cm) are much larger than in the region of r = 2.5 cm,
which was identified as a conversion zone. For the poloidal antenna, the field
amplitudes in these regions are comparable. These results indicate that the m = ± 1,
n = =F3 wave heats the central plasma region quite efficiently while the m = ± 2 ,n = =F 2 wave is absorbed mainly in the boundary region as a result of predominant
excitation of a peripheral slow Alfven wave, which accounts for the low efficiency
heating by the helical antenna, under the given conditions.
The analytic (Eq. (2)) and numerical (Fig. 3) computations show that, for them = - 1 , n = 3 mode, the field amplitudes are smaller at the plasma boundary thanin the plasma core, and RF power is absorbed mainly in the inner region. In contrast
to this, for the m = — 2, n = 2 mode, the field amplitudes at the boundary are muchlarger, and power is absorbed predominantly at the periphery.
Thus, the results of the experiments, the numerical and the analytic computations agree qualitatively. The present experiments cannot, however, set forth all
possible merits of the kinds of antennas in question since in our experiments thispossible effect is masked by the strong dependence of wave generation and absorption
on m and n.

5.

CONCLUSIONS

The analytic and numerical calculations and the experimental results presented
demonstrate, as a whole, that, for efficient plasma heating, modes with large
n/m ratios should be generated. For a wide class of devices (T-10, INTOR), this ratio
varies from 10 to 12. For INTOR, the optimum mode has m = - 1 , n = - 1 0 .
The experiments presented demonstrate a substantial increase in Alfven heating
efficiency, even if one proceeds from n/m = - 1 to n/m = - 3 .

REFERENCES
[1]
[2]
[3]
[4]
[5]
[6]

KIROV, A.G., et al., in Plasma Physics and Controlled Nuclear Fusion Research 1984 (Proc.
10th Int. Conf. London, 1984), Vol. 1, IAEA, Vienna (1985) 643.
KIROV, A.G., et al., in Controlled Fusion and Plasma Physics (Proc. 12th Europ. Conf.
Budapest, 1985), Vol. 2, Part 2, European Physical Society (1985) 260.
ROSS, D.W., LI, Y., MAHAJAN, S.M., MICHIE, R.B., Nucl. Fusion 26 (1986) 139.
APPERT, K., VACLAVIK, J., Plasma Phys. 25 (1983) 551.
APPERT, K., VACLAVIK, J., VILLARD, L., Phys. Fluids 27 (1984) 432.
DE CHAMBRIER, A., et al., in Plasma Physics and Controlled Nuclear Fusion Research 1984
(Proc. 10th Int. Conf. London, 1984), Vol. 1, IAEA, Vienna (1985) 531.

652
[7]

KIROV et al.

LOZOVSKY, S.N., DOROKHOV, V.V., Jr., Fiz. Plazmy 11 (1985) 539;
ELFIMOV, A.G., Fiz. Plazmy 11 (1985) 550;
ELFIMOV, A.G., in Heating in Toroidal Plasmas (Proc. 2nd Joint Varenna-Grenoble Int. Symp.
Como, 1980), Vol. 2, CEC, Brussels (1980) 683.
[8] ELFIMOV, A.G., LOZOVSKY, S.N., DOROKHOV, V.V., Jr., Nucl. Fusion 24 (1984) 609.
[9] SOLOVI'EV, L.S., in Voprosy Teorii Plazmy (LEONTOVICH, M.A., Ed.), Vol. 6, Moscow,
Atomizdat (1972).
[10] ROSS, D.W., CHEN, G.L., MAHAJAN, S.M., Phys. Fluids 25 (1982) 652.

IAEA-CN-47/F-IV-4

RECENT ADVANCES IN THE THEORYAND MODELLING OF RF HEATINGIN INHOMOGENEOUS PLASMAS
D.G. SWANSON, S. CHO
Department of Physics,
Auburn University,
Auburn, Alabama
C.K. PHILLIPS
Princeton Plasma Physics Laboratory,
Princeton University,
Princeton, New Jersey
D.Q. HWANG
Department of Applied Science,
University of California, Davis/Livermore,
Livermore, California
W. HOULBERG, L. HIVELYOak Ridge National Laboratory,
Oak Ridge, Tennessee
United States of America
Abstract
RECENT ADVANCES IN THE THEORY AND MODELLING OF RF HEATING IN
INHOMOGENEOUS PLASMAS.
Theoretical and numerical modelling of rf heating in inhomogeneous plasmas is discussed.
(1) A case is examined near the lower hybrid resonance where two turning points are encountered near
one another, first from a cold to a warm wave and then from the warm wave to an ion Bernstein mode.
This leads to a sixth order equation which is difficult to solve as it has a quadratic coefficient.
(2) Improved asymptotic representations are presented whose accuracy is crucial to the numerical solution of this type of equation. (3) A new method of exact order reduction for higher order equations with
no approximations permits generation of a numerical Green function by integrating only second orderequations. The differential equation is factored with an asymptotically small source term. The resulting
integral equation is more easily solved. (4) A formalism for determining the 2-D structure of the wave
fields and the associated power deposition profiles within the mode conversion-absorption layer has
been developed by applying the parabolic approximation method to the linearized Vlasov-Maxwell
equations. The new method reduces the problem to the solution of one fourth order equation and an
associated parabolic diffusion type equation. (5) Time dependent transport simulations of ICRF heateddischarges in tokamaks with circular and non-circular cross-sections are presented. The simulations are
obtained by interfacing the WHIST 2-D MHD equilibrium and a 1-D flux surface averaged transport
code with a ray tracing algorithm.

653

654
1.

SWANSON et al.

Three Wave Mode Conversion Equation

The dispersion relation for lower hybrid waves near ion cyclotron harmonics
is approximately given by[l]•"(1)where n is the harmonic number being considered and v — -^-.
- With themagnetic field= 5*i/»£,., f = |y = («/*i«*)2,?»
B(x) = Bo/(l + x/R) where R is the major radius (i> — n = nx/R), the dispersion
relation can be further approximated by the cubic equation in A or k\b4x)k\ + (a2 + b2x + cixi)k\- [a0 + box) = 0(2)

which describes the couplings between the incident lower hybrid wave, the ion
plasma wave, and the ion Bernstein wave (cold, warm, and hot wave, respectively). To obtain Eq. (2), we expanded v about n and A about Ao(= 2 \ l
the warm wave value. This approximation is compared to eq. (1) in Fig. 1.

FIG. 1. Comparison of dispersion relations (k\ versus v) of Eq. (1) (dark lines) with Eq. (2) (lightlines) near the 45th harmonic for typical PLT parameters.

This is converted into the differential equationW")"' +++

(3)

with suitable F,- because it has the conserved quantity(4)

IAEA-CN-47/F-IV-4

655

rather than replacing k± by » ^ , since although the simplified coefficients do notguarantee energy conservation, they do guarantee constant power flow asymptotically and enable one to represent the various amounts of power into various
waves.

2.

Numerical Method

Due to the quadratic term in Pi, which is essential to describe couplings
between all three waves, the Laplace transform method is no longer directly
integrable. Thus we use WKBJ approximations and numerical integration.
The asymptotic solutions are:
—oo

<— x

—>

+00

where a = [|(64/2) - y/{bt/2)' - c2\\ \ (3 = [[(64/2) + V(64/2)» - c2|J 82 , and
T = \/^o/c2! w ith 6oj ^4) «2 < 0. Taking the incoming and outgoing cold andwarm waves, and the exponentially decaying hot solution as initial conditions as
x —* +oo, we integrate five functions until x —* —00, then find the transfer matrix
by expanding those functions in terms of six asymptotic solutions. Tunneling andreflection coefficients for a specific branch may be found by using any basis ofasymptotic solutions, while conversion coefficients from one branch to anothercannot be found in this way, since the differential equation is homogeneous andsolutions may be multiplied by an arbitrary constant which must be determinedby a physical quantity we are interested in. Thus, to express conversion of energy
from the cold to the warm wave, we must multiply the energy flux ratio, pcw =S(ipw)IS{ipc)> associated with two waves in a given basis set, by the conversion
coefficient obtained numerically, where if),, and ^m a r e cold and warm asymptotic
solutions as x —* 00.
To improve the accuracy of tunneling, reflection, and conversion coefficients,
we include higher order terms such as

) [ (

5 §

)] (5)

Coefficients up to fourth order are found with the help of the symbolic algebra
routine REDUCE. A matrix formulation is being developed to find asymptotic
solutions more accurately and efficiently, since the overall accuracy of the coefficients is directly related to the accuracy of the asymptotic basis set.

656

3.

SWANSON et al.

Exact Order Reduction in Mode Conversion

Attempts have been made to eliminate the difficulties associated with a classof fourth order differential equations with exponentially growing solutions by
reducing the order of the equation[2,3]. While these methods have been successful in evaluating the transmission, reflection and conversion coefficients withapproximate equations and no localized absorption, the only successful methodsof including the cyclotron absorption are based on fourth order equations[l,4].
An exact method is described whereby only second order differential equations
need be solved numerically, avoiding approximations and growing solutions.
The problem is to solve the fourth order differential equation
yiv + b(x)y"+a{x)y

=O

(6)

where the generally complex coefficients a(x) and b(x) are real and asymptoticallyconstant plus linear. This equation is then factored by using the WKBJ roots ofthe equationd2,,,

where k%(x) = [b(x)±\/b2(x) - 4o(a;)]/2; after a variable change so that b -> X2z
and a —+ \2z+~t asymptotically, the roots are defined in terms of their asymptotic
behavior by the relations k\ w A - 1 and kl « 1 + (l+i)/\2z.
Then Eq. (7),written as D+D-y = g(x), is an exact representation of Eq. (6) with solution
V>(*)= f°° G(x,z)g(z)dz

(8)

J-oo

This is an inhomogeneous integral equation of the second kind, since 5 (z) contains
the unknown 4>[z) and its derivative, and the factors in Eq. (7) have been carefullyarranged so that the inhomogeneous term vanishes asymptotically at least as fastas z~2. Now, G(x,z) is a fourth order Green function which must be obtained
numerically, but it is constructed from solutions of second order equations only.
The first two solutions of the adjoint equation D-D+G = 0 are the numericalsolutions of D + * = <J>" + fc+(«)* = 0. The other two solutions are constructedfrom the numerical solutions of £>_$ = 0 and * i and *2 since D+G = 4>, so *3
and 4*4 are obtained from a second order Green function constructed from thefirst two solutions. Then a fourth order Green function can be constructed tosolve Eq. (7). The integral equation has a well behaved kernel and converges by
iteration.
4.

2D Mode Conversion - Parabolic Approximation

During fast magnetosonic wave heating of tokamaks, mode conversion andabsorption processes are generally important only in a thin layer located about

IAEA-CN-47/F-IV-4657

the cyclotron resonance layer. Within this layer, the parabolic approximation
method[5] has been used to determine the two-dimensional structure of the propagating finite size wavefronts. The layer may be modeled as a uniform plasma slab
in which x, y,z denote the major radial, vertical, and toroidal directions and theequilibrium magnetic field varies as B = Bo(l—x/R), where R is the major radius.
The mode conversion-tunneling equation is derived from the linearized VlasovMaxwell equations by following procedures used by Swanson[6] but explicitly
retaining y-derivatives of perturbed quantities. In the limit me/m,- —+ 0 and neglecting terms of higher order than (kv/kx)2 and (kxpi)2, wherefc,-denotes the ith
component of the wave vector and pi is the ion Larmor radius, the wave equation
reduces to two coupled pde's for Ex and Ev, where E denotes the wave electricfield. Using the paraxial propagation constraint that kv & -g^B < < kx & g^f,
the wave equation is solved iteratively, yielding an algebraic equation for Ex interms of Ev and a fourth order pde for Ev. The paraxial propagation constraint
also implies that Ev may be written as Ev(x,y) = a(x, y)u(x) , where u(i) is a
rapidly varying waveform and a(x, y) is a slowly varying amplitude factor containing the diffraction effects. Using this ansatz, u(z) satisfies the usual fourthorder mode conversion-tunneling equation for normal incidence[6], while a(x,y)
is determined by a parabolic diffusion type equation:

||

| 0

o/.

(9)

Eq. (9) is easily solved analytically using the separation of variables. Hence,
the two-dimensional structure of the wave fields within the kinetic layer canbe obtained by solving one fourth order equation and one parabolic diffusion
type equation. Computationally, this is more efficient than performing a Fourier
decomposition of the fields in y and solving a fourth order equation for eachresulting value of kv.

5.

Transport Simulations of ICRF-heated Tokamaks.

Global time-dependent transport simulations of ICRF-heated tokamaks withcircular and noncircular plasma cross-sections have been obtained by interfacing
the WHIST 2-D MHD equilibrium/1-D flux surface averaged transport code[7]with an ICRF wave propagation and power deposition package. The ICRF
package uses ray tracing/quasilinear damping algorithms, developed initially by
Hwang et. al.[8], but generalized to treat 2-D tokamak equilibria. Simulations of
PLT experiments[9] and the proposed CIT operating scenarioflO] are displayedin Fig. 2. For PLT, the electron thermal conductivity, \o ls based on L-mode
scaling[ll] while the ion thermal conductivity, Xi> consists of the neoclassicalcoefficient plus an anomalous contribution equal to OAxe- Comparison of thedata with the simulation indicates that the experimental energy confinement ismore favorable than L-mode. Since CIT is designed to operate with a divertor,

658

SWANSON et al.
1

(a)

(b)

CIT
60

j . sim>-1 /' rsim

40 -

'e,o

:
f
4

—
20 _

I

RFON RFOFF
t I < I500
TIME(msec)

|

RFON
n

-

I

1000

r

RFOFF
1

10
12
TIME(sec)

FIG. 2. (a) L-mode simulation of PUT ICRF experiments. T"m0 and T"m0 are the central electron andion temperatures from the code results, while T'fg and V*p0 are the experimental measurements
reported by Mazzucato et al. [9]. hi*'"indicates the extent of the experimentally observed sawtooth
amplitude on Te 0.
(b) H-mode simulation of CIT plasmas [10]. Ignition is achieved in the simulation with a 2 s pulse of
20 MW of ICRF heating at a frequency of 105 MHz in a D-T-3He plasma.

H-mode scaling was assumed for Xt while Xi w a s the same as for PLT. The simulation indicates that ignition should be possible in CIT with the current design
parameters.
Acknowledgment
This work was supported by the U. S. Department of Energy, Contracts DEAC02-76-CHO-3073 and DE-AC05-84OR21400,and by Grant DE-FG05-ER53206.
REFERENCES[1] SWANSON, D.G., Phys. Fluids 28 (1985) 2645.
[2] FUCHS, V., KO, K., BERS, A., Phys. Flulids 24 (1981) 1251.
[3] CAIRNS, R.A., LASHMORE-DAVIES, C.N., Phys. Fluids 26 (1983) 1268.
[4] COLESTOCK, P.L., KASHUBA, R.J., Nucl. Fusion 23 (1983) 763.
[5] PHILLIPS, C.K., PERKINS, F.W., HWANG, D.Q., Phys. Fluids 29 (1986) 1608.
[6] SWANSON, D.G., Phys. Fluids 24 (1981) 2035.
[7] HOULBERG, W.A., ATTENBERGER, S.E., LAO, L., in Advances in Mathematical Methods
for the Solution of Nuclear Engineering Problems (Proc. Int. Top. Mtg. Karlsruhe, 1981),
Vol. 2, Fachinformationszentrum Energie, Physik, Mathematik GmbH, Karlsruhe (1981) 425.
[8] HWANG, D.Q., et al., Rep. PPPL-1990, Princeton Plasma Physics Lab., Princeton University,
Princeton, NJ (1983).
[9] MAZZUCATO, E., BELL, R., BITTER, M., et al., in Plasma Physics and Controlled Nuclear
Fusion Research 1984 (Proc. 10th Int. Conf. London, 1984), Vol. 1, IAEA, Vienna (1985) 433.
[10] SCHMIDT, } . , et al., paper IAEA-CN^7/H-I-2, these Proceedings, Vol. 3.
[11] KAYE, S.M., GOLDSTON, R.J., Nucl. Fusion 25 (1985) 65.

IAEA-CN-47/F-IV-5

ELECTRON CYCLOTRON CURRENT DRIVE
AND SUPRA-THERMAL GENERATIONIN LARGE TOKAMAKS
T.J. SCHEP, R.W. POLMAN, E. WESTERHOF, L. KLIEBAssociation Euratom-FOM,
FOM Instituut voor Plasmafysica,
Rijnhuizen, Nieuwegein,
Netherlands

Abstract
ELECTRON CYCLOTRON CURRENT DRIVE AND SUPRA-THERMAL GENERATION IN
LARGE TOKAMAKS.
Top launch of electron cyclotron wave energy into large tokamaks is considered with the aim ofinvestigating the possibilities of spatially localized power deposition in the far tail of the electron distribution function. Such an injection scheme is favourable from the point of view of current drive andgeneration of a supra-thermal electron population. Oblique injection of the extraordinary mode with afrequency below the local cyclotron frequency is most suitable for this purpose.

INTRODUCTION
Recent experiments with strong additional heating on JET [1]and TFTR [2] give evidence of a strong correlation between thecurrent and temperature profiles in tokamak discharges. The decoupling of these profiles could clear the way to increase theenergy content of the plasma and to improve its confinement.
This could be achieved by driving part of the current noninductively or by carrying the current by a supra-thermal electron population. The noninductive current can be provided byvarious methods. The shape of the current profile, however,might well be the important factor which determines eventually
the efficiency, in terms of plasma energy content and confinement time, of the non-inductive current generation [3]. The favourable feature of electron cyclotron power deposition andcurrent drive is its localized character. This could provide
the means for the necessary tailoring of the current profile,even with modest EC-driven currents.

659

660

SCHEP et al.

In this paper, power deposition, current drive, and generation of supra-thermal electrons by electron cyclotron waves are
discussed for JET and TFTR plasmas. In p a r t i c u l a r , top launchof RF power in the extraordinary mode i s investigated.

ELECTRON CYCLOTRON RESONANCE AND RAY TRACING
The r e l a t i v i s t i c electron cyclotron resonance condition is
(0

y

2 2
Y- —
+ p,|/m2c2]2
u - NM pn/mc = 0, where Y = [1 + p f / m c

II II

1

II

and N,| is the parallel refractive index. All other quantities
have their usual meaning. In the Pn"^ plane the resonancecondition is represented by the straight line Y-1 = A+NH PM /me,where A = <o /to-1, inside the hyperbola Y = (1+p,|/m2c2)^. The
angle between the straight line and the pn-axis is determinedby Nil. The ranges of energies E = mc2(Y-1) and parallel momentaof trie resonant particles are determined by the intersection
points of these curves. The resonance condition implies thatelectrons with rather large energies can be resonant if waves
with down-shifted frequencies (w<u ) are injected [JJ]. Undercthe realistic assumption that the electron distribution function is a decreasing function of energy, most of the RF energy
will be deposited near the lower limit (E~) of the energy range.
The ray trajectories and the absorption of electron cyclotron waves injected from a top-port into JET and TFTR are
studied with the ray tracing code TORAY [5], which has been
extended to calculate ray paths in D-shaped equilibria. The
trajectories are calculated from the cold dispersion relation[6]. The power deposition is determined by the absorption coefficient, calculated with the full relativistic cyclotron resonance condition. A contribution due to a supra-thermal electron
population is included in the absorption. This population has
been modelled by a low density, high energy distribution function which is Maxwellian in a frame moving along the equilibrium magnetic field [7].
The magnetic equilibria have been modelled to present actual
discharges in JET and TFTR. In these equilibria the density andIAEA-CN-47/F-IV-5

661

~~

be

temperature profile are given by ne(^) = (no-riL,)[1 - a n ifn ] n +n ,

T(*)
6

being

= (T -T )[1-a * b T ] c T + T ,
U

L

i

l

the poloidal

where ^ 2 = ( ¥ - * „ ) / | 1>,-*J , *

LI

flux

functionOand \>|

LIU

i t s value atthe

Lt t O

limiter and magnetic axis respectively, with suitable choices
for the coefficients n . and T . .

o,Lo,LPOWER DEPOSITION IN REAL AND MOMENTUM SPACE
Power deposition in JET along rays that are injected under avariety of angles from several positions into medium (n5*10 19 m" 3 ) and high (n = 1 0 2 0 m" 3 ) density Maxwellian plasmas
with central temperatures of 5 and 10 keV has been extensively
investigated for various equilibria, density profiles, and frequencies [8], With the proper choice of frequencies the results
for JET and TFTR are analogous for standard discharges. In thecase of TFTR we have investigated super-shot 23260 because ofits broad T -profile and its large shift of the magnetic axis
6[9]. The special features of this shot lead to interesting results even at the relatively low frequency of 90 GHz.
The direction of the injected rays is specified by theangles 9 and <j>, where 9 is the angle between the ray and thevertical direction, and <>f is the angle between the projectionof the ray on the equatorial plane and the major radius. The
dependence of the ray trajectories and of the power depositionon the injection angles is analyzed by scanning in <>f at fixed 9and vice versa. With a proper choice of frequency and launchingangles, complete power absorption with good spatial localization can be obtained in both medium and high density plasmas
with central temperatures in the range 5-10 keV. Figures 1 and2 show typical results of a e-scan at <>j = -80° in JET and TFTR
for a Maxwellian plasma. The projections of the ray paths on
the poloidal plane are depicted in Figs. 1a and 2a. In Figs. 1band 2b the bands, in terms of ty and E~, within which the intensity drops from 90% to "\0% of the injected value, are plottedversus 9. Similar features as in a 9-scan are shown by <(>-scans.
The study yields the following results [8]. The density has
only a weak influence on the position where the wave power isabsorbed up to densities close to the cut-off value. Only minorchanges in the results are found for density profiles withstrongly different peaking factors. For higher temperatures the

SCHEP et al.

662

BO

1.25
Cm)

0 = -80°

CkeV)

1.00

z

position
0.75
0.50
0.25

0 -1.0

0.0

1.0

0.00
120 130 140 150 160 170

Cm)
FIG. 1. Results of a scan in 6 (at <)> = -80°) in JET (70 GHz). The rays are launched fromR = 3.0 m, Z = 1.95 m). The equilibrium corresponds to JET-3901 with n0 = 5 x lO19 m~3,
To = 5 keV (&„, T = 1, bn, r = 2, cn_ T = 1) and B = 3.4 Tat R = 2.96 m. 6 = 125° (1), 135° (2),145° (3), 155° (4), 165° (5), 175° (6).

CkeV)
E"

FIG. 2. Results of a scan in 6 (at <t> = -80°) in TFTR (90 GHz). The rays are launched fromR = 2.55 m, Z = 1.00 m. Equilibrium and profiles as in TFTR-23260: R = 2.46 m, a = 0.80 m,
B = 5T,no=7 x lO19 m~3 (nL= 7 x 1OIS m'3, an = 1, bn = 1.5, cn = 4) and To = 7 keV(TL = 30 eV, aT =l.bT= 2, cT = 1). d = 155° (1), 145° (2), 135° (3), 125° (4), 115° (5).

deposition band narrows somewhat and shifts to higher values of
to /io and rays directed more towards the centre and towards theinboard side show increased absorption. Results for power injection from several positions in a top-port show that rays canbe injected more steeply and more towards the high-field side
when injected from positions with larger major radii. With a

663

IAEA-CN-47/F-IV-5
1.0

1position

• °-a

'

i

1

• = -80°

i

200

i

CkeV)

9= 155°, energy

- 160

//

1:

> \

0.6 -

120

L

0.4 -

-80

" position^^_ energy
0.2 -

-40
(a)

0.0
125

i

i

1

i

135

145

155

165
*•

8

Cb)
t

i

1

1

175-120-110-100 -90 -80 -70 -60*•«

FIG. 3. Ray tracing results in the presence of a supra-thermal population. The dashed bands correspond to a purely Maxwellian plasma, a) a scan in 6 at <t> = —80°, b) a scan in <t> at 9 = 155°.

proper choice of the injection angles for each launching position, the power can be deposited around the same flux surface
and be absorbed by particles in the same energy range. For
larger values of to /u (i.e., lower frequency or higher B-values) the absorption decreases. Also in paramagnetic equilibria(i.e. higher B) the absorption will be lower. Typical values
for the resonant energies are given in the Table for nearcentre deposition in JET-3901. Similar results are found for
TFTR.
In the presence of a supra-thermal electron population the absorption behaviour changes strongly. This is illustrated inFig. 3 for a paramagnetic equilibrium (the low-q JET shot 777*1,
rescaled to B = 3-43 T at R = 2.96 m ) . For a 10 keV, n o =5*10 l 9 m" 3 Maxwellian plasma full absorption is obtained up to
6 = 155° (at <(> = -80°) and for <j> > -90° (at 8 = 155°) with particle
energies around 50 keV. We assume that the power in these rays
produces a supra-thermal population with n s = 5*10 l7 m~ 3 , T =50 keV and < v,, > = 0.331 c inside the surface ty = 0.6 (thisvalue of < Vii > is not realistic from the point of view ofsupra-thermal current, but, here, we consider power absorption). Then, the range in 9 and <Ji for complete absorption is
drastically extended, and the RF power in steeply injected rays
is absorbed by particles with energies up to 200 keV in the
tail of the supra-thermal distribution.

664

SCHEP et al.

CURRENT DRIVE EFFICIENCY
The non-inductive current I, generated by a point spectrum ofelectron cyclotron waves, is discussed in Ref. [10]. According
to this theory, the current that can be carried by relativistlc
particles with energy mc z (Y-1) and parallel momentum P|| isJ

2TTR P .

d

R(m)n2Pu

A/Wn

where R is the major radius, n 2 0 the density in 10 2 0 m"3, W the RF
power and (j/P ) the normalized efficiency [10]. For the energies under consideration we have I/W = [6/(5+Z „„)] (I/W)z f.P_i •61 X

61 X — I

From Table I and Eq. (1), the following rough estimate is obtained for the efficiency of current drive with near-central
power deposition
T (10" eV)o= a -^j—,
A/W
(2)
n 20
R(m) nwhere a = 0.15-0.30 for the temperatures and densities underconsideration. Here, we have assumed that all the wave power isabsorbed by resonant particles with the lowest energy E~.
The profiles of power deposition and current drive for amodelled launching system are given in Fig. 4. In Fig.4a a magnetic equilibrium similar to JET-7771* has been investigatedwith n =0.5'102° m~ 3 and T =5 keV. The toroidal field has beenoorescaled to B = 3.2 T at R = 2.96 m, which optimizes the absorption around the q=1 surface (r = 45 cm). In Fig. 4b theequilibrium and plasma parameters are chosen to represent TFTR23260. The large shift of the magnetic axis in this particulardischarge does allow for central power deposition and current
drive (cf. Fig. 2). The calculations show that j(r) = P(r) whichyields an overall efficiency I/W £ 0.05 A/W in agreement with (2).

TABLE I. E " FOR NEAR CENTRE DEPOSITION
T0(keV)

noOn "')

E"(keV)T0(keV)no(mE"(keV)5

0.5 x 1020

25-35

10

0.5 x 1020

60-80

5

20

10

1 X 1020

35-55

1 X

10

10-25

665

IAEA-CN-47/F-IV-5

I

0.0

0.00

0.0 0.4 0.8 1.2
>• r

Cm)

F/G. 4. a) Power deposition and current profile in JET 7774 (70 GHz). Two Gaussian beams (halfwidth 5°) of 0.5 MW are launched at R = 2.94 and 3.06. The injection angles of the beam centres are
6 = 150°, <S> = -80° and 6 = 155°, <j> = -80°. b) Power deposition and current profile in TFTR
23260 (90 GHz). 1 MWis launched in a Gaussian beam (half-width 5°) from the top of the TFTR vessel(R = 2.55). The injection angles of the beam centre are 0 = 145°, 4> = -80°.

The widths of the profiles are mainly due to the assumed angular spread of 5° of each beam, and are to a lesser extend theresult of the different absorption lengths along the indiviualrays. Hence, the localization of the power deposition and of
the driven current depends to a large extend on the qualities
of an actual launching system.

SUPRA-THERMAL ELECTRON GENERATION
The power deposition and the current drive efficiency are calculated under the assumption that the RF power is absorbed byhigh-energetic, resonant particles without distorting the tailof the Maxwellian. A necessary condition for the validity ofthese results is, thus, that collisions are sufficiently frequent so that the RF power can be transferred to the bulk of
the distribution function by collisional energy loss of resonant particles. When this condition is not fulfilled, resonantparticles will run away in momentum space and strong non-Maxwellian effects are expected [11].
The power lost per unit volume by high energetic particles
with density n
due to collisions with bulk particles ispcoll - n spe

In A M i m v

f 2 v c (v) = 6.10 3 a s n=
3

MW-nT

where v=is the collision frequency (X,nA = 15 has been

666

SCHEP et al.

taken), and a is the fraction of resonant particles. Estimats
ing a for a Maxwellian plasma we arrive at
s

-E/T
p

coii

=

4 103

°- '

0011

i
1

(AE)
n

y—~
(E/T/22T» MW<m"3(3)2T

where (AE)^ is the thermal spread in energy. At the higher
temperatures, the spread in energies E~ is of the order of thethermal spread, AE~ = (AE) = 2T, E/T = 6-9 at n 20 = 0.5 andE/T = ^-6 at n 20 = 1 (see also Table I),The volume in which the
RF power is absorbed is of the order of 10 m3 when the launching is optimized for near-centre depositions, and about 2 m3
when optimized for central deposition. A substitution of thesevalues into (3) shows that with a few MW of RF power, suprathermal effects can be expected at low densities for nearcentre deposition, but also at higher densities for central deposition.
In the run-away regime a depletion of the distribution function will occur unless particles are continuously supplied by
the bulk at the lower energy boundary of the resonant population. Resonant particles will accumulate at the high energy
boundary and be redistributed by pitch-angle scattering. A
non-Maxwellian distribution function will be established in theresonant region. This necessitates a reconsideration of theabsorption coefficient. Also the deposition profile and therange of particle energies where the power i s absorbed willchange drastically, as is i l l u s t r a t e d by the results of Fig. 3.
CONCLUSIONS
The r e s u l t s presented in the preceding Sections lead to theconclusion that electron cyclotron waves at downshifted frequency are completely absorbed in a large tokamak for a wide
range of plasma parameters. The RF power is transferred toelectrons in the t a i l of the distribution function. The power
deposition is spatially localized for a broad range of injection angles and positions, and plasma parameters.
For real launching systems the localization will be limitedby the angular spread of the beams. I t i s shown by Fig. ^ thatall power can be deposited in a volume of - 2 m3 for central andin a volume of - 10 m3 for near-centre deposition ( e . g . , aroundIAEA-CN-47/F-IV-5
the q=1 surface). The overall current drive efficiency
dicted to be >0.O5 A/W in both cases.

667
i s pre-Supra-thermal e f f e c t s are expected for a few MW of RF power.
The creation of suprathermals will d r a s t i c a l l y a l t e r the absorption by the plasma, and will allow the interaction withhigher energetic p a r t i c l e s in the t a i l of the supra-thermald i s t r i b u t i o n , leading to an increased current drive efficiency.
Thus EC wave injection at down-shifted frequencies from a
top-port in a large tokamak (e.g. 70 GHz for B = 3 . 4 T (JET) or
90 GHz for B = 5 T (TFTR)) is an a t t r a c t i v e tool for current
p r o f i l e control.

ACKNOWLEDGEMENTS
Fruitful discussions with Prof.Dr. F. Engelmann are gratefully
acknowledged. This work was performed under the Euratom-FOM
association agreement with financial support from ZWO andEuratom and in part under JET Article 14 Contract no. JJ5/9007.

REFERENCES[1] REBUT, P.H., BRUSATI, M., Plasma Phys. Controll. Fusion 28 (1986) 113.
[2] MURAKAMI, M., et al., Plasma Phys. Controll. Fusion 28 (1986) 17.
[3] FURTH, H.P., Plasma Phys. Controll. Fusion 28 (1986) 1305.
[4] MAZZUCATO, E., FIDONE, I., GIRUZZI, G., KRIVENSKI, V., Nucl. Fusion 26 (1986) 3.
[5] KRITZ, A.H., HSUAN, H., GOLDFINGER, R.C., BATCHELOR, D.B., Ray-Tracing Studyof Electron-Cyclotron Heating in Toroidal Geometry, Rep. PPPL-1980, Princeton Plasma
Physics Laboratory (1983).
[6] BORNATICI, B., CANO, R., De BARBIERI, O., ENGELMANN, F., Nucl. Fusion 23 (1983)
1153.
[7] BORNATICI, M., RUFFINA, U., WESTERHOF, E., Plasma Phys. Controll. Fusion 28 (1986)
629.
[8] POLMAN, R.W., SCHEP, T.J., WESTERHOF, E., KLIEB, L., in Report on JET, Article 14,
Contract No. JJ5/9007.
[9] GREK, B., JOHNSON, D., private communication.
[10] FISCH, N.J., Phys. Rev., A 24 6 (1981) 3245.
[11] DEMEIO, L., ENGELMANN, F., in Mathematical Aspects of Fluids and Plasma Dynamics(Proc. Workshop Trieste, 1984) (1984) 201.

CHAIRMEN OF SESSIONS
Volume 1
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USA

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USA

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