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 3

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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 3

INTERNATIONAL ATOMIC ENERGY AGENCY
VIENNA, 1987

PLASMA PHYSICS AND CONTROLLED NUCLEAR FUSION RESEARCH 1986
IAEA, VIENNA, 1987
STI/PUB/723
ISBN 92-0-130287-8

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 3
INERTIAL CONFINEMENT (Session B)
Laser programme development at Centre d'e'tudes de Limeil-Valenton(IAEA-CN-47/B-M)
3
A. Andre, E. Berthier, E. Buresi, A. Coudeville, J. Coutant,
R. Dautray, M. Decroisette, C. Delmare, B. Duborgel, P. Guillaneux,
J. Launspach, P. Nelson, C. Patou, J.-M. Reisse, D. Schirmann,
B. Sitt, J.-P. Watteau
Discussion
13
Progress of inertial confinement fusion at Lawrence Livermore
National Laboratory (IAEA-CN-47/B-I-2)
15
E. Storm
Discussion
24
X-ray confinement in a laser heated cavity (IAEA-CN-47/B-I-3)
25
T. Mochizuki, T. Yabe, H. Azechi, K. Tanaka, T. Boehly,
N. Miyanaga, H. Nishimura, S. Ido, M. Yamanaka, K. Nishihara,
T. Norimatsu, T. Jitsuno, M. Nakatsuka, K. Mima, S. Nakai,
C. Yamanaka, R. Sigel, G.D. Tsakiris, K. Eidmann, P. Herrmann,
R. Pakula, P. Sachsenmaier, S. Sakabe, S. Witkowski
Discussion
32
Inertial confinement fusion research by GEKKO lasers at ILE Osakaand target design for ignition (IAEA-CN-47/B-I-4)
33
C. Yamanaka, K. Mima, S. Nakai, T. Yamanaka, Y. Izawa, Y. Kato,
K. Nishihara, T. Mochizuki, M. Yamanaka, M. Nakatsuka, T. Sasaki,
T. Yabe, K. Yoshida, H. Azechi, H. Nishimura, T. Norimatsu,
S. Ido, M. Miyanaga, H. Niki, E. Fujiwara, H. Shiraga, M. Nakai,
H. Takabe, T. Jitsuno, K. Tanaka, M. Takagi, A. Nishiguchi
Direct-drive laser fusion in the United States of America(IAEA-CN-47/B-I-5)
47
R.L. McCrory, J.M. Soures, P. Audebert, O. Barnouin, R.S. Craxton,
J. Delettrez, R. Epstein, L. Forsley, R.J. Hutchison, P. Jaanimagi,
S. Jacobs, R.L. Keck, T. Kessler, H. Kim, R. Kremens, S.A. Letzring,
R.S. Marjoribanks, F. Marshall, P. McKenty, G.A. Mourou,
M. C. Richardson, W. Seka, R. W. Short, A. Simon, S. Skupsky,
C. Verdon, J. Wark, B. Yaakobi, S.E. Bodner, J.P. Dahlburg, M. Emery,
J. Gardner, J. Grun, T. Lee, C. Manka, E. McLean, S. Obenschain,
B. Ripin, J. Stamper

Progress in inertial confinement in the United Kingdom(IAEA-CN-47/B-II-l)
55
M.G. Haines, D. Bassett, A.R. Bell, E.M. Epperlein, T.H. Kho, D. Riley,
S.D. Tabatabaei, O. Willi, J. Corbett, M.J. Lamb, C.L.S. Lewis,
E.W. Robertson, P.M. Evans, A. P. Fews, C. Chenais-Popovics,
M.J. Henshaw, G.J. Pert, P. Apte, A.J. Cole, S.J. Rose, W.T. Toner
Discussion
67
Inertial confinement fusion program at Los Alamos National Laboratory(IAEA-CN-47/B-H-2)
69
D.C. Cartwright, P.D. Goldstone, J.A. Hanlon, A. Hauer, R. Mah,
J. McLeod, W.C. Mead, LA. Rosocha, D.C. Wilson, C.A. Wingate
Discussion
77
Experimental studies on inertial fusion at 0.26 ^m laser wavelength(IAEA-CN-47/B-II-3)
79
F. Amiranoff, R. Fabbro, E. Fabre, B. Faral, C. Labaune, A. Michard,
J. Virmont, P. Mora, J.F. Luciani, R. Pellat, D. Pesme, J. Briand,F. Cottet, M. Hallouin, J.P. Romain, H. Baldis, A. P. Fews
Discussion
91
Progress in inertial confinement fusion with light ion beams at
Sandia National Laboratories (IAEA-CN-47/B-II-4)
93
J.P. VanDevender
Discussion
101
Inertial confinement fusion research with light ion beams at ILE(IAEA-CN-47/B-II-5)
103
K. Imasaki, S. Miyamoto, S. Nakai, K. Nishihara, H. Takabe,
C. Yamanaka
X-ray diagnostic developments for laser-matter interaction experiments(IAEA-CN-47/B-III-l)
113
J.L. Bourgade, P. Combis, J. Launspach, J. de Mascureau, A. Mens,
J.M. Mexmain, A. Saleres, R. Sauneuf
Experimental studies of radiation driven laser fusion(IAEA-CN-47/B-III-2)
123
H.Z. Shen, H.S. Peng, Z.J. Zheng, J.G. Yang, S.J. Wang
Research on heavy ion drivers for inertial confinement fusion in the
United States of America (IAEA-CN-47/B-III-3)
129
C. Celata, A. Faltens, T.J. Fessenden, D.L. Judd, D. Keefe, C.H. Kim,
L.J. Laslett, E.P. Lee, L. Smith, M.G. Tiefenback, A.I. Warwick
Microwave simulation of laser-plasma interactions (IAEA-CN-47/B-III-4) ... 137
C.J. Pawley, N.C. Luhmann, Jr., W.L. Kruer
Progress in inertial confinement research at KMS Fusion, Inc.
(IAEA-CN-47/B-III-5)
145
J.T. Larsen

Progress in inertial confinement fusion physics and computational
capabilities at DENIM (IAEA-CN-47/B-III-6)
G. Velarde, J.M. Aragones, J.A. Gago, L. Gdmez, M.C. Gonzalez,
J.J. Honrubia, J.M. Martinez-Val, E. Minguez, J-L. Ocana, R. Otero,
J.M. Perlado, J.M. Santolaya, J. Sanz, J.F. Serrano, P.M. Velarde
Recent experimental results of cryogenic pulsed ion diodes(IAEA-CN-47/B-III-7)
K. Kasuya, K. Horioka, H. Yoneda, K. Mitobe, S. Kato, M. Harada,
S. Morimoto, M. Morioka, T. Obama, T. Taira, K. Nakamura
Experimental study of uniform acceleration with UV laser light(IAEA-CN-47/B-III-8)
K. Koyama, I. Matsushima, Y. Owadano, M. Tanimoto, T. Kasai,
A. Yaoita, I. Okuda, Y. Matsumoto, F. Nemoto, T. Tomie,
S. Komeiji, M. Yano
Inertial confinement fusion research with intense pulsed light ion beams(IAEA-CN^7/B-III-9)
K. Yatsui, Y. Shimotori, Y. Araki, K. Masugata, S. Kawata,
M. Murayama
Radiation transport in laser irradiated targets (IAEA-CN-47/B-IH-10)
A. Ng, Y. Gazit, F.P. Adams, D. Parfeniuk, P. Celliers, L Da Silva

153

161

167

177

187

INTOR (Session G)
INTOR: Introductory remarks (IAEA-CN-47/G-I-1)
S. Mori
INTOR: Overview of the INTOR Workshop (IAEA-CN-47/G-I-2)
W.M. Stacey, Jr.
Discussion on Papers IAEA-CN-47/G-H and G-I-2
INTOR: Impurity and particle control (IAEA-CN-47/G-I-3)
D.E. Post
Discussion
INTOR: Operational limits and confinement (IAEA-CN-47/G-I-4)
F. Engelmann
INTOR: Heating and current drive (IAEA-CN-47/G-I-5)

195
199
211
213
221
223
231

iV. Fujisawa

Discussion

245

TECHNOLOGY AND REACTOR CONCEPTS (Session H)
Concept and parameters of NET (IAEA-CN-47/H-M)
NET Team
Discussion

249
256

A compact ignition experiment (IAEA-CN-47/H-I-2)
259
J. Schmidt, G. Bateman, C. Bushnell, J. Citrolo, P. Colestock,
D. Ignat, R. Izzo, S. Jardin, L.P. Ku, G. Kuo-Petravic, M. Petravic,
C.K. Phillips, N. Pomphrey, D.E. Post, R.B. White, K.M. Young,
D. Blackfield, T. Brown, C. Flanagan, Y.-K.M. Peng, T. Shannon,
D. Strickler, L. Bromberg, D. Cohn, M. Greenwald, B. Lipschultz,
D.B. Montgomery, R. Parker, D. Sigmar, R. Thome, S. Wolfe,
W. Houlberg, J. Sheffield, D. Swain, N. Uckan, M. Hughes, C. Paulson,
M. Phillips, A.M.M. Todd, G. Kerbel, D.G. Swanson, F. Puhn
Discussion
265
Tokamak reactor operation scenario based on plasma heating and current
drive by negative ion based neutral beam injector (IAEA-CN-47/H-I-3) ... 267
5. Yamamoto, K. Okano, S. Nishio, M. Sugihara, R. Saito, T. Uede,
T. Kobayashi, N. Fujisawa, T. Tone, Y. Ohara, S. Tanaka,
H. Horiike, Y. Okumura, T. Shibata, S. Matsuda, K. Tani,
T. Hirayama, M. Azumi
Discussion
276International Large Coil Task: Testing of the largest superconducting
toroidal magnet system (IAEA-CN-47/H-I-4)
279
M.S. Lubell, W.A. Fietz, P.N. Haubenreich, J.W. hue, J.N. Luton,
S.S. Shen, K. Okuno, A.R. Ulbricht, Z.A. Zichy
Discussion
291
Radiological implications of low activation steels as structural materials
in fusion reactors (IAEA-CN-47/H-I-5)
293
R. Hancox, B. Mitchell, G.J. Butterworth
Discussion
300
A muon catalysed fusion-fission reactor (IAEA-CN-47/H-II-l)
301
S. Eliezer, T. Tajima, M.N. Rosenbluth
Discussion
307
Effect of anomalous transport on parameters of a compact torus based
quasi-stationary reactor (IAEA-CN-47/H-II-2)
309
V.A. Burtsev, V.M. Kozhevin, A.N. Makhankov
Initial operation of the high field superconducting tokamak TRIAM-1M(IAEA-CN-47/H-II-3)
321
5. Itoh, Y. Nakamura, N. Hiraki, K. Nakamura, A. Nagao,
S. Moriyama, S. Kawasaki, E. Jotaki
Discussion
331
MFTF-B acceptance tests and operation (IAEA-CN-47/H-III-2)
333
F.H. Coensgen, J.W. Gerich, P.M. Holl, V.P. Karpenko, T.A. Kozman,
K.H. Krause, W.C. Labiak, D.D. Lang, G.H. Lathrop, W.B. Lindquist,
R.L. Nelson, R.L. Peterson, D.L. Podesta, J.R. Reed, P.R. Rupert,
D.W. Shinier, P.D. Siemens, W.H. Sterbentz, R.R. Stone, J.S. Sullivan,
A.L. Throop, J.H. VanSant, J.P. Zbasnik, B. Huff, C.A. Leavitt

Advanced ICF reactor TAKANAWA-I with low radioactivity andtritium self-breeding in the pellet (IAEA-CN-47/H-III-4)
T. Tazima, S. Ido, T. Yabe, T. Shiba, H. Nakashima,
M. Nishikawa, N. Asami
Study of coatings as high heat flux materials (IAEA-CN-47/H-III-5)
M. Okada, M. Fukutond, T. Shikama, Y. Yamauchi
Development of superconducting materials and magnet technology
for the next generation of fusion machines in Japan(IAEA-CN-47/H-III-6)
S. Shimamoto, T. Ando, T. Hiyama, H. Tsuji, Y. Takahashi,
M. Nishi, E. Tada, K. Yoshida, K. Koizumi, K. Okuno, H. Nakajima,
T. Kato, K. Kawano, M. Hoshino, M. Oshikiri, H. Wada,
T. Kuroda, K. Inoue, K. Tachikawa
Engineering study of the feasibility of a fusion power reactor
employing the dense Z-pinch plasma confinement geometry(IAEA-CN-47/H-III-7)
H.R. Bolton, P. Choi, A.E. Dangor, M.G. Haines, M. Javadi,
S.J. Peerless, A. Power, A.E. Robson, S.P. Walker
Deuterium based fuel cycles: The benefits of tritium leanness(IAEA-CN-47/H-III-9)
W. Kernbichler, R. Feldbacher, M. Heindler, A. Nassri, K. Schdpf
Superhigh density operating conditions for a poloidal divertor
in a tokamak reactor (IAEA-CN-47/H-IIMO)
S.I. Krasheninnikov, A.Yu. Pigarov

343

351

357

367

373

387

FUNDAMENTAL PROCESSES AND NEW TRENDS (Session I)
Ignition physics processes and compact experiments(IAEA-CN-47/I-I-1)
397
B. Coppi, R. Englade, S. Migliuolo, F. Porcelli, L. Sugiyama
A new approach to nuclear fusion using a CO2 laser (IAEA-CN-47/I-I-2) .... 407
S. Nakai, Y. Kitagawa, H. Daido, H. Fujita, K. Sawai, M. Fujita,
K. Nishihara, K. Mima, A. Hasegawa, C. Yamanaka
Plasma stabilization in traps with spatial axis (IAEA-CN-47/I-I-3)
421
V.M. Glagolev, A.A. Subbotin, B.A. Trubnikov
Relaxation of plasma toroidal rotation in tokamaks and anomalous viscosity(IAEA-CN-47/I-I-4)
435
D.Kh. Morozov, O.P. Pogutse
Theory of tokamak profiles (IAEA-CN-47/I-I-5)
447
D. Biskamp, D. Pfirsch, F. Pohl

Pellet fuelling of JET plasmas during Ohmic, ICRF and NBI heating(IAEA-CN-47/I-I-6)
A. Gondhalekar, M. Bures, D. Campbell, A. Cheetham, S.A. Cohen,
J.G. Cordey, S. Corti, A.W. Edwards, J. Ehrenberg, R.D. Gill,
R. Granetz, F. Hendriks, W. Houlberg, H. Jacket, T. T. C. Jones,
M. Keilhacker, G. Magyar, P. Mantica, V. Mertens, P.O. Morgan,
P. Nielsen, J. O'Rourke, G. Sadler, F.C. Schuller, M.F. Stamp,
A. Tanga, P.R. Thomas, M.L. Watkins, A. Weller
First experiments on the gasdynamic trap (IAEA-CN-47/I-I-7)
P.A. Bagryanskij, A.A. Ivanov, V.V. Klesov, Yu.L. Koz'minykh,
I.A. Kotel'nikov, Yu.I. Krasnikov, A.A. Podyminogin, A.I. Rogozin,
G.V. Roslyakov, D.D. Ryutov

457

467

SUMMARIES (Session L)
Summary on tokamak experiments (IAEA-CN-47/L-I-1)
A. Gibson
Summary on alternative magnetic systems — experimental results(IAEA-CN-47/L-I-2)
T. Sekiguchi
Summary on magnetic confinement theory (IAEA-CN-47/L-I-3)
K. Lackner
Summary on inertial confinement fusion (IAEA-CN-47/L-I-4)
S.E. Bodner
Summary on technology and reactor concepts (IAEA-CN-47/L-I-5)
R. Hancox
Chairmen of Sessions
Secretariat of the Conference
List of Participants
Author Index

479

497
517
527
535

545
546
547
603

Session B
INERTIAL CONFINEMENT
Chairmen of Sessions
Session B-I

R.L. SCHRIEVER

USA

Session B-II

C. YAMANAKA

Japan

Session B-III (Posters)

IAEA-CN-47/B-I-1

LASER PROGRAMME DEVELOPMENT
AT CENTRE D'ETUDES DE LIMEIL-VALENTON
A. ANDRE, E. BERTHIER, E. BURESI, A. COUDEVILLE,
J. COUTANT, R. DAUTRAY, M. DECROISETTE,
C. DELMARE, B. DUBORGEL, P. GUILLANEUX,
J. LAUNSPACH, P. NELSON, C. PATOU, J.-M. REISSE,
D. SCHIRMANN, B. SITT, J.-P. WATTEAU
CEA, Centre d'e"tudes de Limeil-Valenton,
Villeneuve St. Georges,
FranceAbstract
LASER PROGRAMME DEVELOPMENT AT CENTRE D'ETUDES DE LIMEIL-VALENTON.
A significant effort has been made recently at CEL-V to improve the laser facilities.
Octal, the eight-beam 2 kJ laser, has been equipped with phosphate glass and KDP frequency
tripling systems. The new laser facility Phebus has been built; its two beams deliver 20 kJ of
1.05 /im light which is easily converted to 0.53 jum or 0.35 jum wavelength. New diagnostics
have also been developed and tested. Laser driven hydrodynamics has been studied in various
configurations; there is first evidence of a radiative wave propagating ahead of a supercriticalshock. Theoretical work deals with interaction, transport, target acceleration and lasingtransitions.

1.

LASER PROGRAMME DEVELOPMENT .

1.1. The Octal facility
The eight-beam laser facility Octal has been equipped with phosphate Nd
glass. The total IR energy (X = 1.053 /urn) delivered in a 1 ns pulse is 2 kJ;
routine performance is at a level of 1.2 kJ. Frequency tripling is achieved
between the last turning mirrors and the focusing lenses. The eight converters
are composed of type-II KDP crystals which can be easily aligned by an autocollimation process. The conversion efficiency of the tripling system reaches
55% for routine conditions; the maximum UV energy available on target isequal to 700 J (routinely 370 J).
1.2. The Phebus facility
The two-beam facility Phebus has been built in close co-operation with
the Lawrence Livermore National Laboratory, using the laser technology and4ANDR£ et al.

equipment developed there. In December 1985, Phebus demonstrated a powerof 20 TW on target and is now the most powerful laser in Europe. It is alreadybeing used for plasma experiments. The optical pulse is generated by two kinds
of oscillators: a short-pulse oscillator (0.1-0.9 ns) and a long-pulse oscillator(1-5 ns). A third laser beam, which is not yet activated, will be used for plasma
diagnostic purposes. After frequency conversion by two sets of crystals,
providing 2u> (0.53 jttm) and 3w(0.35 nm) capabilities, the beams are focusedon the centre of the target chamber by two f/4 fused silica lenses of 3 m focal
length. The target is a sphere of 2.3 m in diameter; it has 280 windows whichenable the plasma diagnostics to be installed with an accuracy of ±200 fim withregard to the target centre. The maximum energy on target during 1 ns pulses
i s - 14 kJ.

2.

DIAGNOSTICS

The development of plasma diagnostics has also been continued, particularlyin the field of detectors of high temporal resolution and wide spectral range, andsoft-X-ray imaging techniques [1]. The following performance has been obtainedwith a soft-X-ray streak camera prototype constructed around an image converter
tube with lamellar electron optics:- Useful slit length: 12 mm- Screen dimensions: 24 mm along spatial axis
40 mm along time axis
- Spatial resolution with an image intensifier: 17 line pairs per millimetre
over the screen at 35% contrast- Temporal resolution: 3-5 ps for a sweep speed of 2 X 109 cm-s - 1 .

3.

THERMAL TRANSPORT AND LASER DRIVEN HYDRODYNAMICS

3.1. Energy transport
Energy transport in laser plasmas has been studied with multilayered targets;
this seems to be a powerful technique for the study of thermal front penetration.
We employed this technique in connection with a major improvement, namely
the time resolved observation of the X-ray emission from a target composed ofeither a gold substrate or an aluminium substrate coated with a plastic layer.
The discontinuity of emission when the thermal front crosses the CH substrateinterface appeared well defined. Numerical simulations of these results as well
as of the X-ray conversion efficiency have been performed to infer values ofthe flux limiter. At X = 0.35 /um and an irradiance of = 1014 W-cm"2, satisfactory restitution is obtained for a thermal flux limitation of f = 0.03 [2].

IAEA-CN-47/B-I-1

I

I
0.2

I

j

0.4
0.6
Time (ms)FIG. 1. Velocity of the glass D-T interface for a gold coated laser imploded microballoon.

3.2. Directly driven implosions
Directly driven implosions of microballoons have been performed with
Octal at X = 0.35 /xm. In spite of the two-dimensional irradiation due to thebeam arrangement, a high yield has been obtained by taking advantage of theabsorption and X-ray emission properties of high-Z materials. With bare targets,the end of implosion appeared to be similar to the collision of planar disks;
this 2-D behaviour could be weakened in the case of gold coating, owing to thenoticeable preheat of the whole target which is due to the soft-X-ray emission
from gold [3].
In this type of target, the glass D-T interface appeared to encounter during
its flight a strong reacceleration. An example is shown in Fig. 1 for a 100 jum
diameter glass microballoon (1.3 (zm thickness) coated with 0.19 fxm of gold.
The process can be described as follows: First the glass is heated by the softX-ray emission from the gold, and an exploding pusher type behaviour drives
the glass D-T interface at velocities of approximately 3 X 107 cm-s" 1 . Meanwhile, a strong shock is generated at the ablation front, propagating withincreasing speed as the glass decompresses; when it is exhausted, the glass D-Tinterface velocity reaches approximately 108 cm-s" 1 . High implosion velocities
are thus obtained, resulting in high neutron yield. Moreover, the initial highaspect ratio of the target is rapidly reduced during the explosion phase, so thata rather stable implosion is obtained.

ANDRfi et al.

(a)d (Mm)200

Au/
OBSTACLE
POSITION

Al PISTONEMISSION(b)
RADIATIVEFRONTOBSTACLE
EMISSIONRADIATIVE
FRONT |OBSTACLE^
EMISSION k

FIG. 2. Comparison of streak photographs (hatched zones) with the results of the radiative
heat conduction model (solid lines). The dotted lines represent the experimental radiative
front. The 10 fxm thick Al piston moves in (a) vacuum, (b) 2 bars of xenon gas, and(c) 6 bars of xenon gas. The laser is incident from the right and the irradiance is about
4X 10* W-cm'2.

3.3. Radiative wave
First experimental evidence of a radiative wave propagating ahead of asupercritical shock has been reported [4]. The target was a kind of miniatureshock tube (200 X 200 prni2 cross-section, 1 cm length) filled with xenon at
variable pressure and closed with a 10 jum thick aluminium piston on which thelaser beam was focused. The quartz walls of the tube permitted transverse
observation along the axis.

IAEA-CN-47/B-I-1

7

Figure 2 shows that, in vacuum, a luminous front moves along the shocktube with a nearly constant velocity of (5±1) X 106 cm-s" 1 (Fig. 2(a)). This
corresponds to the motion of the aluminium piston after irradiation with the1.06 fim laser pulse at an irradiance of (4-1) X 1014 W-cmT2. When the shocktube is filled with xenon gas, two additional effects can be observed:(a) A luminous front moves ahead of the piston with a velocity of slightly morethan 107 cm-s"1, which corresponds to the propagation of a radiative wave.
(b) A high-Z obstacle situated at 750 /urn from the piston's initial position becomes
luminous just at the time when the thermal front arrives. The distance of 750 pm
from the piston to the obstacle corresponds to a few Rosseland mean free paths.
Comparison of Figs 2(b) and 2(c) shows that if the gas pressure is increased from
2 bar (p 0 = 10"2 g-cm" 3 ) to 6 bar, the radiative front velocity is decreased by afactor of 1.4, while the piston velocity is unchanged. These observations are ingood agreement with theoretical predictions.

4.

THEORETICAL WORK

4.1. Interaction physics
The acceleration of a Maxwellian test electron population by a resonant
one-dimensional electrical structure has been studied in the convective regime
with the Zakharov equations [5]. Depending on the non-linearity level, theacceleration is due to either diffusion or trapping by the plasma wave. Actually,
diffusion only occurs for low values of the non-linearity parameter and has beenfound to be of the quasi-linear type. Trapping of electrons is the main heatingprocess in most cases. The present model shows that Landau damping deeply
affects the electron acceleration. Whenever electron heating is taken into accountself-consistently in the Landau damping computation, good agreement is foundbetween the hot electron temperatures calculated with the fluid model and those
calculated with a one-dimensional particle code.
The plasma waves generated by resonant absorption of light in the vicinityof the critical density of laser produced plasmas are modelled by a non-linear
Schrodinger equation (NLS), with additional terms accounting for the presenceof a source and the inhomogeneity of the medium:iEt + E x x + 2 ( | E 2 | - a x ) E = s
We study the transition from periodic to 'chaotic' emission of solitons as
a function of the value of the non-linearity parameter p = s 2 /2a 2 . Numericalsimulations of the NLS equation display the transition from a stationary electricfield profile to a periodic regime for p « 0.7, and then the transition from thisregime to a 'chaotic' regime for p « 1.2. In fact, from two-dimensional phase

ANDK£ et al.

T

i

10~2
r

0.5

10-4

•v

\

•r

10-6

\
\
\

'-

r

r
~

I
\

\

20

1

\

\c
\

\
10

-.

\

\

,

\\1
\

V

30

FIG. 3. Reduced distribution functions (unit density and temperature) in a Langdon
equilibrium for various values of the Langdon parameter Zv%mJTe:
A: 0 (Maxwellianj; B: 0.06; C: 0.6; D:6. w is the reduced velocity v(me/Te}ll2,

portraits and three-dimensional projections of the phase space through Poincare
sections, the periodic regime is seen to give place first to a subharmonic regime
after a single-period doubling near p = I and then to the emergence of a second
characteristic period near p = 1.1. The chaotic asymptotic regime is reachedwith p = 1.2 [6]. Other diagnostics are implemented, such as Fourier transformsof the electric field amplitude and a computation of the information dimensionof the attractor in a reconstructed arbitrary (finite) dimensional phase space.
The non-linear evolution of stimulated Brillouin and Raman scattering ina homogeneous plasma has been investigated in several physical situations,neglecting the kinetic effects. On the one hand, we studied the competition
between Brillouin scattering and Raman scattering: In order to find the prevailingmechanism, we first compared the growth lengths of each process as a functionof the plasma parameters; then we numerically investigated the non-linear evolution of the system. On the other hand, we studied the non-linear saturation ofRaman reflectivity due to coupling of electronic plasma waves to ion soundwaves. This work has been performed in collaboration with S. Karttunen fromthe Technical Research Centre of Finland [7]. A code has been written for the
study of the problem and different regimes have been observed. Dependingon the laser flux, the reflectivity can be chaotic in time, which corresponds tothe high flux regime, or it can reach a steady state value. In the first case, spiky
spectra and low values of reflectivity are observed; these features are in agreement

IAEA-CN-47/B-I-1

1.

80

FIG. 4. Reduction 0, of the thermal conductivity
as a function ofZ and v\ mJTe.

with present experimental data. In the second case, the coupling of plasma waves
to ion sound waves still leads to a reduction of reflectivity, although it is lessefficient than in the case of a high flux limit.
Particle simulations with the \j -D electromagnetic relativistic EUTERPE
code have been performed in order to quantify the influence of stimulated
Brillouin scattering (SBS) and, more generally, the influence of ion dynamics
on stimulated Raman scattering (SRS). Reflectivities, forward scattering andelectron heating have been measured in plasmas of both homogeneous andinhomogeneous density, irradiated at 2 X 1015 W-cm"2 laser irradiance. The
electron temperature was initially 2 keV. With the flat profile plasma at
0.2 n c (critical) density, various ion masses and temperatures were used. The
lighter and the colder the ions, the more inhibited were SRX and the associatedfast electrons. Upon the onset of SBS, SRS saturation was observed when SBSdriven sound waves reached a large amplitude [8]. Two types of non-uniform
plasmas were chosen: one with (0.19-0.28) n c and the other with (0.15-0.22) n c
in a linear profile. In both cases, ion mobility led to an increase of the SRS
threshold and a reduction of the scattering efficiency. In the second case, witha long density scalelength (250 laser wavelengths), an increase of SRS has been
observed; this can be explained by a change of the SRS growth behaviour froma convective one to an absolute one, which is due to SBS drive turbulence [9].
4.2. Transport phenomena in non-Maxwellian plasma

A.B. Langdon calculated the electronic distribution function for a plasma
undergoing inverse bremsstrahlung absorption in an intense laser field. This led
us to extending the usual local thermodynamic equilibrium. If the hydrodynamic
quantities n e , T e , V (electron density, temperature and velocity) and the Langdon

10

ANDRE et al.

parameter Zvom e /T e (Z is the ion charge and v0 is the quiver velocity of anelectron in the laser field) all vary slowly in space and time compared with thecharacteristic length and time of electron-electron collisions, then the electrons
will be in a 'local Langdon equilibrium'. The distribution function for heatedelectrons then has a self-similar form, which is calculated numerically with theLandau-Fokker-Planck collision term and the Langdon inverse bremsstrahlung
absorption term. The limiting cases of low field and high field are the Maxwellian('exp(-v 2 )') and super-Maxwellian ('exp(-v 5 )') distributions (Fig. 3). Localtransport coefficients can be calculated from this equilibrium distributionfunction, as Spitzer and Harm did with Maxwellians. In particular, the heat
flux is written as: q = - 0 t K VT e + K'Vn e ; the thermal conductivity is reduced
by a factor of between 0.3 1 and 1 (Fig. 4), and there is a heat flux along thedensity gradient (even at constant temperature). This heat flux can be introduced
in a hydrodynamic code. In some cases, the corona temperature is found to
decrease to the laser temperature, whereas the usual code would give an isothermal
corona. This effect has been seen in Fokker-Planck simulations.
4.3. Laser-target acceleration
During the laser-target interaction, the propagation of an electron conduction
wave gives rise to ablation of the front surface of the target. This ablation creates
an ablation pressure P a which acts on the hydrodynamic behaviour of the innermaterial. Our model [10] describes this action, neglecting a possible radiative
preheating [11], and gives the following phases:
(a)
(b)

te)

Formation of the first shock into the target.
Sequence of shocks (or compressions), followed by rarefaction waves:propagation of the first shock, supposed to be a strong shock driven by Pa
in a material with y = 5/3; formation of centred rarefaction waves at therear surface which move back through the target; the reflection of these
expansion waves on the conduction front (or ablation front) produces
new compression waves; after this first cycle of rarefaction and compression
waves, a sequence of similar cycles begins in which, however, all waves are
mixed; this can be described by the pushing of a constant density target
by PaWhen the conduction waves emerge from the rear surface at time t s (endof ablation), the influence of the laser is stopped and the evolution of therear surface can be described by studying the expansion of a plasma at
uniform temperature.

Scaling laws permit us to link Pa (supposed to be constant during ablation)with the laser characteristics. The mass ablation rate, obtained in the same manner,
gives the duration of ablation. For a comparison with the results of some experiments [11, 12], we calculated the velocity of the rear surface of an aluminium

IAEA-CN-47/B-M

in

2

4

11

6 8 10Initial thickness,

F/C. 5. Rear velocity U as a function of the initial thickness e0 of an aluminium target
submitted to a laser (0.35 fjun, 2±0.5 X 1OM W-cm'2, 1 ns), obtained by: X-ray shadowgraphy (a), 0.53 \xm shadowgraphy (o), and a theoretical model (*).

target, as a function of the laser characteristics and the target thickness, at time t mwhen the velocities are experimentally measured (corresponding to the end of
the laser pulse). For thin targets, a maximum velocity is reached if t m is greaterthan t s . A minimum constant velocity is obtained for thick targets; the thicknessof these targets is such that, during t m , only the first shock can emerge from therear surface. Figure 5 gives a comparison of our model with the results of anexperiment performed at Limeil. For this experiment, as well as for others(performed at Limeil and Palaiseau [10]), the data are in good agreement withthe results of our model. Particularly, we give an explanation for the existenceof a maximum velocity in a thin target and a minimum velocity in a thick target.
4.4. Lasing transitions
Following the accomplishment in 1984 of the first-ever unambiguous
demonstration of soft-X-ray lasing at Lawrence Livermore National Laboratory[13], at wavelengths of 206 A and 209 A in a laser produced selenium plasma,
we focused our efforts on the construction of an atomic model of highly ionizedions. This model includes the thirteen most relevant F-, Ne- and Na-like
configurations: Is 2 2s 2 p s , 2p43J2, 2p 6 , 2p53C and 42', 2p63)2 (£ = 0, 1,2;
8' = 0, 1,2, 3), which represent 129 levels in LSJ coupling. The excited levels
above 4f or 3d through n = 10 are consolidated by principal quantum numbers.
This atomic plasma simulation is completed by the ground states of the otherionization stages, consolidated by n-levels (n < 11). These various states arecoupled by electron collisional excitation and de-excitation, spontaneousradiative decay, collisional ionization, and three-body radiative and dielectronic

12

ANDR& et al.

recombination. The main atomic problem is the determination of all the level
energies, statistical weights and spontaneous radiative probabilities with theSUPERSTRUCTURE [14] computer code, the calculation of the different collision
strengths for four different incident electron energies with the computer
package DW + JJOM [15], and the determination of the dielectronic recombination rate coefficients with the AUTOLSJ [16] code; all these calculations are
performed at the Observatory of Meudon. Compared with published data, they
seem to be satisfactory (maximum of relative errors: 1%) [17, 18] and they aremuch more complete (first evaluation of the collisional excitation rates along
the 4-3 and 4-4 transitions). These data have been or are being established fornine elements (Ar, Fe, Ge, Se, Kr, Sr, Y, Mo and Ag) and we try to detectZ-scalings for all of them, using non-linear least-squares methods. All theseatomic data are used in the time-dependent model LASIX which, coupled withthe hydrodynamic 1-D code CHIVAS, restitutes the correct gains from the LLNL
observations and allows us to clarify the major discrepancies between these experiments and theory (low gain observed on a J = 0 to J = 1 3p-3s transition at 183 A[19]). If we assume that a selenium plasma is created by a 0.53 jum laser with
5 X 10 13 W- cm" 2 , the hydrodynamic evolution of the plasma can be described byan electronic temperature which reaches 1.2 keV and decreases to 700 eV, anda flat compression factor (about 10~3). Under these conditions, the averagenuclear charge of the plasma may be considered to be constant (Z ~ 25.70,
corresponding to a predominance of oxygen-like and fluorine-like ions).
Concerning the 3p-3s transitions in neon-like selenium, the highest gains are
predicted for the J = 2 to J = l lines at 206 A and 210 A (10 cm"1 and 5 cm" 1 ).
The gains predicted for the J = 0 to J = 1 transition (at 183 A) do not exceed
0.25 cm" 1 , which is in good agreement with the Livermore X-ray experiments.
However, LASIX and our atomic model allow us also to explore new lasing
fields, especially with shorter wavelengths (4-3 transitions, with a change of theprincipal quantum number, for example at 30 A for selenium) and with othertarget materials (molybdenum or silver; in these cases, the 3p-3s wavelengths
are reduced to 133, 139 and 131 A for Mo, and to 168, 102 and 177 A for Ag).
All these studies may lead to the determination of the maximum efficiency of
X-ray lasers, i.e. the maximum gain along the shortest possible wavelength, withrespect to the atomic number.

REFERENCES[1] BOURGADE, J.L., et al., Paper IAEA-CN-47/B-IIM, these Proceedings, Vol. 2.
[2] JTJRASZEK, D., BERNARD, M., BILLON, D., BOCHER, J.L., HOLSTEIN, P.A.,
Le BRETON, J.P., LOUIS-JACQUET, M., MEYNIAL, D., SCHNEIDER, Ph.,
"Thermal transport experiments at 0.35 jum wavelength", Laser Interaction and
Related Plasma Phenomena (Proc. 7th Int. Workshop Monterey, 1985).

IAEA-CN-47/B-I-1

13[3] BURESI, E., COUTANT, J., DAUTRAY, R., DECROISETTE, D., DUBORGEL, B.,
GUILLANEUX, P., LAUNSPACH, J., NELSON, P., PATOU, C, REISSE, J.M.,
WATTEAU.J.P., Laser Part. Beams 4 (1986) 531.
[4] BOZ1ER, J.C., THIELL, G., Le BRETON, J.P., AZRA, S. DECROISETTE, M.,
SCHIRMANN, D., Phys. Rev. Lett. 57 (1986) 1304.
[5] GALMICHE, D., NICOLLE, J.P., PESME, D., Laser Part. Beams 4 (1986) 439.
[6] LARROCHE, O., PESME, D., 'Transition to turbulence in the forced non-linear
Schrodinger Equation", Anomalous Absorption (Proc. 16th Conf. Lake Luzerne,
NY, 1986), Laboratory for Laser Energetics, University of Rochester (1986) Fl 1.
[7] CASANOVA, M., KARTTUNEN, S., PESME, D., paper presented at CECAM Workshop
Orsay, 1986.
[8] BONNAUD, G., to be published in Laser Part. Beams.
[9] BONNAUD, G., to be published in Plasma Phys. Controll. Fusion.
[10] BABONNEAU, D., DI BONA, G., FORTIN, X., "A model for the acceleration of laser
irradiated targets", Laser Interaction and Related Plasma Phenomena (Proc. 7th Int.
Workshop Monterrey, 1985).
[11] THIELL, G., MEYER, B., AUSSAGE, P., FORTIN, X., Opt. Commun. 46 (1983) 305.
[12] FABBRO, R., FARAL, B., COTTET, F., ROMAIN, J.P., J. Appl. Phys. 56 (1984) 3204.
[13] MATTHEWS, D.L., Phys. Rev. Lett. 54(1985) 110.
[14] EISSNER, W., JONES, M-, NUSSBAUMER, H., Comput. Phys. Commun. 8 (1974) 270.
[15] SARAPH, H.E., Comput. Phys. Commun. 3 (1972) 256.
[16] DUBAU, J., LOULERGUE, M., J. Phys. B 15(1981) 1007.
[17] HAGELSTEIN, P.L., JUNG, R.K., Rep. UCRL-93588, Lawrence Livermore Lab.,
California Univ., Livermore (1985).
[18] HAGELSTEIN, P.L., Rep. UCRL-93811, Lawrence Livermore Lab., California Univ.,
Livermore (1985).
[19] JACQUEMOT, S., BERTHIER, E., VUILLEMIN, M., CORNILLE, M., Analysis of
lasing transitions in a laser-produced selenium plasma, J. Phys. (Les Ulis), Colloq. C6,
Suppl. No. 10, 47(1986).

DISCUSSION
C. YAMANAKA: Instability due to R-T phenomena is a very interesting
but difficult problem. How do you eliminate such effects in pellets in a stateof high compression?
J.-P. WATTEAU: We hope that with our plane target experiments it will
be possible to model such instabilities in the non-linear regime as well as thecorresponding mixing, in order to improve our implosion target design and to
reduce this effect; improvements will be checked by observation of neutron
emission with time resolution.

IAEA-CN-47/B-I-2

PROGRESS OF INERTIAL CONFINEMENT FUSIONAT LAWRENCE LIVERMORE
NATIONAL LABORATORY*
E. STORM
Inertial Confinement Fusion Program,
Lawrence Livermore National Laboratory,
Livermore, California,
United States of America
Abstract
PROGRESS OF INERTIAL CONFINEMENT FUSION AT LAWRENCE LIVERMORE
NATIONAL LABORATORY.
The attainment of high gain fusion pellet performance with a multi-megajoule driveris required for the full realization of the potential benefits of inertial confinement fusion (ICF).
The achievement of high gain is also necessary for experimental demonstration of the scientificfeasibility of economic ICF power production. The completion of the 100 kJ/100 TW class
Nova laser target facility has permitted extension of the short wavelength laser/target interaction and implosion experiments to target and plasma scales more relevant to the high gain
parameter regime. New results have been obtained on basic target/driver interaction anddiagnostic development.

The full realization of the potential benefits of inertial confinement fusion(ICF) requires the attainment of high gain, i.e. high yield (400-1000 MJ), fusionpellet performance with a multi-megajoule driver. The attainment of ignition
and breakeven will be an important scientific achievement, but high gain isrequired to experimentally establish the scientific feasibility of economic ICF
power production. In addition, the large fluxes of X-rays and neutrons available
from such a target would satisfy the requirements for a broad range of applications.
To achieve high gain with 5-10 MJ of driver energy, the requirements on thetarget implosion are stringent. The 2-10 mg of D-T fuel must be compressed toa density of about 200 g/cm 3 . The density-radius product (pr) of the fuel must
be greater than about 3 g/cm2 for efficient burn. The thermonuclear burn of
the fuel must be initiated in a high temperature (3-5 keV) central hot spot (a few
per cent of the fuel mass). The pr of this region must be above 0.3 g/cm2 so thatthe alpha particles will deposit their energy locally and initiate burn in the cold,
main part of the fuel. The fuel-to-ablator mass ratio must be as large as possible,
Work performed under the auspices of the United States Department of Energy,
under Contract No. W-7405-ENG-48.

15

16

STORM

implying that cryogenic fuel must be used. The implosion velocity must be
uniform over the spherical surface of the capsule to within about 1% in order
to avoid fluid instability breakup of the ablator during the compression. To
allow the lowest driver energy, the driver/target coupling should be very efficientand the fuel preheat low. Preheat from early shocks and X-rays must be minimized
as well as that from hot electrons to keep the fuel on a nearly Fermi-degeneratecompression path. To limit the impact of instability growth, it is advisable tohave a low capsule aspect ratio (ratio of the shell radius to its thickness) and to
gradually accelerate the pusher during the whole implosion. In order toaccomplish this, it is desirable to temporally shape the driver pulse. Our strategy
for accomplishing all of the above requirements is to emphasize radiation (X-ray)driven targets, to seek methods of using cryogenic fuel, to study laser/target
coupling at wavelengths of < 0.5 ^m, and to develop a low cost (< US $50 perjoule) driver capable of igniting a high-gain target.
The completion of the 100 kJ/100 TW class Nova laser target facility has
allowed us to extend our short wavelength laser/target interaction and implosion
experiments to target and plasma scales (~10 4 times A) more relevant to thehigh gain parameter regime. Recent results relevant to basic target/driver interaction and diagnostic development include: confirmation that stimulated
forward Raman scattering is responsible for the production of a 'super hot'component of the X-ray spectrum for laser irradiated targets, and quantification of this process as a function of laser wavelength and intensity;
quantification of X-ray conversion efficiency as a function of laser intensity(5 X 10 1 2 <I <5 X 1015 W/cm 2 ) for X = 0.26, 0.35, 0.53 and 1.06 jum;
generation of 130-150 Mbar pressures; fusion yields of 1013 neutrons andefficiencies of 0.18%, with directly driven D-T filled glass microshells as asource of 14 MeV neutrons for advanced neutron diagnostic development.
In recent indirectly (X-ray) driven implosions, confinement density-time
products of (1.5-3) X 1014 s/cm3 were obtained at D-T fuel ion temperaturesof 1.5-1.7 keV (see Fig. 1). In future experiments, using the full 50-80 kJ(0.35 jum, 1.5-3 ns) drive capability of Nova and more fully optimized targets,
we expect to obtain several-fold improvements in these implosion parameters.
In the area of materials science and target fabrication, perhaps the most
exciting progress has been the development of low density, low Z foams thatact as a carrier matrix for liquid D-T. This permits the production of nearly
ideal, high gain, directly driven ICF targets with thick (several 100 pm) shellsthat consist of about 80% liquid D-T and about 20% CH.
The demonstration of very high damage threshold phosphate, fluorophosphate and fluoride glasses, extremely high efficiency (S 80%) semiconductor
diode arrays, demonstration of high growth rate (10-15 mm/day) techniques
for KDP crystals and other high conversion efficiency non-linear crystals area few of the recent developments that have contributed to our growing confidence
that low cost (< US §50 per joule) solid state lasers in the 5-10 MJ, 300-500 TW
range are technically achievable.

17

IAEA-CN-47/B-I-2

Directdrive
targetFinalconfiguration

Initial configurationrDrive>| energy—-!

rAblator~10keVhot spot (pR~ 0.3)
convergence - 2 0; 1 V Remaining• ablator
Very dense pR ~ 2 - 3, < 200 g/cm3 >D-T at ~ 1 keVjf Efficient coupling of drive energy to target
I Very uniform pressure and implosion velocity
Need i No preheating of fuelI Low capsule aspect ratio
I Gradual acceleration

FIG. la. High gain conditions are difficult to achieve.

.

|

.

|

Hot spot
12 • v ^ p r - 0 . 3 )

Ion
Temp.
(keV)

|

8

0.2

|

pr ~ 2-3

Hot spot pr \ I
must exceed \6 ~ alpha particle \ 1range forU
4 " effective self U
heating from JA
2
0.0

i

l\

>>.

10

i

0.4

.

1 _ Cold main fuel

- 600
500

~ 400 Density\"~ 300\\~~ 20C

\
\ ^

0.6

0.8

100
1.0

Normalized radius

FIG. lb. D-T fuel at ignition for a typical high gain capsule. For ICF capsules, central
ignition and propagation into compressed D-T is utilized to achieve high gain.

i

i

2.1

0.7

-1.4

-0.5

1.25 2.50 3.75 5.00 6.25 7.50 8.75 10.00

Time (ns)Time (ns)

Short pulse
Gaussian

Picket lance
pulse

FIG. 1c. Several different temporal pulses have been generated and propagated on Nova.

18

STORM

XUV laser lines observed to date•Ne-like selenium: 182, 206, 209, 221,263 A•

Ne-like yttrium: 155, 157, 165, 218 A•

Ne-like molybdenum: 104,131, 133, 139,142 A•

H-like oxygen: 102 A•

Nickel-like Europian: Several lines,
80 A - 110 A (preliminary)Measured output power to date is > 1-10 MW for Se: 206,
209 A lines
Propagation in inhomogenous plasma is an issue for XUV lasers
X-ray optics for cavity components have been fabricated and tested

FIG. Id. Status of XUV laser experiments at Nova.

iir

I

I

••I

• XL = 0.53 ^m• T = 0.5 ns• Exploding foil
target

•

10

3

•

Line
energy(A.U.)

•

-

+•

r

LASER

~ 1-10 MW

*

101

-

t
+•

n-1

I
1

I

|

2

3

4

Target length (cm)FIG. le. Demonstration of an amplification of > 10s for the 206 A 5e 24+ line.

19

IAEA-CN-47/B-I-2!),.„ «- Hot electrons""seal

10-

"I

•I

°
10 - 2aFraction ofenergy insuprathermal
electrons

X/
s

P'

Novette 0.53/im x-

X /10 - 3 _
y ,-

^6
10

V - Nova 0.35 /jm a

/
t

vX

s

X-"

-4

X X

'

:

• XX5P
/

*

,.l
I
10"*
10-3
10"2
Raman scattered fraction

i X,

io-

10-

FIG. If. Stimulated Raman scattering appears to be the dominant source of suprathermal
electrons in large plasmas.

i,
£

I3

I
*

0

1

2

3S0
330
Wtvtllnglh Inm)

900
1064
WivtUnjth Inml

Downshifted

1
0

360

370

380

390

400?

/800

840

Wavelength |nml
0.47

0.43

0.39

0.35— w epw

Anti-stokes
Kcat = w o + "epw'

880

92C)

960

1000

0.45

0.47

Wavelength (nm)""w

epw— 0.37

"°

0.40

0.43

Stokes
Kcat = w o ' "epw*

FIG. Ig. Simultaneous observation of Stokes and anti-Stokes radiation has unambiguously
identified forward Raman scattering.

20

STORM

10 u

' I ' ' ' I ' ' ' I ' ' ' I
I > 1 0 1 5 W/cm 2 , low Z plasmas
1 ns
Novette

10" 1

Raman
fraction

Novette2

Novette
4u

2OJ

10.-3

10-4

0

n

Argus

200

•2? . I , . . I . . . I . .
600
1000
1400
1800
Density scalelength L A

FIG. lh. Efficiency of stimulated Raman scattering depends on the density scalelength.

40

1014

10,15

10 16

Incident intensity (W/cm )F/C. 1«. ^ A absorption of large spot, sub-micrometre laser produced plasmas
has been verified.

21

IAEA-CN-47/B-I-2

l.U

1

I

1

Au disks
0.35 fim

0.8

IT

_

1ns

A

0.6
1 ns with
prepulses<he<1.5keV)
0.4
0.2

* !

10"

-

I

i

1

101

10.15

10 16

Incident laser intensity

FIG. 1). X-ray conversion efficiency exceeds 70%forl~10l*

10 -A NovaLaser wavelength• 1.06 pm• 0.53 jim
A 0.35 ,um

3

io-

W/cm2.

I Gekko 12*
A Omega• Shiva

Fusionefficiency• Argus
10"
10"

•

Janus

D a u

for

,|mp|,_ t h j n . w a | | e d i

highInitial adiabat type targett

10"

,

10

,

, , I

100

,

,

, • I

.

. . . I1000.

10,000,• I

100,000

Laser energy (J)FIG. Ik. Simple, direct drive target performance has improved over the past years.
(* Such targets do not scale to high gain for reasonable drive energies.)

22

STORM

NOVA

NOVETTE

20 kJ,3w< 0.353 pm)

8 kJ, 2co (0.53 ;um)nr ~ 2 - 4 X 10 1 4 (estimated)n r ~ 5 - 1 0 X 10 1 3

Tj = 1.7 ± 0.2 keV (measured)T ~ 0.4 keV

Neutron yield = 5.0 t 0.5 X 10 1 0

Neutron yield = 1.0+0.1 X 10 7

(measured)(Near calculated I-D values)(10 - 100 reduction from
I-D calculations)

Needed for high gainm ~ 10 15
Tj ~ 5 keV(In hot spot withrange of alpha particles)

FIG. 11. Recent Nova radiation driven high density implosions achieved a major improvementin the fuel implosion conditions for laser driven 1CF capsules.

SHIVA v \ ^1979 y I

ALCATOR "A"
\
ARGUS
^ • \
JANUS
1975

1.0
(10)

10
(100)
Ion temperature

FIG. lm.

100 (keV)(1000) (million
degrees C)

Thermonuclear conditions achieved in fusion experiments.

23

IAEA-CN-47/B-I-2

Ultimate_performanc
100 -Target gain

/

= 0.8
^

-

~

—

-

•

•

:

/

-

,

in

0

= 0.5

''First high gain
laboratory
experiments
1
5

,

,

,

,
10

Driver energy (MJ)

FIG. In. The present experimental and calculational database gives us high confidencein achieving high gain for multi-megajoule drivers. Improvements in physics understanding
and target technology will raise the gain curves.

10b

1.0

10

100

Peak Power (TW)

FIG. lo. The peak power and energy capability of Nd:glass lasers has progressed rapidly.
It is technically ensured that multi-megajoule, 500 TW solid state lasers can be built.

24

STORM
DISCUSSION

H. HORA: After the very great progress you have made, your aim is now
spark ignition with a 1000-fold increase in solid-state D-T. Taking into accountthe difficulties of spark ignition, could you comment on Yamanaka's volume
compression system, where lower densities can yield the same gains?
E. STORM: It is our opinion that volume compression heating will require
more energy than in fact would be obtained from a spark plug, but I am
unfortunately not in a position to discuss the details at present.

IAEA-CN-47/B-I-3

X-RAY CONFINEMENT
IN A LASER HEATED CAVITY*
T. MOCHIZUKI, T. YABE1, H. AZECHI,
K. TANAKA, T. BOEHLY, N. MIYANAGA,
H. NISHIMURA, S. IDO, M. YAMANAKA,
K. NISHIHARA, T. NORIMATSU, T. JITSUNO,
M. NAKATSUKA, K. MIMA, S. NAKAI, C. YAMANAKA
Institute of Laser Engineering,Osaka University,
Suita, Osaka, Japan
R. SIGEL, G.D. TSAKIRIS, K. EIDMANN,
P. HERRMANN, R. PAKULA, P. SACHSENMAIER,
S. SAKABE2, S. WITKOWSKI
Max-Planck-Institut fur Quantenoptik,
Garching, Federal Republic of Germany
Abstract
X-RAY CONFINEMENT IN A LASER HEATED CAVITY.
The confinement of thermal x-ray radiation in 0.6-2 mm dia. gold cavities heated by
3 kJ/300 ps/0.53 pm laser pulses from the Gekko XII laser has been investigated. A radiant energy flux
corresponding to a brightness temperature of 192 eV has been achieved with one- to two-fold reemission of the x-rays generated at the wall of the cavity. It is concluded that a higher temperature canbe attained with a shorter wavelength laser.

The possibility to implode a fusion capsule by the very
intense, isotropic soft x-ray radiation generated with the helpof modern pulsed power sources, in particular lasers, is of
crucial importance for inertial confinement fusion [1,2]. The
isotropic radiation is generated in an about mm-size cavity ofhigh-Z material whose inner surface is heated by the source tosuch a high temperature (of the order of several hundred eV)that it becomes an intense radiator of thermal soft x-rays.
In the closed geometry of a cavity the wall may provide radiation confinement through multiple re-emission of the source* Work supported in part by the Commission of the European Communities in the frameworkof the IPP-Euratom Association.
1

2

Currently guest scientist at Institut fur Neutronenphysik und Reaktortechnik,
Kernforschungszentrum Karlsruhe, Fed. Rep. Germany.
On leave from Institute of Laser Engineering, Osaka University, Japan.

25

26

MOCHIZUKI et al.

energy in the form of thermal x-rays. It has recently been foundin preliminary experiments that x-ray generation in a laserheated cavity behaves differently due to energy confinement [3].
Through the confinement effect an enhanced x-ray flux and very
uniform conditions may be generated in the cavity, provided thesource is powerful enough [4,5]. So far,the generation of x-raysin laser-heated cavities has been investigated using low energy
laser pulses [3,5]. In this investigation we report experiments
on radiation confinement using more energetic laser pulses where
the effect should become important.
X-ray generation and confinement in a laser-heated cavity
has been studied using empty 0.6-2 mm/30 um thick gold cavities
heated with 0.3~3 kJ/300 psec/0.53 um laser pulses delivered by
one to six beams of the Gekko XII laser system. The laser lightabsorption was measured by a set of plasma calorimeters. The
x-ray spectral intensity in the cavity was monitored by observing x-rays emitted through a small diagnostic hole. Filtered
x-ray photodiodes (XRD) and spatially resolving transmission
grating spectrometers (TGS) produced time-integrated spectra,
while time-resolved spectra were obtained using x-ray streak
cameras (XRSC) in combination with transmission gratings. The
observation angles were chosen in such a way that the x-ray
diagnostics should not see the cavity wall element which is
directly heated by the laser at the first bounce in the cavity.
The measured x-ray flux, therefore, originates from a wall
element which is indirectly heated by cavity-confined x-ray flux
and multi-reflected laser light. The laser focusing was also
carefully arranged not to irradiate the rim of the observation
hole. As precaution, a plastic shield was attached to the cavity
in the vicinity of the entrance hole so that the direct viewingof the entrance hole rim by the diagnostics is obstructed.
Figure la shows a 2 mm diameter cavity with entrance anddiagnostic hole as seen from the direction of one of the transmission grating spectrometers. The spectrometer consists of a 25
um diam. pinhole with an integrated transmission grating of 1000
lines/mm. The pinhole generates in zeroth order a picture of
the radiating cavity (see Fig. lb) on which the entrance anddiagnostic hole as well as some spurious radiation from theouter surface of the cavity may be recognized. The grating bars
disperse the radiation in one (the vertical) direction andenable an absolute, time-integrated measurement of the spectrum
on calibrated Kodak 101-01 film [5].
An example of a time-resolved x-ray spectrum emanatingfrom the interior of the cavity (through the diagnostic hole) isshown in Fig. 2. It represents the instantaneous x-ray spectral
intensity at the time of peak emission and it has been obtainedby deconvoluting the corresponding XRSC record. The absolutecalibration of the time-resolved spectra is performed by comparison of the (numerically) time-integrated XRSC records to theabsolutely measured TGS spectra. The XRSC measurements show aradiation time of typically ~ 550 ps, i.e. somewhat longer than

27

IAEA-CN-47/B-I-3

cavityL-30

F/G. 7. (1^ Spherical, 2 mm dia. gold cavity with an entrance hole for the six laser beams from the
Gekko XII laser and a diagnostic hole.
(b) Soft x-ray pinhole image of the cavity projected onto film by a 25 nm dia. pinhole with an integrated
transmission grating of 1000 lines per mm. Note the spectra generated in the vertical direction by theradiation emanating from the entrance and diagnostic hole.

the laser pulse duration of 300 ps (a detailed account of thetime-resolved XRSC spectra will be given elsewhere).
By integrating the TGS spectra over wavelength anddividing through the radiation time, the radiant energy flux S
in the cavity is obtained. This flux may be characterized by atime-average brightness
temperature. The highest measured
brightness temperature is 192 eV (associated with the spectrum
shown in Fig. 2) and corresponds to a radiant flux of 1.5 x 1 0 1 4
W/cm2. For the same shot the XRD signal gives a temperature of
195 eV.
Figure 3 shows the radiant energy flux S versus theabsorbed laser flux S. for all the shots made in this experimental campaign. The pT.asma calorimeters indicated an absoptionof typically ~ 0.6 for a wide range of intensity, indicating
efficient confinement of laser light inside the cavity. The

28

MOCHIZUKI et al.

H1MICO (closed geometry)

Planckianexperiment

HIMICO (open geometry)MA)
FIG. 2. Instantaneous spectral intensity at peak emission of the radiation emitted from a gold cavity(D = 600 fim) heated by three beams of the Gekko XII laser with an average laser intensity of
3.0 x ]014 W/cm2. The solid line represents the experimentally obtained spectrum. The accuracy inthe absolute values is estimated to be ±20%. The dashed lines represent the HIMICO code predictions
for open and closed geometry. The theoretical spectra have been numerically smoothed to give a spectral
resolution corresponding to the experimental one (~ 5 A). For comparison, the Planckian spectrum isshown (dotted line) corresponding to an average brightness temperature of 192 eVand a radiation timeof550ps.

large scatter in the data for the three shots with the highest
intensity may be due to plasma filling of the cavity and tomisalignment of the laser beams resulting in deposition of laser
energy at the rim of the entrance hole. This energy, althoughabsorbed, does not contribute to the cavity heating.
The radiation confinement in the cavity was studied inparallel theoretically, using 1) self-similar solutions [4] incombination with the code HOLRAD [5] and 2) the 1-D sphericalhydrocode HIMICO [6]. Results of these calculations are includedin Figs 2 and 3.
The first approach gives the basic scaling relations andlimits under idealized assumptions. A planar cavity of two
distant walls is considered. The density and temperature profileof the expanding hot wall material is obtained from the LTE
solution of the radiative energy transport problem into thewalls, the ablative heat wave [4]. Consequently, HOLRAD solves
the radiation transfer equation for these profiles on the basisof spectral opacities for gold, calculated in hydrogenic average
ion approximation [5]. In this way, spectra and a corrected x-ray
flux are obtained, taking the finite optical thickness of the
hot material into account. Figure 3 shows the HOLRAD prediction

29

IAEA-CN-47/B-I-3^SS"**-closed^

ATGS-I\HIMICO

ATGS-2

DXRD

10'3
10 u
absorbed laser flux SL (W/cm2)

10

15

FIG. 3. Radiant energy flux Sx through a diagnostic hole of a laser heated cavity versus averaged(over the cavity inner surface) absorbed laser flux SL. The points represented by triangles are obtained
from the TGS data with a radiation time of 550 ps. The open squares are the peak emission values fromthe XRD signal. Also shown is the scaling law predicted by the self-similar solution ofRef. [4] for LTE
conditions and maximum possible material opacity as well as for realistic gold opacities and correction
for finite optical thickness (HOLRAD code). Finally, the predictions of the fully self-consistent code
HIMICO for closed and open geometry are represented by the shaded areas.

for gold under the optimistic assumption that the total absorbed
laser flux is converted into x-ray flux which serves as input tothe ablative heat wave in the cavity wall. Also shown is the uncorrected LTE prediction for a wall with the maximum possible
opacity [4] . It is considered as the upper limit of the radiantenergy flux one can hope to achieve in the cavity.
A more detailed, self-consistent simulation of theexperiments is afforded by HIMICO [6]. This code includes multigroup radiative transport, non-LTE atomic physics, detailed
physics of laser-plasma interaction, the actual laser pulse
shape and the effects of the spherical geometry and the finite
cavity size. As input,the measured values of the absorbed laser
intensity are used. The closed geometry is simulated by imposing
the boundary condition of vanishing net x-ray flux at theplasma-vacuum interface. Physically, this condition means thateach wall element accepts, in addition to the direct laser intensity, an amount of x-ray intensity from the opposite wall
equal to the amount emitted. To simulate the case of open geometry, we remove this boundary condition, thus allowing the x-ray

30

MOCHIZUKI et al.

intensity generated by each wall element to escape without
reaching the opposite wall.
A measured XSRC spectrum and the corresponding timeresolved HIMICO spectra for open and closed geometry are compared in Fig. 2. As is seen, there is good agreement, both inabsolute value andoverall shape of thespectrumwiththeprediction for closed geometry. The spectrum calculated for open
geometry is seen to be lower by a factor of about 3 in absolutevalue than the measured one. It is found in these experiments aswell as in previous ones [5] that as the absorbed laser intensity increases, the spectrum approaches the Planckian corresponding to the LTE prediction for the wall temperature.
However, spectral structures attributed to N- and O-shell excitation of multiply ionized gold still remain visible,as seen in
the spectrum of Fig. 2. This is a result of incomplete equilibrium in the cavity at the achieved conditions.
As seen in Fig. 3, the more realistic HIMICO simulations
give a somewhat lower radiant flux than HOLRAD. However, theslope is very similar for both codes in the intensity range of 2
to 7 x 1 0 1 3 W/cm2 where the finite cavity size should not be
important. It is found that HIMICO reproduces well the experimental points in this range. The data show the ratio S /S., to beclose to unity.
Based on this agreement the quality of radiation confinement achieved in the experiment may be estimated by comparison with HIMICO predictions for open geometry. A simple energybalance consideration gives the ratio of the radiant energy flux
for closed and open geometry at constant source flux as
S (closed)-*= 1 + NS (open)
where N(circulating radiant energy flux/ablative heat wave
source flux) is the quality factor for radiation confinement [4]in the cavity (being numerically equal to the number of re-emissions of the source energy in the cavity). It should be emphasized that the flux into the ablation heat wave is only a fraction of the absorbed laser flux because, as suggested by planar
target experiments [7], the absorbed energy is only partiallyconverted in the transition region into thermal x-rays. With S(closed)/Sx (open) - 2 - 3(see Fig. 3), one finds N = 1 - 2?
i.e. a one- to two-fold re—emission of the energy converted into
thermal x-rays at the wall of the cavity. This interpretation of
the results is, however, only approximate because the differencein laser conversion efficiency and re-emission from the ablative
heat wave between open and closed geometry and the influence ofthe radiation from the transition region has been neglected.
A radiant energy flux exceeding the absorbed laser flux
could not be achieved in these experiments because the con-

IAEA-CN-47/B-I-3

31

finement effect is more than compensated by the low instantaneous conversion rate of laser light into x-rays (approximately 20 % according to HIMICO calculations for open geometry).
The large fraction of energy consumed by the plasma in the transition region makes the production of radiation in the thermal
wave region inefficient. Due to the inaccessible geometry of acavity, it is difficult to keep track of the energy partition
into the various channels.
It may be possible to improve the radiation confinementin the cavity by optimizing the wall material with respect toopacity. However, because the achievement of high temperaturesin the cavity requires even for maximum opacity an intense laser
flux, it is of concern that at the high intensity end of theexperimental range the radiation energy flux seems to increase
more slowly than at lower intensities, both in the experimentand simulation. This may be attributed to the degrading laserplasma interaction in combination with plasma filling of thecavity, and also to poor radiation - ablative heat wave couplingsince at higher laser intensities the spectrum produced in thetransition region is considerably harder and, because of its
longer mean free path it penetrates deeper inside the target
wall. Filling can be avoided by choosing larger cavities, butthen larger pulse energies are needed. A more favourable
laser-plasma interaction and improved efficiency of x-ray
conversion should be obtainable by shorter wavelenght lasers[8]. It remains to be seen whether in this way the conditions
required for inertial confinement fusion can be met by lasers.

REFERENCES
[1]

[2]
[3]

[4]
[5]

YAMANAKA, C , et al., in Plasma Physics and Controlled Nuclear Fusion Research 1984
(Proc. 10th Int. Conf. London, 1984), Vol. 3, IAEA, Vienna (1985) 3.
NISHIHARA, K., YUCHI, T., Quart. Prog. Rep. QPR-83-5, Institute of Laser Engineering,Osaka University (1983) 24.
AZECHI, H., et al., Jpn. J. Appl. Phys. 20 (1981) L 477.
MOCHIZUKI, T., et al., Jpn. J. Appl. Phys. 22 (1983) L 133.
MOCHIZUKI, T., et al., Quart. Prog. Rep. QPR-84-11, Institute of Laser Engineering, Osaka
University (1984) 16.
OKADA, et al., J. Appl. Phys. 59 (1986) 2332.
PAKULA, R., SIGEL, R., Rep. MPQ-85, Max-Planck-Institut fur Quantenoptik, Garching(1984); Z. Naturforsch., A 41 (1986) 462.
TSAKJRIS, G.D., et al., Europhys. Lett. 2 (1986) 213.
HERRMANN, P., et al., Z. Naturforsch., A 41 (1986) 767.
EIDMANN, K., et al., Laser Part. Beams 4 (1986) 521.

32

MOCHIZUKI et al.

[6]

YABE, T., YAMANAKA, C , Comments Plasma Phys. Controll. Fusion 9 (1985) 169.
ITOH, M., YABE, T., KIYOKAWA, S., Jpn. J. Appl. Phys. 25 (1986) L 45.
ITOH, M., et al., accepted for publication in Phys. Rev., A (1986).
MOCHIZUKI, T., et al., Phys. Rev., A 33 (1986) 525.
KODAMA, R., et al., J. Appl. Phys. 59 (1986) 3050.

[7]
[8]

DISCUSSION
H. HORA: In your curve showing the calculation using the HOLRAD code, theX-radiation attained higher power than the laser power. I am sure that this can be
explained by changes in the pulse length. Could you comment on this?
S. WITKOWSKI: The measured internally circulating radiation phase may behigher than the laser phase because of radiation confinement. The radiation losses
through the cavity holes are small and can be disregarded. Therefore, the radiation
inside the cavity may persist for a longer time than the laser pulse, even if its intensity
is higher.

IAEA-CN-47/B-I-4

INERTIAL CONFINEMENT FUSION RESEARCHBY GEKKO LASERS AT ILE OSAKAAND TARGET DESIGN FOR IGNITIONC. YAMANAKA, K. MIMA, S. NAKAI, T. YAMANAKA,
Y. IZAWA, Y. KATO, K. NISHIHARA, T. MOCHIZUKI,
M. YAMANAKA, M. NAKATSUKA, T. SASAKI, T. YABE,
K. YOSHIDA, H. AZECHI, H. NISHIMURA, T. NORIMATSU,
S. IDO, M. MIYANAGA, H. NIKI, E. FUJIWARA,
H. SHIRAGA, M. NAKAI, H. TAKABE, T. JITSUNO,
K. TANAKA, M. TAKAGI, A. NISHIGUCHI
Institute of Laser Engineering (ILE),
Osaka University,
Suita, Osaka, Japan

Abstract
INERTIAL CONFINEMENT FUSION RESEARCH BY GEKKO LASERS AT ILE OSAKAAND TARGET DESIGN FOR IGNITION.
Direct and indirect drive implosions have been investigated by GEKKO lasers with new
diagnostics and computer simulations. For direct implosion, a new compression scheme withlarge high aspect ratio targets has been demonstrated by GEKKO XII green beam D-T experiments to yield 9.4 X 1012 neutrons. The areal mass density of fuel was estimated by thesecondary fusion reactions in D-D fuel. The implosion dynamics investigated by an X-ray
streak camera show shock wave multiplexing by successive acceleration of a pusher. A radiation driven Cannonball implosion has been used to investigate how the laser pulse shape isrelated with the X-ray pulse and the implosion dynamics. The radiation driven ablationand implosion processes could be clarified by X-ray streak pictures with X-ray
backlighting. It has been found that a flat top laser pulse with shorter FWHM is bettera higher neutron yield. For igniting fuel at the centre of a pellet, the symmetry of the hot
spark and the related energy transport are crucial. The high density plasma effects on theheat transport and the hydrodynamic stability were investigated with regard to ignition
target design. Since the contact surface between fuel and pusher is strongly unstable in thestagnation phase, a shock multiplexing implosion is of interest for obtaining data on laserfuel coupling without stagnation. In a one-dimensional simulation of this implosion it has
been shown that D-T fuel could be ignited at a laser energy higher than 100 kJ with
X = 0.35 jxm.

1.

INTRODUCTION

The GEKKO glass laser systems have been used for investigating both direct
and indirect implosions. In 1984, the GEKKO XII was converted to a green
beam laser. The other lasers, GEKKO IV and GEKKO Mil, were changed to33

34

YAMANAKA et al.

0.25 jum and 0.35 /itn, respectively. Because of the lower fluence of hot electrons
in these shorter wavelength lasers, a new implosion scheme (Large High Aspect
Ratio Target, LHART) has become applicable. This is a stable ablative implosionin the shock multiplexing mode [1].
Necessary conditions for a perfect implosion [2] are the following:(a) Good laser-fuel coupling efficiency (better than 5%);(b) Low isentrope (lower than that for a temperature of 10 eV at liquiddensity);(c) High ablation pressure (more than a few tens of Mbar);(d) Small non-uniformity of implosion velocity (less than 3%).
These GEKKO laser experiments confirmed that the short wavelength
laser implosion fulfils conditions (a)-(c). However, the fourth condition was
still in question. With a new implosion scheme, the 'shock multiplexing
implosion', a highly symmetric compression and high neutron yield could beachieved.

2.

DIRECT IMPLOSION

Directly driven implosion has been extensively studied in the wake of thecompletion of short wavelength laser systems. In a long wavelength laser (forinstance 1 nm), hot electrons play a significant role in the energy transport bypreheating the fuel and exploding the pusher. However, in a short wavelength
laser, the energy transport is dominated by thermal electrons, which ablate theouter surface of the pusher, generating high pressure (>20 Mbar) to drive theresidual pusher and fuel inwards.
The general features of conventional implosions with thick-shell targets
are described as follows. The pusher is accelerated by the ablation pressure anda shock wave in the fuel travels faster than the pusher implosion. In a onedimensional simulation, the temporal evolution of the neutron yield has ashelf between the first shock reflection and the imploded pusher stagnation.
The neutrons observed in experiments are close to the value at the shelf.
Therefore it is suspected that the pusher mixes with the fuel during the stagnation period, as has been suggested theoretically [3].
Implosion dynamics and temporal evolution of neutron yield for a largehigh aspect ratio target experiment and simulation are shown in Fig. 1. In this
implosion mode, the neutron production rate has one peak in time since thepusher is accelerated fast enough to follow the shock front in the fuel, where
the maximum neutron production and the maximum compression occur almost
simultaneously when the reflection shock reaches the contact surface. In thisscheme - the 'shock multiplexing implosion' - the highest neutron yield of
9.4 X 1012 has been achieved with the GEKKO XII green beam. A glass micro-

35

IAEA-CN-47/B-I-4• : X-ray Streak Data

(a)
2

r

• j

li

my
.. illr
200
Radius

~

0

600

400
(urn)

200

Core SizeMay' X-ray Pinhole Camera

100
•o

~

a
a.
1 I l l l l l l l l 4 ! " I'll

IX

' '

"

X-ray Streak Camera
Pusher pAR (mg/cm 2 )Fuel pR (mg/cm 2 )Fuel Density
Ion Temperature (keV)

"=

10" r

I io< r

/

exp.

"1

...i..

2.0
2.2
2.4
Time (ns)

2.0
2.2
2.4
Time (ns)FIG. 1. Flow diagram and temporal evolution of neutron yield, ion temperature, fuel density
and area! mass density for GMB of 1235 pm diameter and 1.31 [im thickness with 6.2 atm D-T.
(a) Flow diagram and positions of the pusher-fuel contact surface, measured with an X-ray
streak camera.
(b) Neutron yield, ion temperature, fuel density and areal mass density of a simulation; arrows
indicate the experimental results.

balloon with a large diameter (1235 /im) and high aspect ratio (470), filledwith D-T fuel of 6.2 atm, was irradiated by green beams of 13 kJ with a 1 ns
pulse. This neutron yield corresponds to 2.0 X 10~3 times the input laser
energy. Figure 2 shows the laser energy dependence of the neutron yield andthe pellet gain.as obtained by ILE, Osaka University; Lawrence Livermore
National Laboratory, University of California (LLNL); and Laboratory for
Laser Energetics, University of Rochester (LLE),

36

YAMANAKA et al.

10"

10'
10*

10*
Laser Energy

10

(J)

FIG. 2. Neutron yield and pellet gain versus laser energy from ILE, Osaka, LLNL and LLE.

For the shot with the highest neutron yield, the ion temperature, theneutron yield and the timing of the neutron pulse are shown by arrows in
Fig. 1. Figure 3 presents a comparison of neutron streak measurements ofthe temporal evolution of neutron production with simulation results. The
timing and duration of the observed neutron pulse reasonably agree with the
ILESTAR simulation prediction.
Figure 4 shows the dependence of the neutron yield on the target aspectratio. Although the data points are scattered, the best shots for various aspectratio targets are connected by a curve. The curve indicates that the optimum
aspect ratio, A, is 400-500, and the neutron yield is proportional to A6 for
A < 400.
Another feature of the LHART implosion is the high coupling efficiency
between laser and D-T fuel. The D-T fuel accumulates on the inner surface.
of the pusher and forms a hot dense fuel layer which is accelerated to a very
high velocity (~ 108 cm/s), as shown in the space-time diagram of Fig. 1. In
the final stage of the implosion, the SiO2 pusher is almost ablated and only
the accelerated fuel layer remains. Therefore, we can expect that the coupling
efficiency becomes very high. For the velocity V and the total mass Mp ofthe fuel layer, the total fuel energy E F is approximately given by
1+

I) 2

where 7 is the specific heat capacity ratio.

IAEA-CN-47/B-I-4

37

1.4 1.6 1.8 2.0 2.2 2.4 2.6 2.STime (ns)FIG. 3. Comparison of neutron streak measurements of the temporal evolution of neutron
production with simulation results.

Evaluating the implosion velocity V by the rocket model, we obtain ascaling law for the coupling efficiency, T?C = E F /W L , where WL is the input laser
energy. Figure 5 shows the scalings together with the LHART experimental
results, which are estimated from the ion temperature measured by neutron
time of flight. The curves A, B and C are for target diameters of 1200 nm,
1000 /um and 800 fim, respectively. The mass ablation rate rh is assumed to bem = 1.5 X(I a b /10 1 3 W/cm 2 ) 0 - 5 X 10s g/s-cm2
All experimental data for the large diameter targets («1200 /urn, indicated by °)are approximately on curve A. The experimental data for the smaller diametertargets (•, A and x) are on curves B and C. We note in Fig. 5 that a coupling
efficiency of more than 5% has been achieved in the LHART experiment.

38

YAMANAKA et al.

V

•
•
-

O

o §

In

O

1

o
\

/

e

o
a
o

-

o
o

D

"3

1

V

)

O

o

1
Ix

o
o

O

/

*

A
1

z

\
o

\
o

10'

:

\

X

-

A

A

^

_
-

•

X

X
X

101
200

1

,

1

400
Aspect600
Ratio

FIG. 4. Aspect ratio dependence of neutron yield, o is for target diameters of 1000-1500 jin
a is for 900-1000 urn O4 atm fuel pressure), A is for 900-1000 \im «4 atm fuel pressure),
and X is for less than 900 pm.

UJ

o

o

1.2

EffectiveLaser

E n e r g y •?£-(kJ)

FIG. 5. Coupling efficiency of laser to D-T fuel for large high aspect ratio target.
Target diameters: o is for 1000-1400 fim, n is for 900-1000 [im, and X and & are for
700-900 (Jim.

IAEA-CN-47/B-I-4

Glass

R/AR: 4 5 0 / 1 . 3
10

39

,C8H8

Laser
0.53 .urn
l5.8kJ/800ps
350/1.5^

12(PAR),

2
>
c
o
3

o

z
10 10

10!

0.2 0,4 0.6 0.8 1.0
Initial IpAR (mg/cm2)

FIG. 6. Measured pusher areal mass density (p&RJp: • if for Si activation, a is for Pr activation
and * is for fuel pR deduced from (pAR)p as a function of the initial areal mass density.
Simulation predictions of(pt\R)v , °; (pRj(, A; and neutron yield: *, are also plotted.
(pR){ measured by the secondary reaction is shown by •. The Pr activation signal is shown at
the bottom.

The space resolved X-ray spectroscopy measurements and the X-ray pinhole
images show conclusively that the core is round (30% sphericity) and that the
pusher does not mix with the fuel. Plausible mechanisms for stabilization are
the decrease of the target aspect ratio during the implosion and the decreaseof the At wood number on the pusher-fuel contact surface.
Radiochemical measurements of areal mass density have been developed forthe diagnostics of compressed fuel [4]. The multi-activatable tracer method isapplied for measuring the pusher areal mass density by means of a highly pure
germanium detector. Figure 6 shows the measured pusher pR for high neutron
yield implosions. We were able to clarify how the pusher (pAR) p increases andthe neutron yield decreases with the increase of the initial pusher areal massdensity.

40

YAMANAKA et al.

U

d)
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•

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-ray
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IAEA-CN-47/B-I-4

41

In the first experiments of the secondary nuclear fusion reaction method
we succeeded in measuring the fuel areal mass density of imploded D-D fuel [5].
There are two branches in the secondary fusion reactions:
Neutron branch: D + D = P(3.02 MeV) + T(1.01 MeV)
D + T = a + n(ll.8-17.1 MeV)
Proton branch:

D + D = n (2.45 MeV) + 3 He (0.82 MeV)
D + 3 He= a + P ( l 2 . 5 - 1 7 . 4 MeV)

When the fuel areal mass density (pR)f is lower than the stopping range of T
and/or 3 He, the ratios of the secondary reaction rates (R.^2 for the n-branch andR p 2 for the P-branch) to the primary D-D fusion reaction rates (RT>] and R ^ , ] )are proportional to (pR)f as follows:
n-branch R n;2 /R T ,i = 0-09 (pR) f
3

He-branch Rp )2 /R H e,l = ° - 1 4 CpR)f

The results for the dependence of (pR)f on the initial pusher 2 (pAR)p are shownin Fig. 6. This method is applicable to the fuel pR measurements for the n-branch,
up to (pR) f ~ 2 0 mg/cm2 at Te = 1 keV, and (pR) f ~ 200 mg/cm2 at Te = 10 keV.

3.

INDIRECT IMPLOSION

In order to achieve high density compression, our implosion studies areconcentrated on the radiation driven Cannonball target which has a large outer shelland a large inlet hole. In the Cannonball target, the laser absorption efficiency
depends not only on laser intensity but also on cavity size and pulse width [6]. The
observed absorption efficiencies are duplicated well by the HISHO code withthree-dimensional laser ray tracing. The hard X-ray spectrum and the electron
spectrum from the Cannonball radiation experiments indicate less hot electron
generation and low preheating; however, in the small cavity Cannonball experiment the suprathermal electrons are produced around the inlet hole to preheat
the fuel and degrade fuel compression [7].
The intensity and pulse shapes of soft X-rays which mainly drive the fuelpellet are especially important for achieving high density and high neutron
yield. In the GEKKO XII experiments, we investigated how the soft X-ray
intensity and pulse shape are related with the laser pulse. Figure 7 shows the
temporal evolution of cavity X-ray emission. The X-ray emission from thecavity wall heated directly by laser beams rises simultaneously with the laser
pulse (see (a), (b) and (c)). On the other hand, the X-ray emission from thewall that is not heated directly by laser beams is delayed by more than 300 ps

42

YAMANAKA et al.

Initial Diameter
1
1

4

2

0

2

4

Radius (xlOOjjm)
FIG. 8. X-ray streaked shadow image of a Cannonball target inner shell and the relevant
simulated flow diagram. The closed circles represent the density peaks derived from theshadow data using Abel inversion and assuming axial symmetry of the density profile. The
Cannonball outer shell was a 2000 pm (p gold shell and the inner shell was a 6. 75 pm CF
coated glass microballoon, 401 pm (j> and 0.98 pm thick.

with respect to the laser pulse (see (d), (e) and (f)). The intensity of the emission
from the indirectly heated wall is 2-3 times lower than that from the directly
heated wall for the long pulses ((d) and (e)). For both the tailored laser pulse
and the long square pulse, the conversion efficiency of laser energy to X-ray
energy decreases significantly in the rear side of the pulse, as shown in (a) and(b). The overall conversion efficiency is bad for the tailored pulse and goodfor the short square pulse. The mechanism of these tendencies is as follows.
After about 1 ns, the cavity (diameters 2.5-2.0 mm) is filled with a low density
corona plasma and the injected laser energy is mainly absorbed in it; the corona
plasma can be thermally decoupled from the dense wall plasma [8]. We concluded
that the tailored laser pulse is not appropriate for efficient radiative implosion atpresent laser energy levels. Actually, the neutron yield was higher in the shortsquare-pulse irradiation than in irradiations with the long and/or tailored pulses.
The GEKKO IV results also indicate that long pulse irradiation is not favourable
for the generation of soft X-rays {hv < 1 keV) although the X-ray energy
increases [8]. It should also be noted that in this case the intensity of the
X-ray emission from the indirectly heated wall becomes comparable to thatfrom the directly heated wall (Fig. 7(c) and (f)).

43

IAEA-CN-47/B-I-4

Uranium cavity(a)Gold cavity(b)FIG. 9. Temporal evolution of soft X-ray spectra from a uranium cavity (a) and a gold
cavity (b). The colours indicate the spectral intensity.

With the X-ray backlighting method, we observed implosion dynamics
of fuel pellets driven by X-ray radiation in the gold cavity for the square laser
pulse of 1.3 ns. Figure 8 shows an X-ray shadow image of a Cannonball target
shell and the flow diagram of the HISHO simulation. The observed implosion
velocity of 2 X 107 cm/s agrees well with the value predicted in the simulation.
The ablation pressure obtained by the simulation is 15-20 Mbar. which also
agrees with the theoretical scaling [9] for the observed brightness temperatureof about 130 eV.
In the square pulse experiments, we investigated the shell thickness
dependence of the neutron yield. We compared the simulation results of
HIMICO with the experimental results for neutron yield. The experimental
results for the shell thickness dependence of neutron yield agree reasonably
well with the simulation results for neutron yield at the first shock. We
realized from these results that the pusher fuel mixing is also serious in radiation
driven implosions.
Finally, we succeeded in taking the temporal evolution of the soft X-ray
spectrum of a uranium cavity. Figure 9 shows a comparison of the uranium
cavity spectrum and the gold cavity spectrum. These spectra consist of
nitrogen shell emissions and oxygen shell emissions; the photon energies foruranium are higher by 100-300 eV than those for gold. In Fig. 9, the spectrum
shifts of uranium towards the shorter wavelength are compared with those of gold.

44

YAMANAKA et al.
100p• a-heating ON
A a-heatlng OFF
0.1
10

I
100

I

i

i

I i mlI

1000

FIG. 10. Driver energy dependence of maximum ion temperature (T-J and areal massdensity (pR) of the imploded cryogenic D-T fuel. The case with alpha particle transport
in the simulation code is denoted by solid circles, the case without alpha particle transport
is denoted by solid triangles.

4.

DESIGN OF IGNITION TARGET

For the ignition of fuel at the pellet centre the hot spark symmetry andthe energy transport in fuel are crucial. Since plasmas around the hot spark
are compressed to high density, the high density plasma effects on the radiative
and atomic processes are important. These considerations led us to investigate
the implosion stability and the high density plasma effects.
It was found that the ablation front is rather stable during short wavelength irradiation because of the high mass ablation rate; vortex shielding
and/or fire polishing stabilize the Rayleigh-Taylor modes [10]. However, thecontact surface between fuel and pusher is strongly unstable in the stagnation
phase since there is no ablative stabilization [3]. As a result, the fuel mixes
with the pusher. Therefore, it is difficult to expect high neutron yield in the
maximum compression phase of the thick pusher shell target implosion. We
propose that the fuel shell should be self-compressed, without strong pusher
action. An implosion simulation with a stability analysis indicated that theneutron yield is less sensitive to the initial perturbation in the 'stagnation free'
implosion. The simulations also showed that the D-T fuel could be ignited ata laser energy of more than 100 kJ at AL = 0.35 (im through 'stagnation free'
implosion. Figure 10 shows the results of a simulation of the laser energy
dependence of the ion temperature Tj and the areal mass density pR. The

45

IAEA-CN-47/B-I-4
100 IT]

1 I l|llll|

1 I I|IIM|

1 I I|llll|

1 I I | illlj

1 I ' IU l l |1 I I|llll|

1 I I |MII|

Direct0. S3 gin 1.0 Sum
Single Shell
Double Shellas•Cannonball ( 1 .05um)
No Coal A u C o a t
O•Single Shell«
Double Shell
o

• M

0.01

.I

I

I

10«

10 7

10 s

,

. ,

I

10 9

. . .

I..., I10.
10

, , li.nl

10"I. ,l,,i.l• i .lm,l

10'

10'J

, , ,1,.,.

D-T Neutron Yield (Y N )FIG. 11. Compressed fuel density versus neutron yield. When the pusher mass increases,
the fuel is compressed more, but neutron yield is less. The high density is achieved for the
Cannonball targets and the direct driven double shell targets.

difference between the ion temperature with alpha heating and the ion temperature without alpha heating becomes significant when the laser energy is higher
than 30 kJ. where the fuel pR is 0.6 g/cm 2 . In the stage where E L = 30 kJ.
the neutron yield is about 10 15 , i.e. 3 kJ of energy yield, and the pellet gain is 0.1.
However, it should be noted that with this type of implosion scheme thepellet gain required in a commercial power reactor (~ 100) is difficult to obtainbecause the whole volume of the fuel is heated and such volume ignition prevents
high gain. The actual device is expected to overcome this difficulty.

5.

SUMMARY

Figure 11 presents data for direct drive and indirect drive implosion. We
note that the direct drive double-shell target filled with 110 atm D-D achieved
a high density compression (up to more than 10 g/cm 3 ). The neutron yield of
the direct drive implosion by shock wave multiplexing is very high, as shown by
the data for N y = ] 012—1013 in Fig. 11. The reason for this is that the coupling
efficiency between input laser energy and compressed fuel energy is excellent.
We derived a gain scaling on the basis of the shock multiplexing implosion
experiments. By using a one-dimensiona! simulation, the scaling law was
extrapolated to a case with a larger driver for the design of an ignition target.
When the blue 1 MJ laser is used, the pellet gain is expected to be 30.

46

YAMANAKA et al.

REFERENCES[1] YAMANAKA, C, NAKAI, S., Nature (London) 319 (1986) 757.
YAMANAKA, C, et al., Phys. Rev. Lett. 56 (1986) 1575.
[2] BODNER, S.E., J, Fusion Energy 1 (1981) 22.
[3] HATTORI, F., TAKABE, H., MIMA, K., Phys. Fluids 29 (1986) 1719.
[4] M1YANAGA, N., et al., Rev. Sci. Instrum. 57(1986) 1731.
[5] AZECHI, H., et al., Appl. Phys. Lett. 49 (1986) 555.
[6] BOEHLY, T., et al., Opt. Commun. 60 (1986) 169.
[7] NISHIMURA, H., etal., Opt. Commun. 56(1986)409.
[8] KODAMA, R., et al., submitted to Appl. Phys. Lett.
SAKURAI, K., KONDO, K., NISHIMURA, H., Rep. QPR-86-18, ILE, Osaka University(1986).
[9] NISHIHARA, K., Jpn. J. Appl. Phys. 21 (1982) L571.
[10] TAKABE, H., MIMA, K., MONTIERTH, L., MORSE, R.L., Phys. Fluids 28 (1985) 3676.

IAEA-CN-47/B-I-5

DIRECT-DRIVE LASER FUSION
IN THE UNITED STATES OF AMERICA
R.L. McCRORY, J.M. SOURES, P. AUDEBERT, O. BARNOUIN,
R.S. CRAXTON, J. DELETTREZ, R. EPSTEIN, L. FORSLEY,
R.J. HUTCHISON, P. JAANIMAGI, S. JACOBS, R.L. KECK,
T. KESSLER, H. KIM, R. KREMENS, S.A. LETZRING,
R.S. MARJORIBANKS, F. MARSHALL, P. McKENTY,
G.A. MOUROU, M.C. RICHARDSON, W. SEKA, R.W. SHORT,
A. SIMON, S. SKUPSKY, C. VERDON, J. WARK, B. YAAKOBILaboratory for Laser Energetics,
University of Rochester,Rochester, New YorkS.E. BODNER, J.P. DAHLBURG, M. EMERY, J. GARDNER,
J. GRUN, T. LEE, C. MANKA, E. McLEAN, S. OBENSCHAIN,
B. RIPIN, J. STAMPER
Naval Research Laboratory,
Washington, D.C.
United States of America

Abstract
DIRECT-DRIVE LASER FUSION IN THE UNITED STATES OF AMERICA.
Direct-drive experiments at the University of Rochester's Laboratory for Laser Energetics (LLE)
and the Naval Research Laboratory (NRL) address issues in pellet compression and heating: efficiencyof the coupling of laser energy to the target and the coupling of absorbed energy to the fuel, drive uniformity, hydrodynamic stability, preheat arising from laser plasma instabilities and x-rays, and target
diagnostics. The 24 beam, 2500 J, 351 nm OMEGA laser system at LLE has been used in anexperimental effort to achieve high compressed DT fuel densities. Detailed hydrodynamic computer
simulations at NRL predict that the growth rate of the ablative Rayleigh-Taylor instability is less thanthe classical values. Recent Rayleigh-Taylor experiments at NRL test these predictions.

1.

UNIVERSITY OF ROCHESTER PROGRAM

1.1.

Introduction

Direct-drive laser fusion requires the efficient absorption of theincident laser light by the target, high (>10%) hydrodynamic efficiencyof energy coupling to the encapsulated DT fuel mass with low preheat,and uniform compression of the fuel to densities >2000 times liquid
47

48

McCRORY et al.

density.
The thermalized plasma corona must produce an ablationpressure which drives the implosion with a uniformity of 1% or betterto achieve the compression and high spark plug temperatures (>4 keV)necessary for ignition.
Symmetric uniform laser irradiation in therange of 10 14 to 10 15 W/cm 2 using a laser wavelength of 530 nm or
less, in order to minimize suprathermal electron generation byparametric plasma instabilities which can preheat the fuel and preclude
high gain (100-300), is required to achieve these parameters.
The OMEGA laser is presently in use in a series of experiments todemonstrate the compression of thermonuclear fuel to 100-200 times
liquid density before 1989. The primary elements of the LLE directdrive program are:(a) the definition and control of the laser
irradiation uniformity; (b) the development and implementation of anarray of high-density target diagnostic systems; (c) target design andanalysis of experiments using one-dimensional and two-dimensionalhydrodynamic simulation computer codes; and (d) the implementationof a high quality liquid or solid DT fuel target system.
1.2. OMEGA Laser System
The OMEGA twenty-four beam Nd:phosphate glass (LHG-8) laser
employs rod amplifiers (up to 90 mm diameter). The low nonlinear
index of refraction, high gain, and low thermally induced distortion
provided by LHG-8 allows operation of OMEGA at peak power levels of
0.5 TW per beam (at 50 ps) and peak energy of 175 J/beam(1054 nm at 1 ns) with a 28 min shot cycle.
The beam quality generated by OMEGA is limited by the number
and quality of optical components in each beamline, and results in abeam approximately 10 times diffraction limited.
The frequency
tripled system employing the polarization mismatch scheme [1] withan overall energy conversion efficiency of 80% from 1054 nm to351 nm has been operated for over one thousand full system shots.
To achieve high density implosions from direct-drive laser fusion,
irradiance nonuniformities over the surface of the target must be
restricted to a few percent or less. Studies using two-dimensionalhydrodynamic simulation codes together with a three-dimensional
beam superposition code which calculates refraction and absorption of
the incident radiation in the plasma corona, with absorption
nonuniformities decomposed into a series of spherical harmonics, havedemonstrated the sensitivity of the hydrodynamic performance oftarget designs to target irradiance and sources of nonuniformity: beam
to beam energy imbalance, beam mispointing, single beam intensity
fluctuations and the fixed, 24-beam (f/3.5) geometry. This sensitivity
analysis, together with an experimental characterization of these major
sources has shown that irregularities in each of the 24-beam profiles
are the dominant causes for irradiation nonuniformity.

IAEA-CN-47/B-I-5

49Beam calorimetry can now measure absolute individual beamenergies to 1%, and beam to beam energy imbalance is presently
5% RMS. Beam pointing accuracy is 10 urad. Present emphasis is on
improvement of the illumination uniformity by identification of thosefactors which affect the intensity distribution in the UV target plane,
primarily the intensity and phase distribution of the beam incident on
the focussing optics. Measured beam profiles on OMEGA yield an RMS
nonuniformity, summed over all surface harmonic modes, of 15%.
The 15% nonuniformity on OMEGA should be reduced by morethan a factor of three in the future, due to improvements in opticalcomponents and harmonic conversion crystals, decreased turbulence
levels in the beam propagation path with the installation of beamcorridors, and free propagation spatial filtering of the oscillator
output. [2]Potentially, additional uniformity can be achieved on
OMEGA by using ISI (see Sec. 2 below). Design studies are in progress
to define the engineering problems associated with implementing ISI
on OMEGA.

1.3. High Density Diagnostics and Experimental

Analysis

To characterize the performance of high compression laser fusion
targets, a number of complementary diagnostic techniques must be
deployed. At LLE we are using knock-on DT ion spectrometry [3] andnuclear activation [4] to measure the density-radius product of the fuel(pR) and tamper (pAR); particle and x-ray imaging to characterize thesize and uniformity of the imploded core [5]; neutron and charged
particle spectroscopy to measure fuel temperature; time-resolved andspatially-resolved x-ray emission and absorption measurements tomeasure implosion velocity [6], energy transport [7] and shell pAR [8];and a number of other diagnostic techniques to characterize the low
density coronal plasma [9].
Time-resolved and time-integrated spectroscopic measurementsof thermal transport in spherical geometry have been made using 6, 12,
and 24 beams of OMEGA for incident irradiances in the range of 1 to
4 x 10 1 4 W/cm 2 using Gaussian temporal pulses 600-750 ps FWHM.
Figure 1 summarizes the scaling of the measured mass ablation rate,
deduced using the methodology discussed in [7]. The 24-beam data
most closely matches theoretical predictions, confirming our previous
conclusion [7] that irradiation nonuniformities dominate the 6- and12-beam data, for m as a function of absorbed irradiance for fluxinhibited transport with f=0.1. [10]In 24-beam implosion experiments we have performed the firstsimultaneous measurements of fuel and shell areal density [5]. The
results of some of these experiments, conducted on DT gas filled glassshell targets [11] (in some cases overcoated with parylene) are

50

McCRORY et al.

10 6

o

\

Charge collectors

X

SI, spectr.
A 24 beams
Ca, spectr.
AAl, spectr.
o 6 beams
LILAC simulations•(charge collectors& 10% level)_
500 eV burnthrough D

•E

10 5
10

14

10'

Absorbed Irradiance (W/cm 2 )FIG. I. Mass ablation rates obtained from charge collector traces and from time-integrated x-ray
emission spectra on OMEGA.

summarized in Table I. The primary objective of these experiments,
predicted to achieve compressed densities up to 150 times that ofliquid DT, was to activate and calibrate the various diagnostic systems
to be used in future OMEGA liquid or solid fuel experiments.

2.

NAVAL RESEARCH LABORATORY PROGRAM

2.1.

Introduction

The symmetry problem in "direct drive" laser fusion has now
been solved in principle with an optical technique that we call ISI, or"induced spatial incoherence ". This technique can produce the extreme
laser smoothness that is required for direct drive laser fusion, and also
produce the overall required symmetry on the pellet during an
implosion [12]. We have invented variants of ISI that eliminate thetechnological complexity of echelons [13]. We have also shown
experimentally that ISI has the side benefit of controlling most of thedeleterious laser-plasma instabilities [14]. Recently we extended theseexperiments to study the Raman mode, and we found similar effects.
In order to obtain economic electrical energy with realistic laser
efficiencies, the energy gain of the pellet must be rather high — of order

51

IAEA-CN-47/B-I-5

TABLE I. DATA FROM LLE EXPERIMENTS AND PREDICTIONS3
Neutron Yield

Density

(pAR)

(PR)

Meas.

Pred.

Pred.

8

27±19

33

24.4

3

1.1 ±0.1

3

1.0

4

17±7

11

9.2

2.1±0.3

3

—

3

0.8

244

—

3

2.1±0.1

4

1.3

Meas.

Pred.

6

20 ±10124

0.6±0.06—

Shot No.

Meas.

Pred.

11162

0.002

11200

2.70

11218

0.01

4

11233

11.30

233

11310

6.70

11311

13.20

416

—

3

1.5+0.1

4

1.3

11332

0.02

6

—

4

28 ±6

14

11.1

11333

0.02

10

—

6

19±5

22

13.8

13092

4.5

150

—

3

1 ±0.1

3

1.4

13096

0.02

83

—

16

9.6±4

77

36.9

16

2.2±0.5

54

36.2

13111

0.12

228

-

' Measured (Meas.) and Predicted (Pred.) neutron yield (1010 neutrons), areal densities (mg/cm2) anddensities (g/cm3).

150-300 for a laser efficiency of 5-7%. The only pellet designs that are
currently predicted to achieve this high gain are directly driven pellets,
using relatively thin shells with limited laser pulse shaping. At one
limit, one can use pellets with an initial aspect ratio R 0 /AR 0 of order
100 and a laser wavelength of about one micron [15]. At the otherlimit, one can use a somewhat thicker shell with an initial aspect ratio
of about 10 and a laser wavelength of about 1/4 micron [16]. In both
limits the pellet shell is still thin enough such that it would not implodeproperly if it were subject to the classical Rayleigh-Taylor fluid
instability at the pellet's ablation surface.

2.2. ISI CouplingExperiments

We have now extended our previous laser-plasma coupling
experiments with ISI to include the potentially dangerous Raman
instability.
Using one- micron laser light (to maximize any Raman
effect), at intensities up to 2x10 14 W/cm 2 , we found that the ISI
inhibited the Raman backscatter.

52

2.3. Rayleigh-Taylor

McCRORY et al.

Computer

Simulations

Detailed computer simulations predict that the growth rate of theRayleigh-Taylor mode is well below the classical value [17]. In our
most recent computer simulations, the most dangerous of the shortwavelength perturbations has a growth rate about 60% of the classicalvalue for one-micron laser light, dropping to about 30% of the classicalvalue for quarter-micron laser light.
Quarter-micron laser light istherefore doubly advantageous for the control of the Rayleigh-Taylor
instability: the growth rate is less, and the pellet shells can be thicker.
These hydrodynamic computer simulations use an FCT Euleriancode, with a sliding zone, variable width mesh that has a quartermicron zone size at the ablation surface. Eulerian codes can follow the
Kelvin-Helmholtz roll-up of the spikes, and they can provide very finezoning at the ablation surface. In order to obtain well-defined growth
rates we also now utilize a new "quiet start" method. The 2D code is
run for several nanoseconds without any perturbation, until the initialsound waves decay and the ablation profile reaches a nearly steadystate distribution.
A small 2D perturbation is then inserted at theablation surface. This technique avoids the mixing of Rayleigh-Taylor
phenomena and simple 2D sound waves bouncing off the density
perturbations.
Although these 2D computer simulations are believed to bereliable, they necessarily leave out some of the physical phenomena.
Therefore, there is a need for experimental measurements of theRayleigh-Taylor growth rate with laser-driven ablation. The remainderof this article describes our recent advances in Rayleigh-Taylor
experiments.
2.4. Rayleigh-Taylor

Experiments

There have been several attempts to measure the growth rate of
the Rayleigh-Taylor instability at the ablation surface, but progress has
been very limited. The two best experiments to date, we believe, havebeen at Rutherford-Appleton Lab. and at the Naval Research Lab. [18].
Both used flat targets, not spheres, since spherical convergence effects
are not believed to be important in the early stages of growth of theshorter wavelength perturbations.
The Rutherford-Appleton
experiment used face-on x-ray backlighting with a streak camera, tomeasure the growth of A(pR). The measurements showed that thegrowth rate was well below the classical value, but the signal-to-noise
ratio was small and the error bars large. The Naval Research Lab.
measurement used an internal x-ray face-on source without streaking
the x-rays. The signal-to-noise ratio was better, and the growth of A(pR)
was consistent with a rate below the classical value, but the time
evolution was not measured.

53

IAEA-CN-47/B-I-5

TARGET MOTIONRAYLEIGH-TAYLOR
A{er) GROWTH

FIG. 2. Streaked x-ray radiographs of the motion (left) and of the areal-mass distribution (right) in an
accelerated, perturbed foil. The parameters for both shots were: target thickness 8.5 \im, perturbation
wavelength 100 nm, perturbation amplitude ± 1 pun, irradiance ~ 10u W/cm2. The time increase isin the upward direction. The scales are approximate because the pincushion distortion in the streak
camera intensifier has not yet been removed in these photographs.

We have recently made several technical advances in our
measurement of the ablative Rayleigh-Taylor mode. We used ISI optics
in order to produce a smooth laser profile. The net uniformity ofacceleration was measured by the two-foil method and by shock
breakout.
We used a multinanosecond laser pulse and a smoothnanosecond rise of the laser pulse, in order to prevent strong shockheating or impulsive acceleration.
The rear side temperature wasmeasured to be less than 2 eV during the acceleration.
The
acceleration was measured directly with an x-ray sidelighting source
and an x-ray streak camera. This allowed us to directly determine the"classical" growth rate. And finally, a separate x-ray source was usedwith a separate x-ray streak camera to measure the growth of A(pR).
Although the long-term goal must be measurements with quartermicron laser light, it would be significant to check the computer

54

McCRORY et al.

simulations at any laser wavelength. To date, NRL Rayleigh-Taylor
experiments have been with one-micron laser light, since we have themost laser energy at that wavelength. Initial grooving of the flat foil
target had 50, 100 and 150 micron wavelengths. The depth of thegrooves was typically one to two microns, and the initial thickness of
the CH foil was typically 8 to 15 microns. As of the date of thismanuscript we are still reducing the data. See Figure 2.

ACKNOWLEDGMENTS
The University of Rochester work was supported by the U.S. Department ofEnergy, Office of Inertial Fusion., under agreement No. DE-FC08-85DP40200, and by
the Laser Fusion Feasibility Project at the Laboratory for Laser Energetics, which
has the following sponsors: Empire State Electric Energy Research Corporation,
General Electric Company, New York State Energy Research and Development
Authority, Ontario Hydro, Southern California Edison Company, and the
University of Rochester.
Such support does not imply endorsement of the content
by any of the above parties.
Work at the Naval Research Laboratory was
supported by the U.S. Department of Energy.

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

CRAXTON, R.S., Opt. Commun. 34 (1980) 474.
KESSLER, T., et al., Lasers and Electro-Optics (Tech. Digest Conf. San Francisco, 1986),
Institute of Electrical and Electronics Engineers (1986) 238.
KACENJAR, S., et al., J. Appl. Phys. 56 (1984) 2027.
RICHARDSON, M.C., et al., Rev. Sci. Instrum. 57 (1986) 1737.
RICHARDSON, M.C., et al., Phys. Rev. Lett. 56 (1986) 2048.
GREGORY, G., et al., Proc. Soc. Photo-Opt. Instrum. Eng. 569 (1985) 141.
JAANIMAGI, P.A., et al., Phys. Rev., A. 34 (1986) 1322.
YAAKOBI, B., et al., Rev. Sci. Instrum. 57 (1986) 2124.
SEKA, W., et al., Phys. Fluids 28 (1985) 2570.
MALONE, R.C., McCRORY, R.L., MORSE, R.L., Phys. Rev. Lett. 34 (1975) 721.
KMS Fusion, Inc., Ann Arbor, MI (provision of glass shell targets).
SCHMITT, A.J., GARDNER, J.H., J. Appl. Phys. 60 (1986) 6.
LEHMBERG, R.H., GOLDHAR, J., Fusion Technol. (Jan. 1987).
OBENSCHAIN, S.P., et al., Phys. Rev. Lett. 56 (1986) 2807.
AFANAS'EV, Y.V., et al., JETP Lett. 21 (1975) 68.
GARDNER, J.H., BODNER, S.E., Phys. Fluids 29 (1986) 2672.
EMERY, M.H., GARDNER, J.H., BODNER, S.E., Phys. Rev. Lett. 57 (1986) 703.
COLE, A.J., et al., Nature (London) 299 (1982) 329;
GRUN, J., et al., Phys. Rev. Lett. 53 (1984) 1352.

IAEA-CN-47/B-II-l

PROGRESS IN INERTIAL CONFINEMENT
IN THE UNITED KINGDOM
M.G. HAINES, D. BASSETT, A.R. BELL, E.M. EPPERLEIN,
T.H. KHO, D. RILEY, S.D. TABATABAEI, O. WILLIThe Blackett Laboratory,
Imperial College of Science and Technology,
London, United Kingdom
J. CORBETT, M.J. LAMB, C.L.S. LEWIS, E.W. ROBERTSON
Department of Pure and Applied Physics,
The Queen's University of Belfast,
Belfast, United Kingdom
P.M. EVANS, A.P. FEWS
H.H. Wills Physics Laboratory,
University of Bristol,
Bristol, United Kingdom
C. CHENAIS-POPOVICS
Ecole poly technique,
Palaiseau, France
M.J. HENSHAW, G.J. PERT
Department of Applied Physics,
University of Hull,
Hull, United Kingdom
P. APTE, A.J. COLE, S.J. ROSE, W.T. TONER
Rutherford Appleton Laboratory,
Chilton, Didcot, Oxfordshire,
United Kingdom
Abstract
PROGRESS IN INERTIAL CONFINEMENT IN THE UNITED KINGDOM.
Eight experimental and theoretical projects are presented, covering important physics aspects ofinertial confinement. An experiment on energy transport employed green laser light on massive plastic
spheres with buried diagnostic substrates that emit X-rays at different temperatures. Consistent withtheory there was no evidence of a foot in the non-linear heat flow. Time and space resolved density
profiles were obtained from the Stark spectroscopy of aluminium tracer dots in plastic targets. Whentightly focussed laser beams are employed, megagauss magnetic fields are generated by VT x Vn
effects and measured by Faraday rotation. Theoretically, the convection and amplification of magnetic
fields by inward heat flow has been extended to non-linear transport by solving the Fokker-Planck equation. Implosion studies employing X-ray back-lighting and spectral emission techniques have shown theinfluence of variation of the aspect ratio ro/Ar of polymer shell targets in the range 10—60 driven ablatively by twelve beams (0.53 /xm, 0.8 ns) at an absorbed irradiance of 1014 W -cm"2. Rayleigh-Taylor

55

56

HAINES et al.

instabilities in accelerated plane foils have been studied by a novel technique of alpha particle transmission. Non-linear simulations of this instability in spherical geometry show the phenomenon of
coalescence of bubbles, while no Kelvin-Helmholtz mushrooming of the spike tips is observed. Finescale jets have been diagnosed by a Schlieren technique on spherical targets, showing that the jets occur
towards the end of the laser pulse. A non-local treatment of the heat flow driven collisional Weibel
instability has shown that inclusion of the Nernst convection of magnetic fields leads to stabilisation.

1.

INTRODUCTION

This paper describes experimental and theoretical physics
carried out by universities in the U.K., in collaboration with
the Central Laser Facility, Rutherford Appleton Laboratory, inthe field of inertia! confinement. Areas covered include
energy transport, magnetic field generation and convection,
implosions, and Rayleigh-Taylor and thermal instabilities.
2.

ENERGY TRANSPORT

Extensive experimental observations have been carried out
to investigate radial thermal transport in solid sphericaltargets. An aluminium tracer layer (0.025 to O.lym thickness)
was sandwiched between 0.5 to 3ym layers of CH~ which were
deposited on the surface of a silicon sphere doped withcalcium.

Si Ly -_

10

.

Ally

Si

/n

\ f!1
/
1

Ul
tIll

1

z

0-5

1

1

-10

SH0T6 20/5/86

10
TlME(ns)

FIG. 1. Temporal emission intensities of Al Lya, Si Lya and 2Ca Hea. The time-scale is with respect
to the peak emissions from the Al Lya line.

57

IAEA-CN-47/B-n-l

(a)
2.0

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# 16050686

Experiment" " I Theory

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TIME FIDUCIAL

1ns

-25

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+25 mA

FIG. 2. (a) Experimental spectral profile of the Hey line at the peak of the laser pulse compared withtheoretical modelling for electron densities of 5 X JO21 cm'3 and 6 X 1021 cm'3, (b) Raw data,
including the time fiducial.

The 12-beam Vulcan laser was used to irradiate these 155ym
diameter targets with green light (X = 0.53um) at an irradianceof 2.3 x 1014 w-cm-2 and 800 ps duration. An X-ray streak
camera recorded the temporal emissions of the Al, Si and Calines, and results are shown in Figure 1. (An auxiliary laser
beam of 70 ps duration was used to produce an X-ray flash for atime fiducial.) Because of the simultaneous turn-on of the Si

58

HAINES et al.

and Ca X-ray emission, it can be concluded that the heat front
is steep and no 'foot', or precursor, in the electron energy
spectrum between 300 and 600 eV is present, which is consistent
with theoretical modelling [i].
Time resolved density profiles were obtained by means ofStark spectroscopy. The prime diagnostic was a toroidally
focusing PET crystal mounted on to a X-ray streak camera.
Solid plastic spherical targets had Al tracer dots (75ym indiameter and O.lutn in thickness) buried from 0.2 to 2ym below
a plastic layer. Under the same conditions of laser irradiation
the time resolved X-ray spectrum from helium-like y-transitionof aluminium was recorded. The measured Stark profile at thepeak of the pulse is shown in Figure 2. It was fitted to anumerical profile [2]2 to give3 a value for electron density
between 5 and 6 x 10 l cm " . (The measured Stark width of
40 mA dominates instrumental source. Doppler and opacity
broadening of 2 mA'.) A simultaneous spatially resolved, time
integrated spectrum from the side showed that peak emission was
at 25pm outside the initial location of the target surface.

3.

MAGNETIC FIELDS

Simultaneous measurements of mass ablation rates, density
profiles and vT x vn generated magnetic fields were made on
spherical targets irradiated with 12 tightly focussed infra-red(>, = 1.05ym) laser beams. Each beam had an irradiance of 3 x
10"15 W-cm"S focal spot size 40ym, energy 35J and duration 1.2ns.
For interferometry, a 3 wo (A = 0.35ym) probe beam and for
Faraday rotation a Raman-shifted second harmonic probe beam wereused in a multiframe mode and in conjunction with a time-resolved
X-ray spectrometer. Magnetic fields in the megagauss range havebeen measured.
Such magnetic fields could be convected towards the ablation
surface through the Nernst effect [3]. By solving the FokkerPlanck and Maxwell equations in 1| dimensions it has beenshown [4]that this convection of magnetic field is strongly
coupled to the non-linear heat flux q . To a good approximation,
the convective velocity is qe/(j> P e ) where p is the electronpressure. The Righi-Leduc heat flow is strongly reduced underconditions of a depleted tail in the distribution function, as
occurs near the critical density. The transverse heat flow issensitive to the depletion because of the difference of two terms
which depend on high velocity moments.

59

IAEA-CN-47/B-IM

0.01

0

10

20 30 40
radius (um)05001000 1500 2000
time (ps)FIG. 3. Radiograph with spectrally resolved Au backlighting at a time near peak compression, together
with the transmission profile (o indicate experimental data, — and — are simulations separated by
100 ps) and the r(l/2) simulation and experimental data point *.

60

HAINES et al.

2.4

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FIG. 4. Comparison with simulation for low aspect ratio shells. The solid circle represents a systemcalibration shot for an unirradiated target.

IAEA-CN-47/B-II-l

61

4.

INFLUENCE OF ASPECT RATIO ON IMPLODING TARGETS
Experiments have been carried out using both X-ray backlighting and spectral emission techniques to study the
influence of aspect ratio in the range ro/Ar = 10-60 on empty
CH polymer targets driven ablatively by 12 beams (x = 0.53ym,
duration 0.8ns) at absorbed irradiance levels of about 10^4 w•cm"2. Quasi-point projection radiography with Bragg crystal
and multilayer X-ray mirror dispersers were used to record 2-D
spatially resolved shadows of imploding low aspect ratio CH
shells with frame times of order 100 ps. At early times theradiograph with spectrally resolved Au back-lighting at about
2.5 keV showed well-behaved shell structure. Figure 3 is aradiograph near peak compression,at which time (marked *) theshell has collapsed to about 20% of its initial radius. The
conditions for a given shot are simulated by the 1-D hydro-code
Medusa, and density profiles at the time of the data record are
post-processed to find the radial distance r{l) Med to a chord
giving 50% transmission at the probe energy. In Fig 4, thecorresponding experimentally measured chord radius (r (|) exp) is
normalised to this and plotted against the ratio T(probe)/T(imp),
where T(probe) is the experimental delay time of the X-ray probefrom the start of the drive pulse and T(imp) is the Medusa
simulation time for peak compression. Perfect agreement withsimulation would place all data points along the horizontal
dotted line. The discrepancies can be explained in terms ofa random fluctuation in implosion time of the order of ±200 ps
compared to simulation, due to lack of beam balance, beam
pointing and other shot-to-shot variations. All these targets
had aspect ratios ro/Ar in the range 7 to 12, and from Abelinversions of opacity profiles gave peak densities of the orderof 10 g-cm-3.
Since the low-aspect-ratio targets seemed prone to shot-toshot fluctuations, it was of interest to see whether this sensitivity became catastrophic at higher aspect ratios. 3 A crudeself-regulating hydrodynamic model suggests that if ro Ar is
kept constant, then targets of any aspect ratio ro/Ar will
implode in approximately the same time for fixed absorbed laser
energy and pulse length. In the experiment, the aspect ratio
was accordingly varied from 10 to 60» and Medusa simulationsindeed showed a weak variation of the implosion time for fixedro^Ar. Two time resolving crystal spectrometers measured thecarbon continuum emission from the cores at 0.6 and 2 keV.
Figure 5 shows the measured implosion times with respect to thelaser pulse peak, with shots normalised to an absorbed energyof 80J. However, throughout the range of ro/Ar the discrepancies remain of the order of 200 ps. It can be concluded thathigher aspect ratio targets behave in this respect just as well
as low aspect ratio targets.

HAINES et al.

62

400MEDUSA

S

200

UJ

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to
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a.

0

-200-

LASER

-400

—r20

40

60

80

ASPECT RATIO, r/dr

FIG. 5. Implosion times relative to the peak of the drive pulse as a junction of aspect ratio. The
MEDUSA simulation corresponds to 20% coupling of 400 J incident energy in a 0.8 ns FWHM pulse
which was typical for the focussing conditions used.

80.0

60.0

40.0

20.0

0.0

0. 0

_i_

3. 0

_i_

6. 0
9. 0
12- 0
RANGE LOSS IN TARGET (pml

S. 0

FIG. 6. Alpha particle count versus range loss computed from CR-39 tracks in the accelerated target.

IAEA-CN-47/B-II-l

5.

63

RAYLEIGH-TAYLOR EXPERIMENTS

A novel technique for investigating the non-linear development of Rayleigh-Taylor instabilities is reported. A planar
foil is accelerated by irradiating from one side with two laser
beams, and it is backlit by thermonuclear alpha particles froman adjacent implosion of a deuterium-tritium filled microballoon,
timed in a controlled way with respect to the acceleration of
the foil. After transmission through the foil, the alphaparticles are recorded in CR-39 plastic. The thickness distribution across the foil can be directly determined from the energy
spectrum of the alpha particles found from a high resolution
measurement of the etched tracks. Multiple Coulomb scatteringin the foil limits the spatial mapping capability of thistechnique to about 40ym.
Experiments have been conducted either with corrugated planetargets (corrugation length of 10 to 40pm) or with spatially modulated laser beams (modulation length 20 to 60ym at the foil).
The alpha particle source was a lOOym diameter DT filled glass
microballoon with wall thickness lym irradiated by six laser beamsof lOOps duration. Irradiation by another two longer pulse beams(Ins) caused a foil to be accelerated towards the exploding
pusher, initially 1mm away. The relative timing of the pulseof alpha particles and the start of the acceleration of the
foil was varied by up to 600ps 5 so that different stages ofgrowth of the instability in the foil could be investigated.
Figure 6 shows the thickness distribution of an Al foil
irradiated with 1.05ym light at an intensity of 1014 W-cnr 2 ,with the timing set so that the alpha particles passed throughthe foil l-9ns after the start of the accelerating pulse. The
foil was corrugated on a transverse scale of 10pm to act as an
initial source of perturbation. The presence of some tracks
having essentially no energy loss suggests that the foil hasburnt through in places.
6.

RAYLEIGH-TAYLOR SIMULATIONS

2-D computer simulations of the Rayleigh-Taylor instabilityof the outer surface of a spherical CH~ shell target have been
carried out using the code POLLUX [5]. This used a two-temperature model with flux-limited electron thermal conduction (f =0.06), while the laser energy was absorbed by inverse bremsstrahlung, with a dump of 20% of the energy reaching criticalto simulate anomalous processes. Figure 7 shows density contour
plots for a shell of initial radius 150ym, aspect ratio 30,
irradiated by 0.53ym laser light with an intensity of 4 x 10'^W-cm~2. A sinusoidal perturbation of initial amplitude 0.2ym

64

HAINES et al.

FIG. 7. The non-linear growth of Rayleigh-Taylor instabilities are shown in density contour plots at
0.5, 0.75, 1.0 and 1.2 ns, where the shaded regions correspond to p > 1.2 g-cm'3 (solid density).

and wavelength 15ym was applied to the outer interface. While
the spikes tend to grow according to a gt^ law (g being
acceleration) the bubble head shows a period of constant ?
velocity followed by an amplitude that also varies as ngt .
An extensive series of simulations has shown thatr\ lies between
0.05 and 0.15 and that a linear growth occurs only if theinitial perturbation amplitude is less than 0.04 times the
wavelength. The lifetime of the shell is determined by thisrate of penetration of the bubbles. Competition can occur
between bubbles and can result in merging. As shown in
Fig.7, no Kelvin-Helmholtz mushrooming of the spikes due tovorticity effects is evident.

65

IAEA-CN-47/B-II-l

1630ps1660ps1690ps1750ps1880ps1910ps

19A0ps2000psFIG. 8. Two sequences ofSchlieren photographs of gold-coated microballoons, illustrating the growthofjei-like structures.

66

7.

HAINES et al.

THERMAL INSTABILITIES

Spherical gold-coated targets were irradiated with 270 J
from 12 infra-red (A = 1.05pm) laser beams with a pulse
duration of 1.8 ns and D/R = 10, where D is the distance fromthe target centre to the focus of each lens. Fine-scale jetlike structures develop and their temporal evolution wasrecorded with a four-frame optical probing system with exposure
times of 15 ps. Two sequences of Schlieren images are shownin Fig.8. It is interesting, but not yet explained, that thejets begin to grow near the end of and after the laser pulse.

8.

COLLISIONAL WEIBEL INSTABILITY

A non-local theoretical analysis of the collisional
Weibel instability has been carried out for a laser-ablated
plasma in planar geometry. Earlier work [6]had shown, in a
local analysis using a distribution function derived from atime-evolving solution of the Fokker-Planck equation, that thecollisional Weibel instability
driven by heat flow had agrowth rate typically of 1 0 1 0 s-1 and a wavelength of 20 pm(similar to the observed'thermal instabilities' discussed above)
The present analysis employed a simpler distribution function
derived from linear heat flow, but in a non-local analysis
which included longitudinal magnetic diffusion, electron
viscosity and magnetic convection effects £3]. The exponential
time dependence of the linear Weibel eigenmodes was obtainedby numerically solving the electron Fokker-Planck andMaxwell equations in the inhomogeneous direction with aharmonic perturbation in the transverse direction. The main
result was that the collisional Weibel mode was stabilised by
the Nernst convection of magnetic field to the cold, dense
inner region where the perturbation was damped.

ACKNOWLEDGEMENTS
The authors thank the staff of the Central Laser Facility for their assistance inthe experiments. They are also grateful to the Centre Europeen de calcul atomiqueet moleculaire, Universite de Paris Sud, Orsay, France, for providing useful workshops for theoretical investigations.

IAEA-CN-47/B-H-1

67

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

BELL, A.R., Phys. Fluids 28 (1985) 2007.
LEE, R.W., personal communication.
NISHIGUCHI, A., YABE, T., HAINES, M.G., Phys. Fluids 28 (1985) 3683.
KHO, T.H., HAINES, M.G., Phys. Fluids 29 (1986) 2665.
PERT, G.J., J. Comput. Phys. 43 (1981) 111.
EPPERLEIN, E.M., KHO, T.H., HAINES, M.G., Plasma Phys. 28 (1986) 393.

DISCUSSION
R.L. McCRORY: In the high aspect ratio shell experiments performed at the
Rutherford Appleton Laboratory, is it possible that radiative preheat decreases the inflight aspect ratio of the target as the initial aspect ratio increases so that the agreement with the one-dimensional MEDUSA simulations is not unexpected but does nottell us about ablative dynamics of high aspect ratio shells?
M.G. HAINES: Radiative preheat calculations have been carried out for low
aspect ratio CH shells with a wall thickness of > 5 /xm. Here the radiative preheat
was negligible. For the shell with the thinnest wall (about 2.2 fim) used in the experiments it is possible that radiative preheat was not negligible, but this would be offset
by the lower irradiance used in the r'Ar scaling employed.

IAEA-CN-47/B-II-2

INERTIAL CONFINEMENT FUSION PROGRAM
AT LOS ALAMOS NATIONAL LABORATORY
D.C. CARTWRIGHT, P.D. GOLDSTONE, J.A. HANLON,
A. HAUER, R. MAH, J. McLEOD, W.C. MEAD,
L.A. ROSOCHA, D.C. WILSON, C.A. WINGATE
Los Alamos National Laboratory,
Los Alamos, New Mexico,
United States of America

Abstract
INERTIAL CONFINEMENT FUSION PROGRAM AT LOS ALAMOS NATIONALLABORATORY.
Los Alamos National Laboratory has completed a thorough evaluation of the CO2 laser as adriver for inertial confinement fusion (ICF) and concludes that in the case of a large driver the cost of
this laser is too high to be attractive. Recent studies, at Los Alamos and elsewhere, show that 0.25 /*m
is near the optimum coupling wavelength, and Los Alamos is involved in experimental and theoreticalstudies of laser-matter interaction near that wavelength. Since KrF lases at 0.25 ftm and has thepotential for 10% wall plug efficiency and relatively low capital cost, Los Alamos investigates thefeasibility of the KrF laser as a future ICF driver.

1.

INTRODUCTION

The ICF program at Los Alamos is designed to determine
the feasibility of achieving fusion in the laboratory and iscomposed of four major components: (£) driver-matter interaction, (it) materials technology, (Hi-) development of
improved drivers and (iv) the Centurion program.
The Centurion program at Los Alamos is a "classified
theoretical and experimental effort to investigate the design
characteristics of efficient ICF targets" and, as was recently noted by the US-NAS [1], the Centurion (Los Alamos)/Halite (Livermore) programs have been "quite successful so
far ". The Centurion program will not be discussed further
here.

2.

DRIVER-MATTER INTERACTION
A.

.C02_

The main thrust of fusion laser development at Los
Alamos has, until recently, been directed toward C 0 2 laser
systems at a wavelength of 10.6 jum. The goal of Antares was

69

70

CARTWRIGHT et al.
1000

10
100
DRIVER ENERGY (MJ)
FIG. 1. Comparison of required and predicted ICF target gains,for an assumed reactor condition of r\G - 10.
The theoretical curves are based on results ofJ.D. Lindl and J. W. K. Mark
(Laser Part. Beams 3 (1985) 37).

2

3
4
TIME (ns)FIG. 2. Hot-electron temperature (inferred from hard x-ray bremsstrahlung) versus time. The insert
shows a typical Antares pulse shape, stretched to 5.5 ns FWHM for this experiment. The hot-electron
temperature declines somewhat at later times, though not enough to modify the unfavorable targetcoupling characteristics of CO2 laser radiation.

to complete evaluation of the C0 2 laser as a laser fusion
driver by extension of the Helios data to higher energy(-40 kJ) and longer pulse lengths (1-5 ns).
The most important factor in C0 2 laser-plasma couplingis the collisionless nature of the target corona for thislong wavelength. This results in the dominance of collective absorption and scattering mechanisms, limiting theachievable laser absorption efficiency and producing highly
energetic "suprathermal" electron distributions [2-4]. Theseenergetic electrons tend to couple their energy into a small

IAEA-CN-47/B-II-2

71

number of high-energy ions, which are an inefficient source
for the required implosion driving pressure.
The dominance of suprathermal processes led to the needfor innovative approaches to C0 2 fusion target design in order to effectively use the hot electron energy. We separated
the problem into studies of: (1) fuel capsule performance
under idealized drive conditions, and (2) laser-target coupling efficiency. Capsule design calculations defined thecapsule gain for a given amount of energy absorbed into thetarget ablator. Based on the capsule gain predictions andthe assumed driver electrical efficiency r\, we determined thetarget coupling efficiency X required for fusion
applications. For drivers with r) < 10%, X > 0.25 is required
to achieve adequate overall target gain (thermonuclear
output/driver input). For reasons of cost and containment,
adequate gain should be available using driver energies lessthan - 10 MJ. Figure 1 shows the required and predicted target gains for various driver energies and X-.factors.
The Los Alamos C0 2 evaluation effort has determined theX-factors achievable from a variety of C0 2 target designs atreactor scale, including both directly- and indirectly-driven
concepts, but we have been unable to find C0 2 target concepts
which will achieve adequate overall coupling efficiency.
Recent Antares data (Figure 2) indicate that some significant
mitigation of hot-electron production was possible at morereactor-like pulselengths [5]), but this improvement appears
to be inadequate to make X for C0 2 competitive with shortwavelength lasers. However, a comparable analysis of shortwavelength (KrF) lasers indicates good potential (Figure 1 ) .

B.

Short Wavelength Theory and Experiments

Los Alamos has increasingly shifted its emphasis to ICF
driven by Short-Wavelength Lasers (SWL), particularly KrF,
and continued design and theory work for light- and heavy-ion
drivers. The favorable coupling characteristics of SWL light
had been generally established [6] and the Los Alamos shortwavelength effort now includes work in laser-plasma coupling
theory, target design, and the design, execution, andanalysis of SWL interaction experiments.
Hot-electron production is still of some concern, though
no longer dominant, for SWL-driven reactor targets since some
important high-gain designs have fairly low preheat
tolerances. Instead of the steep, collisionless plasmas of
the C0 2 target corona, with its associated nonlinear collective instabilities, coupling theory for KrF light must deal
with collisionally-dominated plasmas in which the plasma
instabilities are weakly driven (near threshold), and the72

CARTWRIGHT et al.

growth occurs in relatively homogeneous regions extending
over thousands of laser wavelengths. Simulations of such
SWL-produced plasmas at Los Alamos have indicated two classesof plasma processes which might lead to low-level productionof suprathermal electrons in ICF reactor targets: (1) turbulence and the coupling of low- and high-frequency instabilities and (2) laser light thermal self-focusing. Work continues to determine the impact of such processes on ICF
target designs.
Conservative designs using low-aspect-ratio shells
similar to those advanced by Llvermore [7] have been tailored
to KrF applications at gains of 100-200 with input energiesof 2-5 MJ. We are also exploring advanced designs which
could offer substantial increases in target gain.
Los Alamos has been increasingly involved in the design,
fielding, and analysis of SWL Laser-Matter Interaction
experiments. Several other ICF laboratories have supported
our collaborative involvement, and we have been able to contribute to experiments studying laser-plasma coupling and xray conversion scaling to 0.26 /im wavelength (Lawrence
Livermore National Laboratory), high-Z plasma dynamics(University of Rochester/Laboratory for Laser Energetics[LLE]), and electron thermal transport and ablation (withRutherford-Appleton Laboratory).
The Los Alamos/LLE High-Z Plasma Dynamics Experiment [8]at 0.35 fim wavelength explored the mechanisms governing theconversion of laser light to x-rays, especially reexamining a
low-intensity discrepancy noted in the analysis of gold disk
experiments at Livermore [6]. Targets were solid CH sphereswith gold coatings of various thicknesses to probe the energy
transport, hydrodynamics, and x-ray emission of high-Z
plasmas. The measurements were performed using six and 24
beams of the OMEGA spherical irradiation facility, and showed
unexpected scaling differences between the two irradiation
sets, Fig. 3a. We also determined the x-ray emission as afunction of gold coating thickness, Fig. 3b, and observed thesignatures of laser burn-through, verifying the LASNEX
predictions for energy-penetration and x-ray-emission depths.
To take advantage of the unique opportunity provided by
the AURORA proof-of-principle KrF laser, we are designing
experiments at 5 kJ, 5 ns, to begin in late 1987. We will
study laser-target coupling effects dependent upon special
properties of the KrF laser system: beam quality, bandwidth,and Induced Spatial Incoherence [9]. Reactor-like plasmas
extending over thousands of wavelengths can be produced by
exploding thin foils, Fig. 4. Other areas to be emphasized
include x-ray conversion processes, hohlraum scaling andphysics, and planar-foil studies of fluid instabilities.

73

IAEA-CN-47/B-H-2(a)1.0

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a

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INTENSITY (W/cm*)

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1.0

10.0

GOLD THICKNESS (fim)FIG. 3. Results from the high-Z plasma dynamics experiment.
(a) X-ray conversion and apparent target-size scaling at low intensities.
(b) Conversion efficiency fall-off for thin gold coatings, indicating depth of energy penetration
into the high-Z plasma, in agreement with calculations.

w 1000v>

100a.
o{APPROXIMATE THRESHOLDS)

0.01

01

1
\\\

10

1000

<IO U W/cm 2 -/Ltm 2 )FIG. 4. Plasma parameter space. Plasma size (L/\^), interaction intensities (Fhl) and temperatures(in keV, indicated by numerals on the figure) accessible using the AURORA laser are highly relevantto the reactor target regime. Thresholds for plasma instabilities, shown by the diagonal lines, indicate
roughly the regions where filamentation, stimulated Raman scattering (backward and forward) andstimulated Brillouin scattering are expected to be operative.

74
3.

CARTWRIGHT et al.
MATERIALS TECHNOLOGY

Fusion targets, whether for use with laser or particle
beam drivers, consist of a series of concentric sphericalshells (pusher and ablator shells). The targets are small,less than a millimeter in diameter, and each shell must possess specific properties, ranging from strength at low
densities to high atomic number and density for improved
compression. Considerable success has been achieved in target fabrication, developing new technologies, new materials,
and new tools. Metal pushers can be deposited on nonconducting substrates by autocatalysis or "electroless" plating.
Some advanced designs use plastic or glass foam ablators
saturated with liquid DT. We have prepared small-cell, lowdensity, rigid foams from such polymers as polystyrene,poly(4-methyl-l-penetene), tradenamed TPX, and substituted
celluloses and starches. In some cases, specially developed
micromachining techniques can be used to improve the dimensional properties of the foam material.
The multifaceted task of target fabrication has led tomaterials and technologies of interest and benefit to otherresearch and development programs at Los Alamos. Some examples are• The polymeric foams (25 /jm cells) which are of benefit
to biomedical research.
• Miniature coaxial cables (0.1 mm dia.) fabricated insupport of ICF experiments which may simplify surgicalremoval of brain tumors.
• High-strength single-crystal whiskers, which were
fabricated and studied for possible use as support
structures in laser fusion targets, are now a key component in the Los Alamos structural ceramics program.

4.

DEVELOPMENT OF IMPROVED DRIVERS

Because KrF lases at 0.25 p,m (nearly the optimum coupling wavelength), and has the potential for 10 per cent wall
plug efficiency and low capital cost, Los Alamos is investigating the feasibility of the KrF laser as a future ICF
driver. At present, Los Alamos is constructing a prototype
KrF laser system called AURORA [10], which is a short-pulse
high-power krypton-fluoride laser system that serves as an
end-to-end technology demonstration of the applicability oflarge-scale ultraviolet (uv) laser systems to ICF
applications. AURORA is designed to employ optical angularmultiplexing and serial amplification by electron-beam-driven
KrF laser amplifiers to deliver 5-ns duration multi-kilojoule

75

IAEA-CN-47/B-H-2

1-5ns-250mJ-

FRONT END

ANGLE ENCODER,.50
96-*.nS'^

FOLD ENCODER

8 FOLD ENCODER50
PUPIL
INPUT PUPILARRAYFOCAL PLANE
96-5ns-10mJ
IN VACUUM

LAM INPUTARRAYPA
FOCAL PLANE
S6-5ns-500 mJ
IN VACUUM

LAM^ C-jr-lT-

1)6-5 n?200 J

FIG. 5. Conceptual layout of the AURORA laser system, showing the main optical and laser elementsfrom the front-end to the final amplifier output. Stage gains, number of beams and energy per beam areindicated at various points along the beam path. A final optimized design output of 10-20 kJ in a 480 ns
pulse composed of a 96 element train of S ns pulses is expected at the output of the final Large Aperture
Module (LAM) amplifier.

LARGE APERTURE MODULE

DECODER BEAM TUNNELLAM FEED ARRAY,
96-CHANNELMIRROR ARRAY
12 F0LD-RECOLLIMA

ENCODERSAM AMPLIFIER

FIG. 6. An artist's conception of the completed A URORA system, with the 48 beam decoder and target
building in place. As configured, the system will deliver multi-kilojoule 5 ns laser pulses to fusion
targets. The decoder and target chamber are currently in fabrication.

76

CARTWRIGHT et al.

laser pulses to ICF targets. Figure 5 shows a conceptual
layout of the multiplexed AURORA laser chain and illustrates
all of the main optical and laser elements from the front-end
to the final amplifier output.
The main amplification chain for AURORA consists of fourelectron beam-driven KrF laser amplifier ranging in aperturesize from 10 cm x 1 cm to 100 cm x 100 cm. These devices are
specified as the Small Aperture Module (SAM), thePreamplifier (PA), the Intermediate Amplifier (IA), and theLarge Aperture Module (LAM).
The AURORA design replicates the front-end output pulse
using aperture slicers, beamsplitters, and mirrors to produce
a 480-ns long pulse train consisting of 96 beams each of 5 ns
duration. These encoded pulses, which are spatiallyseparated, are then separately aligned at the entrance pupilof an optical relay system. The beam train is then relayed
through two single-pass laser amplifiers (the PA and IA), anda double-pass laser amplifier (the LAM). Beam train alignment is accomplished with two automated alignment systems[11]; one system points each of the 96 beams through thePreamplifier entrance pupil and the second maintains alignment of the LAM primary mirror.
The AURORA project is divided into two phases: (1)operate the laser system to produce an optically-encodedenergy output of 10 kJ in a 480 ns pulse at the final
amplifier, and (2) decode 48 of the 96 beams and deliver
multi-kilojoule 5 ns pulses to fusion targets. When bothgoals have been reached, AURORA will have demonstrated thesystem technology and enable critical tests of KrF-related
target physics issues. Figure 6 shows an artist's conceptionof the entire AURORA system as it will appear with the 48beam demultiplexer and target area in place.
Integration of the entire amplifier chain is due to takeplace late in 1986 and the decoder and target subsystems
scheduled to be completed in 1987. In addition to AURORA,
Los Alamos is addressing the technology issues which affectthe system cost and wall-plug efficiency of large-scale(megajoule) KrF laser fusion systems. Together with theresults from AURORA, this should enable an assessment of the
KrF technology as a candidate fusion driver in the approximate 1991 time frame.

REFERENCES[1] Review of the Department of Energy's Inertia] Confinement Fusion Program, National Academyof Sciences, Washington, DC (1986).
[2] HAUER, A., et al., in Laser Interaction and Related Plasma Phenomena, Vol. 6, Plenum Press,
New York (1984) 479.

IAEA-CN-47/B-n-2
[3]
[4]
[5]
[6]
[7]
[8]
[9]
[10]
[11]

77

FORSLUND, D.W., GOLDSTONE, P.D., Los Alamos Science No. 12, 2 (Spring/
Summer 1985).
BACH, D.R., et al., Phys. Rev. Lett. SO (1983) 2082.
MUELLER, M.M., et al., Bull. Am. Phys. Soc. 30 (1985) 1365.
MEAD, W.C., CAMPBELL, E.M., ESTABROOK, K.G., TURNER, R.E., et al., Phys. Fluids
26 (1983) 2316 (and references therein).
LINDL, J.D., persona] communication, 1983.
GOLDSTONE, P.D., GOLDMAN, S.R., MEAD, W.C., RICHARDSON, M.C., et al.,Rep. LA-UR-86-2856, Los Alamos Scientific Lab., NM (1986); submitted to Phys. Rev. Lett.
LEHMBERG, R.H., OBENSHAIN, S.P., Opt. Commun. 46 (1983) 27.
ROSOCHA, L.A., BOWLING, P.S., BURROWS, M.D., KANG, M., HANLON, J.,
McLEOD, J., YORK, G.W., Jr., Laser Part. Beams 4 (1986) 55.
KORTEGAARD, B.L., 1985 Proc. Soc. Photo-Opt. Instrum. Eng. 534 (1985) 159.

DISCUSSION
H. HORA: You mentioned that your KrF laser would use multiplexing irradiation. Does this mean that you plan to use Yamanaka volume compression and volumeignition1 rather than central spark ignition of pellets?
L.A. ROSOCHA: No, we are interested in spark ignition, too, but AURORA
will mainly concern planar target irradiation, with 48 beams, and hohlraum scaling.

HORA, H., RAY, P.S., Z. Naturforsch., A 33 (1978) 890.

IAEA-CN-47/B-II-3

EXPERIMENTAL STUDIES ON INERTIAL FUSION
AT 0.26 /an LASER WAVELENGTH
F. AMIRANOFF, R. FABBRO, E. FABRE, B. FARAL,
C. LABAUNE, A. MICHARD, J. VIRMONT*
Laboratoire de physique des milieux ionises,
Ecole poly technique,
Palaiseau, France
P. MORA, J.F. LUCIANI, R. PELLAT, D. PESME*
Centre de physique theorique,
Ecole polytechnique,
Palaiseau, France
J. BRIAND*
Centre de physique atomique,
Universite de Toulouse,
Toulouse, France
F. COTTET, M. HALLOUIN, J.P. ROMAIN*
Laboratoire d'energetique et detonique ENSMA,
Poitiers, France
H. BALDISNational Research Council,
Ottawa, Canada
A.P. FEWS
University of Bristol, Bristol,and Central Laser Facility,
Rutherford Appleton Laboratory,
Chilton, Didcot, Oxfordshire,
United Kingdom

Abstract
EXPERIMENTAL STUDIES ON INERTIAL FUSION AT 0.26 fim LASER WAVELENGTH.
Data are presented on the experimental analysis, at 0.26 ^ m , of the behaviour of Raman
instability in a homogeneous plasma, on the generation of high ablation pressures (50 Mbar),and on the multiplication by impedance mismatch and foil collision, up to 400 Mbar, as well
as some preliminary results on the implosion of D-T targets, yielding 1.5 X 10 8 neutrons, andthe behaviour of implosion with target parameters.

Groups participating in the scientific programme of GRECO Interaction Laser-Matiere,of the Centre National de la Recherche Scientifique.

79

80

AMIRANOFF et al.

Short wavelength lasers (SWL) have been demonstrated to be essential for directdrive laser inertial fusion. Experiments have evidenced [ 1 ] efficient coliisional absorption, reduction of fast electron generation by resonance absorption and thecapability of generating high ablation pressures when short wavelength lasers areused. However, there are still fundamental problems to be analysed so as to provide a good understanding of the interaction, transport and hydrodynamic physics
necessary to achieve good compression.
First of all, the advent of large lasers gives rise to new situations where interaction takes place with large scalelength plasmas and processes such as parametricinstability find favourable circumstances for their growth. Secondly, energy transport by electron conduction has been observed to be inhibited for al! wavelengths,
and a description of this problem is necessary for a good numerical modelling of
the experiments. Third, high pressures are needed to bring about efficient compressions. They can be generated by ablation and amplified by techniques as
shock impedance mismatch and foil collisions. However, two-dimensional effects
or radiation preheat can introduce significant perturbation in the behaviour ofhydrodynamics. Finally, the drive of efficient ablation compression has to bestudied by experiments in order to analyse the behaviour of hydrodynamic motion
and the instability associated with non-uniform illumination. In this paper, we shalldescribe some work done in our group on these topics.

1.

EXPERIMENTAL FACILITIESThe laser is a neodymium phosphate glass system, the final amplifying heads
being 90 mm glass rods. The laser is structured as a two-beam system used for oneor two-beam plane target experiments and followed by a six-beam amplifier for
spherical target illumination. Each beam can deliver, in short pulses, a peak powerof 250 GW and, in long pulses, i.e. of 600 ps and above, an energy of 120 J. The
peak power of the six-beam system is 1.5 TW with an energy of 700 J in 600 ps.
In all the experiments the beam frequency is converted by KDP crystals. The conversion cells are located just before the entrance window of the experimental
chamber: they allow frequency doubling with 65% efficiency and frequency
quadrupling with 45% efficiency under the best conditions. However, damage
problem introduce some limiting operating conditions, and in the six beam experiments the best output has been 500 J in a 600 ps pulse at 1.06 /im, and 200 J in
a 500 ps pulse at 0.26 fjm. Moreover, uncoated optics for 0.26 jum reduce theenergy on targets by nearly 30%. Nevertheless, focusing on targets with fast UV
optics gives, on a plane target, an intensity in the range of 2 X 1015 W-citf2, and,on a spherical target, an intensity of 5 X 1014 W-cm"2 at 0.26 pm wavelength.

81

IAEA-CN-47/B-H-3

300

200 "

P

100 -

360

0.07

380
400
420 440 460
Raman Light Wavelength (nm)0.1
0.15
Emitting Density (n/nc)

480

360

0.2

0.06

300
400
420 440 4G0 480
Raman Light Wavelength (nm)0.1
0.15
Emitting Density (n/nc)

0.2

FIG. 1. Typical time resolved spectra of Raman scattered light in (a) forward and (b) backward
direction. Incident energies: fa) 26 J, (b) 13 J.

2.

INTERACTION STUDIES

Studies on laser-plasma interaction physics are related to parametric instabilities in homogeneous plasmas. This is the corresponding situation for large size
targets. As to short wavelength lasers, it appears, at present, that Brillouin backscattering [2] does not have a high conversion efficiency even, for plasma density
scalelengths of the order of several hundred times the laser wavelength. Plasma
collisionality and Landau damping of ion waves are to be invoked to explain
instability saturation at a level of < 1 % for intensities of 2 X 1015 W-crrf2 at 0.26 jum
wavelength. The Raman scattering instability [3] causes, however, some concern because
it is related to very energetic electron generation, leading to possible target preheat. To investigate SRS, long scalelength plasmas have been generated by irradiating thin plastic foils at 0.26 jum with intensities of up to 2 X 1015 W-cm"2. The
fact that the foils become underdense during the laser pulse generates plasmas withparabolic density profiles, with density scalelengths of several hundreds of wavelengths. Time resolved Raman spectra were observed, as is shown in Fig. 1, bothfor backward and forward Raman instabilities. The parameterization of the experimental conditions, i.e. foil thickness and, consequently, plasma density scalelengthand temperature, lead us to conclude that the threshold conditions for SRS arereached only in a narrow region, near the top of the parabolic density profile when
the foil is exploded before the maximum intensity of the laser pulse. This is confirmed by very narrow spectral broadening of Raman scattered (ACJ/CO < 0.05)
light. SRS occurs for electron densities of 0.07 to 0.15 of the critical density.

82

AMIRANOFF et al.

i

i

i

F dN-dE"1-sr-l

107
10s -

10s

10* -

\

A

\
A V
c K) ioo f§o :E (keV)

FIG. 2. Energy distribution of hot electrons.

The temporal behaviour of Raman spectra shifting towards wavelengths with timeis explained by the time decrease in the electron density for the unstable region due
to plasma expansion. The conversion efficiency of SRS has been measured at alevel of 10~8 at 0.26 jum, and at a significantly lower level (by ten times) the
conversion efficiency was measured for comparable experiments at 0.53 jxm. It
appears that the behaviour of Raman instability, conversion level and laser wavelength dependence can be explained by the plasma collisionality which introduces a
damping of the electromagnetic waves and of the electron plasma wave by electronion collisions. The dependence of the Raman level on the incident laser energy
shows a very fast growth around 2 X 1O1S W-cirf2, which is in agreement with thethreshold computed for a parabolic density profile. Fast electrons correlated withSRS at 0.26 fim have been detected. Their energy distribution (Fig. 2) corresponds
to a hot temperature of 20 ± 2 keV, and the total number of electrons is between
5 X 107 and 5 X 108 sr"1, which is again ten times lower than that for the experiment at 0.53 fim.

3.

ELECTRON HEAT TRANSPORTElectron heat conduction has been analysed [4] by delocalized heat transport
theory, which is an extension of the linear Spitzer-Harm laser theory, when thetemperature gradient becomes comparable to or smaller than the mean free path ofthese electrons, which carry the heat flux. An expression of the heat flux has been
derived:q ( x ) = / dx'q S H (x')w(x, x')A)

IAEA-CN-47/B-n-3

83

0.01

FIG. 3. Non-local propagator for heat flux due to steep temperature profile w is normalized
such that w(0j = 0 and £ is normalized distance from source point. Curve includes effect of nonlinear electric field, ensuring zero current condition.

where qsn( x ') is the linear Spitzer-Harm heat flux and w(x, x'.) is a delocalization
propagator. This first proposed expression for w was:
= [l/2A d (x')]exp[-|s|/A d (x')]
where

dx"n(x")and
Xd = 31 (Z + 1 ) 1 / 2 [T2/47rne(Z + 1) e4 lnA]
More elaborate forms of w were proposed since, based on a detailed analytic justification of such a delocalization formula. A more satisfactory propagator [5] isgiven in Fig. 3, where £ = 5.5 s/Xd is a normalized space co-ordinate. In particular,
it includes the effect of the non-linear electric field, ensuring the zero current condition. Introduction of this transport theory into numerical simulation yields, asa first order description, equivalance with a flux inhibition factor of 0.1 plus some
additional transport term associated to the fastest electrons. Experiments at
0.26 (im based on ablation measurements in spherical geometry [6] or spectroscopic
determination [7] of temperature and density profiles gave reasonably good agreement with this description of transport.

84

AMIRANOFF et al.
100 T P(Mbar)

Theoretical law at 0.26jum
P=1SI03/t—• •• 0 . 2 6 /io : 1.05/j*

Experimental law at Q26um

10

IQ(W-cm-2)
1On

FIG. 4. Ablation pressure as a function of absorbed intensity at 0.26 /dm and 1.05 p.m laser
wavelengths.

_93 Mbpr*V~—Experiment

Af-Au targetAt

Experiment Au-AJ target
Fluid velocity
(m
20

30

FIG. 5. Polar shock of Al and Au for incident pressure of 45 Mbar in firs} material; determination of theoretically transmitted pressure in second material; corresponding experimental points.

85

IAEA-CN-47/B-H-3(ps)

O 9um : depth of the groove (experimental points)/

25

Pressures (Mbar)

400

two-dimensional simulation300

one-dimensional simulation

200

10

20

30

40

SO

()im)

initial distance of separation

FIG. 6. Variation of induced shock wave transit time, through 9 \xm depth groove, at rear of
15 p.m Mo impact foil, as a function of initial distance of separation between accelerated andimpact foils.

3.

HIGH PRESSURE GENERATION

The goal of this experimental work is to prove the feasibility of very highpressure by the laser ablation technique [8 , 9]. Experiments were conducted onplane targets irradiated at 0.26 nm. The pressure is determined by measuring
the shock velocity through a foil, by the usual step technique. Photographing,with a fast streak camera, the emergence of the shock on either side of the step
we find the shock velocity if we know the step thickness. An improvement of thistechnique has been achieved by using, instead of a step, a groove inside a foil which
results in higher precision in the time interval determination and reduces the edge
effects in the shock propagation in the material. Figure 4 shows the results obtained
for ablation pressure and a comparison with the theoretical scaling law, Pab[ °c I 3/4 .
Pressures in excess of 50 Mbar were reached. Some discrepancy exists between theexperimental and the theoretical scaling law. This has been explained by the effectof some lateral conduction. In fact, the intensity on the target is plotted by taking
the focal spot area in vacuum. However, when the correction was made by taking,as the focal spot size, the extended region where the shock is observed, much better
agreement was reached. For the purpose of implosion and also of fundamental
studies of the equations of state, higher pressures are needed. Two multiplication
techniques were analysed. First, shock impedance mismatch was obtained when ashock is driven by ablation pressure on a material such as aluminium or plastic and transmitted through a heavier material such as gold in intimate contact with the light material.
The results of such experiments are as shown in Fig. 5 for an experiment on the Al-Au
sandwich. Starting from a 45 Mbar pressure in Al, a pressure of 100 Mbar wasmeasured in gold. More recent experiments of this type have shown that in this86

AMIRANOFF et al.

pressure range very good agreement was obtained when comparing the intersection ofof the symmetrical of the shock polar curve of Al and the p 0 D slope line, where D is theshock velocity in gold, with the shock polar curve for gold obtained from Sesamedata.
The third high pressure generation technique of interest for laser compressionis the foil collision technique. A foil accelerated by ablation pressure collides witha second foil, where a strong shock is induced. To prevent radiation preheat, theaccelerated foil was made of plastic. The main parameter in the experiment was
the initial distance between the two foils. The collided foil was made of molybdenum, to combine the impedance mismatch effects. The results are plotted inFig. 6, with a comparison with ID or 2D numerical simulation. At is the transit
time of the shock through a groove of 9 ixm depth at the rear of a 1 5 jum Mo foil.
For small distances the shock velocity increases, as a result of a longer acceleration
time and higher velocity of the colliding foil. Beyond some separation distance,
the shock velocity is reduced. This is explained by a two-dimensional effect.
Indeed, if the foil velocity is still increasing, simultaneously the radial matter flow
and the areal mass of the accelerated foil are reduced. Then, if a very high pressure
is obtained at the collision time, the shock is not maintained and is damped significantly when propagating through the modybdenum foil. The pressure measured
after propagation in the Mo foil is, consequently, much lower than the shockpressure at the first instant of collision. A comparison with the simulation showsthat the initial pressure has been of 400 Mbar, in order to induce the measured
shock velocity.

FIG. 7. Pinhole photograph of 150 jim diameter microballoon, D-T filled, imploded at
0.26 urn laser wavelength.

IAEA-CN-47/B-U-3
4.

87

IMPLOSION EXPERIMENTS AT 0.26 urn

We have proceeded to a series of preliminary implosion experiments of D-T
filled glass microballoons with the energy on target in the range of 80 to 120 J ina light pulse of 500 ps duration at 0.26 jum laser wavelength. Targets were as follows:
diameter between 90 //m to 250 /xm, wall thickness between 0.8 and 3.5 jzm, andfill pressure of 10, 20 and 40 atm of equimolar D-T mixture.
To obtain good symmetry, the focal spot was localized at a distance D from thetarget centre beyond the target wall with D/R > 2. For these conditions the
implosion symmetry was correct at first sight, as is shown in Fig. 7 by the pinhole
photograph. However, these conditions did not correspond to the maximum ofneutron emission which was obtained when D was nearly <1 , enhancing the effectof the initial shock wave which seems to be the major reason for neutron emission.
The diagnostics on these experiments were neutron measurement, measurement
and spectroscopy of a-particles and protons with CR 39 foil technique and X-ray
photographs.
The highest level of neutron emission was obtained for targets having thefollowing parameters:

N

W/T
(Mm)

P
(atm)

D/R

Gum)
170

1.3

40

1.6

112

2.5 X 108

163

1.4

20

1

110

1.1 X 10s

Diameter

E

y

(J)

These levels are, however, much below — by two orders of magnitude the numerical prediction of the ID Lagrangian FILM code. Owing to the strong
dependence of the fusion reaction on the ion temperature, this is, however, amoderate discrepancy and can be explained by the end of the implosion process
before the stagnation point, because of the macroscopic non-uniformity of the
implosion associated with the initial non-uniformity of the illumination. The
reaction rate has also been analysed as a function of the target parameters. The
experiments show that neutron emission increases with the target diameter. Thisis explained by the fact that the implosion time depends on the target size; forsmall targets, the end of the implosion may occur before the maximum of thelaser pulse, resulting in a less efficient coupling of laser energy to the target.
The second set of shots was made by varying only the wall thickness of thetarget. For full stable ablation, numerical simulation of the implosion would predict a slight increase in neutron emission with the wall thickness in the range ofthe parameter of these targets. The experiment shows a significant decrease in

88

AMIRANOFF et al.

50.0 .

0.0

0.0

3.0

6.0
9.0
RANGE LOS5 IN TARGET

12.0

1S.0

FIG. 8. Energy spectrum ofa-particles from tracks in CR 39. x-axis is ploted in range ofloss target, expressed in pm of plastic; p&R is 1.04 X 10'3 g-cm'3; 7], deduced from spectra,
islAkeV.

10.0

8.0

6.0

D
O
U
2.0

0.0
-50.0

-20.0

10.0
40.0
RANGE LO55 IN TARGET

70-0

100.0

FIG. 9. Proton energy spectrum in the same conditions as in Fig. 8.

89

IAEA-CN-47/B-II-3'2x io-3•io- 3

SHELL THICKNESS

FIG. 10. phR measurement versus initial target wall thickness.

Ao-PARTICLES()

D-T NEUTRONS

0
.10'

.10°

0.1

1

10

T F (fuel) (keV)

FIG. 11. a-particle or neutron yield versus ion temperature. Data outside line probably
come from inhomogeneous implosions.

90

AMIRANOFF et al.

neutron yield for increasing wall thickness. This discrepancy leads us to the conclusion that the implosion does stop before stagnation and, indeed, when one
examines the neutron emission as a function of wall thickness at an earlier stageof the implosion, for example at the first shock convergence and also at the firstshock reflection from the imploding wall, then the neutron emission decreases
when the wall thickness increases.
The technique of track analysis on CR 39 foils has been used for a-particle
and proton spectroscopy. Spectra are shown in Figs 8 and 9. The position of
the peak of distribution gives the energy shift due to the particle slowing downwhen crossing the shell and then thepARof the shell. The broadening of the distribution yields the ion temperature. In Fig. 10, the value of pAR is plotted as a functionof initial target wall thickness. Comparison was made with the numerical simulation at the time when numerical and experimental neutron emissions were inagreement. In general, the experimental pAR were larger than the numerical ones,
this discrepancy decreasing for larger wall thickness. These results seem to showthat the implosion is proceeding beyond the first shock convergence time whenexperimental neutron emission will more or less agree with the numerical simulation but imply, then, that the fuel temperature is lower than the code prediction,
because of a possible mix between the inner side of the wall and the fuel, resulting
in cooling of the D-T mixture. Better agreement for thicker wall targets implies
a better stability of the inner wall and confirms, however, that implosion must stop
between the first shock convergence and the shock reflection on the inner wall.
The ion temperature determination is shown in Fig. 11, where Tj is plotted versus
the neutron or a-particle yield. One can see here two groups of experiments: thefirst one with NaTf, where 3.5 < a < 5. These data agree with the dependence of the
D-T reaction cross-section on the temperature which scales as T? around T; = 2 keV.
The second set of data in the figure can be explained by broadening of distribution
due to shell inhomogeneities.
From the previous results and their comparison with numerical data, thefollowing picture of the process can be drawn: the implosion is relatively stable until
the time corresponding to the reflexion of the shock wave on the imploding wall;at this time, the magnitude of neutron emission and its dependence on shell thickness as obtained by experiment and numerical simulation are in qualitative agreement. However, the inner face of the shell goes unstable, involving some mix with
D-T and cooling of the fuel. After this time, the implosion ends because of theshell instability associated to slowing down before stagnation, resuling in a breakingof the shell and, finally in the strong bright central spot emission as observed on
the pinhole photographs. The envelope of this region, whose radius is nearly r o /8,
corresponds to the values for the radius of the wall-fuel interface at this time, which
is r o /7 by numerical simulation. Neutron emission has been obtained at the time
confirmed by the experimental values of pAR.
In conclusion, we may state that this series of experiments at 0.26 pm has
shown that, as far as interaction physics is concerned, SWL implies strong plasma

IAEA-CN-47/B-H-3

91

collisionality which not only enhances absorption, but also reduces the conversion
efficiency of parametric instabilities and, consequently, reduces the emission of
fast electron associated with Raman scattering. Very high ablation pressure canbe generated which should be of the order of 100 Mbar at 1015 W-ctrf2 in sphericalgeometry. These pressures can be enhanced by convergence effect and collision
technique on structured targets, using impedance mismatching. Efficient implosion
can be driven at a high level of neutron emission. However, non-uniform illumination results in unstable behaviour of the implosion, which becomes dominant for
wail slowing before the stagnation time. When this problem is solved, it is shownthat very efficient ablative implosion will be obtained with direct drive using shortwavelength lasers.

REFERENCES
[1]
[2]
[3]

[4]
[5]
[6]
[7]
[8]
[9]

FABRE, E., in Plasma Physics and Controlled Nuclear Fusion Research 1980 (Proc. 8thInt. Conf. Brussels, 1980), Vol. 2, IAEA, Vienna (1981) 263.
LABAUNE, C, et al., Phys. Rev., A 32 (1985) 577.
LABAUNE, C, et al., in Controlled Fusion and Plasma Physics (Proc. 12th Europ. Conf.
Budapest, 1985), Pt. I, European Physical Society (1985) 686.
TURNER, R.E., et al., Phys. Rev. Lett. 54 (1985) 189.
LUCIANI, J.F.,et al., Phys. Fluids 28 (1985) 835.
LUCIANI, J.F., et al., Phys. Rev. Lett. 55 (1985) 2421.
FABRE, E., et al., Rapport d'activite du GRECO Interaction Laser-Matiere,
Centre National de la Recherche Scientifique (1983)223.
ALATERRE, Ph., et al., Phys. Rev., A 32 (1985) 324.
FABBRO, R., et al., Laser Part. Beams 4 (1986) 413;
COTTET, et al., Appl. Phys. Lett. 47 (1985) 678.
TRAINOR, R.J., et al., Appl. Phys. Lett. 43 (1983) 542.

DISCUSSION

S. WITKOWSKI: Could you quantify, or comment on, the influence of thefast electrons produced by stimulated Raman scattering (SRS) at 0.26 Mm on thepreheating of the pellet.
E. FABRE: In our present experiments the percentage of SRS and the energy
in the associated fast electrons is negligible — 10~8 of the laser energy. Moreover,a fast electron temperature of 20 keV is moderate, and these electrons can be
stopped. However, we need experiments with much larger plasmas, closer to future
large target conditions, in order to verify whether the saturation observed is still
occurring for the backward Raman scattering, and also to see that the forward
Raman scattering does not reach a high level since this instability could produce
very fast electrons.

92

AMIRANOFF et al.

K. SATO: Although you mentioned that experimental results of neutron
yield dependence on temperature are consistent with the cross-sections, the actual
dependence of the cross-section on temperature is much higher than the value of
3.6 which you report. Could you explain this discrepancy?
E. FABRE: The experimental neutron or a-particle yield scaling with thefuel temperatures cannot be determined with good accuracy owing to experimental
error in neutron measurement. It scales withTf when 3.5 < a < 6, as is written
in my paper. What we expect from the D-T cross-section dependence on Tjfor
values of Tj around 2 keV is a scaling where a = 5.

IAEA-CN-47/B-H-4

PROGRESS IN INERTIAL CONFINEMENTFUSION WITH LIGHT ION BEAMSAT SANDIA NATIONAL LABORATORIES
J.P. VanDEVENDER
Sandia National Laboratories,
Albuquerque, New Mexico,
United States of America
Abstract
PROGRESS IN INERTIAL CONFINEMENT FUSION WITH LIGHT ION BEAMS AT SANDIA
NATIONAL LABORATORIES.
The construction phase of the Particle Beam Fusion Accelerator (PBFA II) was finished with asuccessful first shot on 11 December 1985. The paper describes progress in bringing the pulsed power,
power flow and ion diode systems to operational status.

PBFA II Introduction
Light ion beams [1] offer the possibility of a very
efficient and low-cost driver for Inertial Confinement Fusion(ICF). The energy deposition is straightforward since the ions
deposit their energy in a dense plasma that prevents
microscopic instabilities from producing preheating electrons.
The difficulty with light ions has been the focusability. In
1984, a proof-of-principle experiment on Proto I, at the same
current density and charge density required for inertial fusionon the Particle Beam Fusion Accelerator II (PBFA II), showedthat intense ion beams can be focused to the required
divergence with the correct local physics [2]. In 1985, diodeand accelerator technology was examined [3] on PBFA I at thesame current and diode radius required for fusion on PBFA II.
PBFA II [4] is the latest of a series of particle beam
generators designed and tested to provide a power source fordriving ICF targets. It is the first one that has been
designed as an intense ion accelerator and the first with thepotential of igniting thermonuclear fuel. Since the lithium
ions deposit their energy efficiently in a dense plasma whichshields the beam ions from each other and remains very
collisional, instabilities and the resulting preheat are
precluded. Consequently, energy deposition is classical andbenign. The principal uncertainties are the generation andfocusing ofa million joule (MJ) ion beam to the required
100 TW/cm 2 power density, which requires a 5 kA/cm'2 Li + ion
source and a divergence of 10 to 15 mrad. Consequently, almost
all of the work on the light ion approach to ICF has centered93

94

VanDEVENDER

FIG. 1. Artist's cutaway drawing of PBFA II. Energy from the capacitors in the outer annulus isdelivered to the target in the center through the multi-module power-conditioning network.

on the generation and focusing of light ion beams. The target
implosion experiments performed with other ICF drivers are
generally applicable to the targets for light ions also, so
target issues have not been neglected by this strategy.
PBFA II is shown schematically in Fig. 1. The accelerator
was activated on December 11, 1985, with all components outside
the target in place for testing. Preparing it for target
experiments requires three steps: a series of "shakedown"
shots to stress the accelerator and identify weaknesses, aseries of experiments to optimize the pulsed power and todevelop the plasma opening switch (POS), and a final series todevelop the beam technologies. The first of these steps iscompleted and the second is underway. The major technicalissues for each of the PBFA II sections will be presented andevaluated.
Energy Storage Section
The energy is stored in 36 Marx generators, large banks inwhich 60 capacitors are charged in parallel and discharged inseries. The 36 modules store 13 MJ of energy and discharge
into a pulse-forming section in 1 us by firing 1080 spark gaps

IAEA-CN-47/B-II-4

95

inside the generators. Reliability, prefire control, andsystem timing uncertainty or jitter were the major issues.
Experience on the PBFA I predecessor to PBFA II indicated thatthese would be major problems. The solution [5] required thechanging internal resistors to promote a smooth spark-gapbreakdown sequence in the first two rows of the Marx to makespark gap firing sequence repeatable and, therefore, lowjitter. The stable firing sequence was subsequently verified
with new photonic diagnostics [6]. The same system changes
were adapted on PBFA I and the timing spread was reduced from-70 to 120 ns to the 20 ns range, and prefires were virtuallyeliminated. The 36 PBFA II generators each have a reliabilityof >0.9995, which corresponds to a single gap reliability of>0.99998, and a PBFA II verified timing spread of 20 to 40 ns.
This spread is adequate and does not contribute to target
asynchrony.
Pulse-Forming Section
The pulse-forming section [4] in the water-filled annulus
of PBFA II compresses the electrical power pulse from the Marx
generator by a factor of 8, synchronizes all 36 modules,
inverts the electric field direction on half of the accelerator
to add voltages in series, and transforms the impedance tomatch that of the vacuum section.
All 36 modules must be synchronized within 15 to 20 ns for
maximum accelerator power and sub-nanosecond asynchrony at thetarget. Modules are synchronized with a laser-triggered gas
switch. The strong coupling [7,8] between KrF laser light andSF6 insulating gas has been developed into a reliable, 5 MV,
low prefire, low jitter, voltage insensitive switch technology[9] for PBFA II. The novel multi-stage switch design hasseries gaps arrayed along a central column and isolated fromthe segmented gas-switch housing. The laser trigger ensures
low jitter. A new model [10] for gas switch breakdown has beendeveloped and utilized to reduce prefires, jitter, and voltage
sensitivity of high voltage gas switches. This gas switch is a
key component of PBFA II and operates nominally at 5.2 MV,
0.9 us charge time, with 2 ns rms jitter, and with >99.7& (fromDEMON data) reliability against prefires. The single 4 J,
injection-locked KrF laser beam is divided into 36 beams which
fire all 36 modules simultaneously. The laser itself had a
misfire rate of 5% which has been improved to a rate of lessthan 0.2%. However, before this misfire improvement was made,a worst case accident occurred on the fifth "shakedown" shot
and with the accelerator set to fire at full energy. A laser
misfire could have been catastrophic if that fault mode had notbeen accommodated in the design. The energy left in theintermediate store might have destroyed the energy-storage
section, the storage capacitors, and the gas switches, andmight have mixed the oil and water. The robust accelerator
survived with only minor damage to the gas switches and useful

96

VanDEVENDER

data was obtained. Early tests of such fault modes areimportant in making a large pulsed-power system fully
operational and survival indicates the integrity of the design.
The pulse-forming lines of PBFA II compress the energy
into a 55 ns wide pulse by a two-staged, water-dielectricsvitched, pulse-forming network. A double-bounce, charging
scheme [11] was developed and adopted [4] for PBFA II tominimize the voltage on the lines. The pulse-forming section
produces a 5.8 MJ, 100 TW, and 3.2 MV pulse from its 36 modules
into the polarity-inverter-transformer combination.
Since a 30 MV pulse is desired to give the lithium ion
optimum range in the target, the 36 modules are' combined in aseries-parallel combination to provide a 12 MV source for thevacuum section. The final voltage gain is accomplished by the
POS [12,13] in vacuum. Each of the coaxial transmission linesof the pulse-forming section are split, with an 88% energy
efficiency, into two parallel-plate transmission lines. The
one further from the mid-plane of the accelerator has a crossover network that inverts the electric field vectors.
Parasitic losses from the inverter reduce the energy in that
portion of the line to 80% of its initial value. The line
without the inverter has no such loss, so approximately 120 kJ
of energy is injected into the transformer. Measurements ofthe energy and power transport through the inverter agree withcomputed values [14].
Transformer Losses
Although approximately 4.6 MJ of energy is injected into
the 36 transformers, only 76% is available at the output for
the water section. The impedance transformation in each of the72 parallel plate transformers is from 4.3 ohm input to10 ohm output in a 2.2 m long linear taper. The computed
efficiency is 95% if the lines are uncoupled and if radiative
losses are negligible. Experiments on a 1/5 scale model of 18
of the modules showed the losses are not dependent on adjacentmodules. It appears that the electromagnetic coupling between
the various lines in the transformer causes destructive
interference and reduces the output pulse by 30%. These losses
are still present in PBFA II. Alternate transformer
configurations are being investigated through 3-D computer
simulations to reduce the parasitic losses. The potential forsignificantly increasing the PBFA II energy makes this researcha high priority.
Vacuum Insulator Stack
The central vacuum section of PBFA II is 3.6 m in diameterand 4.8 m high. The 59 square meters of insulator are required
to feed 100 TW of power in the baseline design. The insulator

IAEA-CN-47/B-II-4

97

performed well in the "shakedown" shots and should accept theadditional energy resulting from improvements in thetransformer efficiency.
Magnetically Insulated Transmission Lines (MITLs)
The MITLs provide azimuthal smoothing of the wavefronts
from the 9 modules feeding each level. This azimuthalsmoothing is a key step in reducing the beam asynchrony at thetarget to the required sub-nanosecond level. In addition,these transmission lines provide the magnetically insulated
storage inductor for the last stage of power compression andvoltage gain. Since each half of the accelerator has 4 biconic
MITLs added in series, electron losses at the junctions arepotential energy losses. These losses are computed to be small
because the POS provides the initial low impedance to keep theelectron flow overly trapped.
The computed and measured current waveforms agree towithin a few percent and indicate the monitoring system isadequate, the energy per module is as expected, and the MITLs
are functioning to azimuthally and vertically smooth the wave
as designed. More testing with optimal POS conduction andopening and normal module timing is required to fully test the
MITL performance, but the shakedown tests are encouraging.
Plasma Opening Switch (POS)The POS is a joint research and development project with
the Naval Research Labortory (NRL), the lead laboratory in the
research. Substantial contributions have been made byindividuals at Sandia National Laboratories (SNL) and at theLaboratory of Plasma Studies of Cornell University.
Experiments and theoretical studies in many other institutions
have added to the understanding of POS operation [15-17].
Nevertheless, that understanding is still incomplete and the
POS is still a moderate risk component. The PBFA II POS could
not be fully tested without PBFA II.
Early results are encouraging. The switch conducts forthe designed current but opens in approximately 30 ns insteadof the 12-15 ns required. A series of tests are in progress to
optimize the POS performance and to determine whether or not asecond iteration in the design will be required. Recent
experiments with a small cathode radius [18] or a series fieldcoil in the cathode [19] have shown improved POS performance.
These innovations offer possible improvements in the PBFA II
POS. The voltage and power gain expected of the POS are
essential for the PBFA II ignition experiments. However, thelonger-term experiments with pulse shaping, beam extraction andbeam transport will not require the POS. If the POS
performance is found to be inadequate, the schedule for thepulse-shaping option will be accelerated.

98

VanDEVENDER

Lithium Ion Source
At the last meeting of this conference, the development ofa suitable lithium ion source was identified as a major
priority for the light ion approach to ICF. Since that time,
more than 20 proposed sources were evaluated and 5 wereselected for development: a glow-discharge-cleaned LiF
flashover source [20], an electrohydrodynamic-instability
driven liquid lithium source [21], an impact ionized Li gas gun
approach [22], an approach [23] producing a Li vapor by flash
heating a LiAg thin film with an electrical heater and ionizing
the vapor through the LIBORS [24] process with a visible dyelaser, and laser vaporization followed by single-photon
ionization with an ArF laser [25]. All 5 sources have shown
encouraging results. The last 3 have shown
evidence forlithium ion densities in excess of the 10 1 7 cm"3 specified by
the design criteria. The first one has been fielded on PBFA
II, but the lithium content of the ion beam was negligible atthe 8 MV operating level achieved in the "shakedown" shots.
This source will be tested as the voltage increases with POS
optimization. The second, third, and fourth sources are being
prepared for PBFA II testing during the next year. The last is
being examined for feasibility. The number of promisingsources is encouraging; only one has to work. Any of the 5should be acceptable for the PBFA II experiments, but only thesecond can be readily extended to multi-shot or repetitive
operation required by the ICF drivers following PBFA II.
Ion Beam Generation and Focusing
The PBFA II ion diode is shown in Fig. 2; it has the sameradius as the PBFA I diode, but twice the height. The
operating point is to. be 30 MV and 5 MA compared to the 1.8 MV
and 5 MA of PBFA I. The desired ion is lithium instead of the
protons accelerated on PBFA I. The PBFA II diode features amuch larger field coil in the anode and one in the mid-plane of
the cathode. The new magnetic field geometry reduces themagnetic field energy required to insulate the electrons by afactor of 10, and compensates for the Li + canonical momentum
change as it strips to L i + + + in the gas-cell membrane. The
proton impurities that could preheat the target are excluded
from the target region by their canonical momentum [26] in the
PBFA II geometry.
The diode was designed [27] with the MAGIC [28] 2-D, fully
electromagnetic, PIC code merged with the accurate magnetic
field profile coputed with the TRIDIF code [29]. The insights
gained from these computations have been generally encouraging.
However, the ion current is significantly delayed, even with
the ion-source space-charge limited at the beginning of thecalculation. A new model [30] for the impedance of ion diodes
has been advanced substantially since the last conference and99

IAEA-CN-47/B-H-4
TOP CATHODEASSEMBLYGAS CELLANODE
PLASMA(ION SOURCE)

BOTTOM CATHODEASSEMBLY
POWER FEEDFIG. 2. PBFAII ion diode region and plasma opening switch (POS) which will produce the high-energy
ion beam and direct it onto the central target.

the data from many ion diode experiments has been successfully
explained with the electron lifetime (or, equivalently, the ion
production efficiency) as the only free parameter. The MillerMendel [30] model suggests that new techniques for quickly
establishing the electron charge - which determines the ion
current for a given geometry and voltage - in the accelerating
gap will be necessary to efficiently couple the ion diode to a
10 to 15 ns PBFA II pulse. In addition, techniques forcontrolling the electron lifetime after the desired ion current
is reached will be required to maintain or preferably increase
the diode voltage during the pulse. Until now, only limited
attention has been given to electron control, such as theelectron limiter [31] on the anode. There is a great need andopportunity for significant advances in ion diode physics
through new ideas for electron control. The recent
demonstration of electron retrapping in strong applied magnetic
fields [19] may stimulate such new approaches.
Although the experiments in 1986 emphasized pulsed power
development, many of the experiments had an ion diode. The
coupling to the diode was poor when the plasma opening switch

100

VanDEVENDER

was used because the top and bottom switches have not beenadequately simultaneous and because the diode impedance is toohigh for too long at the 8-9 MV of recent experiments.
Experiments without the opening switch provide a longer pulse
at 8 MV and permit the diode model in Ref. [30] to be tested inthis regime. The model predictions agree well with themeasured current and voltages. The model indicatesapproximately 1.5 MJ of ions and 100 KJ of electrons areproduced. No direct measurements of the total ion energy havebeen obtained. However, x-ray measurements can account for
only 200 KJ of the 1 to 2 MJ available and are consistent witha megajoule ion beam. Direct measurements are now underway.

Acknowledgements
This work was supported by the U.S. Department of Energy
under Contract DE-AC04-76DP00789.

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

VanDEVENDER, J.P., COOK, D.L., Science 232 (1986) 837.
JOHNSON, D.L., et al., J. Appl. Phys. 58 (1985) 12.
MAENCHEN, J.E., et al., Rep. SAND-87-0844, Sandia National Laboratories (1987).
TURMAN, B.N., et al., in Proc. 5th IEEE Pulsed Power Conf. (Arlington, VA, 1985), Instituteof Electrical and Electronics Engineers, Inc., New York (1985) 155.
SCHNEIDER, L.X., LOCKWOOD, G.J., ibid., p. 780.
LOCKWOOD, G.J., et al., ibid., p. 784.
W00DW0RTH, J.R., et al., J. Appl. Phys. 53 (1982) 7.
RAPOPORT, W.R., et al., IEEE Trans. Plasma Sci. PS 58 (1980) 3.
HUMPHREYS, D.R., et al., in Proc. 5th IEEE Pulsed Power Conf. (Arlington, VA, 1985),
Institute of Electrical and Electronics Engineers, Inc., New York (1985) 262.
MARTIN, T.H., ibid., p. 74.
SMITH, I.D. (Pulse Sciences, Inc., Oakland, CA), Rapid Charging of Transmission Lines,
Interim Report for Sandia National Laboratories, Apr. 1981.
MENDEL, C.W., Jr., GOLDSTEIN, S.A., J. Appl. Phys. 48 (1977) 1004.
WEBER, B.V., et al., Appl. Phys. 45 (1984) 1043.
OTTINGER, P.F., et al., J. Appl. Phys. 56 (1984) 774.
NEAU, E.L., et al., in Proc. 5th IEEE Pulsed Power Conf. (Arlington, VA, 1985), Institute ofElectrical and Electronics Engineers, Inc., New York (1985) 772.
KAROW, H., etal., ibid.
IMASAKI, K., in Plasma Physics and Controlled Nuclear Fusion Research 1984 (Proc. 10th Int.
Conf. London, 1984), Vol. 3, IAEA, Vienna (1985) 71.
STINNETT, R.W., et al., Plasma opening switch development for the Particle Beam Fusion
Accelerator (PBFA-II), submitted to IEEE Transactions.
OTTINGER, P.F., WEBER, B.V., internal report, Naval Research Laboratory,
Washington, DC.

IAEA-CN-47/B-II-4
[19]
[20]
[21]
[22]
[23]
[24]
[25]
[26]
[27]

[28]
[29]
[30]

[31]

101

MENDEL, C.W., Jr., et al., Bull. Am. Phys. Soc. 31 (1986) 1412.
BIEG, K.W., et al., Flashover lithium ion source development for large pulsed power accelerators, submitted to J. Vac. Sci. Technol.
PREGENZER, A.L., J. Appl. Phys. 58 (1985) 4509.
GOLDSTEIN, S.A., et al., internal report, G-T Devices, Inc., Alexandria, VA.
DREIKE, P.L., TISONE, G.C., J. Appl. Phys. 59 (1986) 371.
MEASURES, R.M., CARDINAL, P.G., Phys. Rev., A 23 (1981) 804.
TANG, K., LIN, S., Western Research Corporation, San Diego, CA, personal communication.
DREIKE, P.L., MILLER, P.A., J. Appl. Phys. 57 (1985) 1589.
SLUTZ, S.A., et al., Bull. Am. Phys. Soc. 29 (1984) 1343;
MIX, L.P., et al., Bull. Am. Phys. Soc. 30 (1985) 1514;
ROVANG, D.C., in Proc. 5th IEEE Pulsed Power Conf. (Arlington, VA, 1985), Institute ofElectrical and Electronics Engineers, Inc., New York (1985) 567.
DREIKE, P.L., TISONE, G.C., J. Appl. Phys. 59 (1986) 371.
GOPLEN, B., et al., User's Manual for MAGIC, Version - Sep. 1983, Rep. MRC/WDCR-068, Mission Research Corporation, Alexandria, VA (1983).
FREEMAN, J.R., J. Comput. Phys. 41 (1981) 142.
MILLER, P.A., MENDEL, C.W., Jr., Bull. Am. Phys. Soc. 30 (1985) 1513;
MILLER, P.A., Rep. SAND-86-0015, Sandia National Laboratories, Albuquerque, NM (1986);
MILLER, P.A., MENDEL, C.W., Jr., J. Appl. Phys. 61 (1987) 529.
SLUTZ, S.A., Rep. SAND-86-0016, Sandia National Laboratories, Albuquerque, NM (1986);
MAENCHEN, J.E., et al., in High-Power Particle Beams (Proc. 6th Int. Conf. Kobe, 1986),
Laser Society of Japan (1986).

DISCUSSION
D.D. RYUTOV: Do you see any possibility of developing a Li 2+ or Li 3+ ion
source? This might well facilitate the production of a 30 MeV beam.
J.P. VanDEVENDER: The beam must be a pure, single species to avoid preheating the ions. If we wanted a lower voltage and higher current, we would turn todeuterons, since they have almost the same charge-to-mass ratio as Li 3+ and would
be purer. However, we need the 1 to 7 ratio of Li + to reduce magnetic bending to
low levels, so we will be continuing with Li + .
K. YATSUI: Does the appearance of EHD instability strongly affect the divergence angle of the extracted ion beam? And have you measured these dependences
experimentally in several parameter spaces?
J.P. VanDEVENDER: No, the divergence has not been measured. Calculationsof the divergence under PBFA II conditions by Klingham in the United Kingdom
indicate 2 mrad. It will be measured on PBFA II.
K. KASUYA: You said that your LIBORS would be very expensive. Could you
give a few more details of that?
J.P. VanDEVENDER: Yes. For example, the dye laser (200 J) costs
US $600 000 and the system for flash heating a lithium film anode would cost about
US $1 400 000; manpower is additional.

102

VanDEVENDER

K. KASUYA: And what happens if your plasma opening switch fails? Would
you go back again to a proton or a deuteron beam?
J.P. VanDEVENDER: No, we would use an alternative opening switch, as
recently invented by C.W. Mendel at Sandia Laboratories; or else we would stack
all modules in series to generate 30 MV without a switch, in which case there would
be no power gain.

IAEA-CN-47/B-II-5

INERTIAL CONFINEMENT FUSION RESEARCH
WITH LIGHT ION BEAMS AT ILE
K. IMASAKI, S. MIYAMOTO, S. NAKAI,
K. NISHIHARA, H. TAKABE, C. YAMANAKA
Institute of Laser Engineering (ILE),
Osaka University,
Suita, Osaka, Japan

Abstract
INERTIAL CONFINEMENT FUSION RESEARCH WITH LIGHT ION BEAMS AT ILE.
The diode physics with pulsed power technology has been investigated on a Reiden IV-H
machine with a 180 kj Marx generator (output 6.4 MV, 0.3 MA, 1.8 TW). A brightness of1014 W/(cm2-rad2) was achieved using a planar inverse pinch diode. For the focusing diode, theaberrations induced by the self-magnetic field and the external magnetic field could be correctedby the anode with a non-spherical surface structure, which might improve the focusing. The
hydrodynamic processes in the targets caused by irradiation with focused light ion beams were
investigated for plane and cylindrical geometry. Driving pressure scalings for various targets
with different ion energies were obtained from experiments, and computer simulations.
Irradiations of Cannonball targets were performed; these were composed of a 100 jum thickpolyethylene foil sandwiched between 10 /um and 20 /Lim thick copper foils. The pressure in thecase of the 3 MeV proton irradiation was 5-6 times higher than that in the case of the ablative
acceleration target. The deposition profiles of the light ion beam in the solid target may be
advantageous in Cannonball targets. The temperatures of the front and rear surfaces of various
targets were monitored by diodes to study the conversion processes of the beam energy to theradiation in the target.

1.

INTRODUCTION

The first and most important milestone for inertial confinement fusion (ICF)
is to achieve ignition. Because of the recent progresses in high power particle beamtechnology, light ion beams can be considered to be one of the candidates for
ignition energy drivers.
The advantages of light ion beams are a large output energy, a high efficiency
and a low cost of the total system. An important issue is the focusing of the beamand therefore large efforts were made to obtain sufficient focusing.

103

104

IMASAKI et al.
PR ~ 0.4
107

r||~0.05_ _O , A=0.05

r|,~O.J
1 0 6 ^-o.oi \La<«erignitionTe ~ T| ~ 4 keV
LIB

Ignition
Xng'i

10s

1

'H

-10-"

10"

" 10"3
103
102

10'

10 2

10 3 10" 10 6

Beam Energy

10 6

10 7 10 8 10"

onto Target( j )FIG. 1. Ignition region in the jj. and EQ plane. The compression ratio JX is limited by rijand ffj.

FIG. 2. Schematic picture of the Mahobin target. K is the tamper thickness, AR is the radiator
thickness and Rfis the fuel pellet radius.

2.

REQUIREMENTS FOR IGNITION BY LIGHT ION BEAMS

A rough estimate obtained from energy balance and mass conservation
equations1 indicates the required driver energy for the ignition [1 ]:Ejg~4X109(l +AJUI/3)V"2T?I2(1)

where A is the initial non-uniformity of the implosion, ju is the final compression
ratio normalized by the liquid D-T fuel density, and rii is the implosion efficiency.
1

The concept of using energy balance and mass conservation to obtain the target gain
was noted by S. Bodner in J. Fusion Energy 1-3 (1981) 2 2 1 .

IAEA-CN-47/B-II-5

105

In this case the ignition conditions are assumed to be pR ~ 0.4 g/cm2 and thetemperature is assumed to be approximately 4 keV.
The upper limit of the compression ratio ju is roughly determined by the preheat ratio fH:
3(2)

Figure 1 shows the relation of the driver energy to the compression ratio for
the ignition estimated from Eq. (1). The dashed lines are the critical lines of theignition forT?i~0.1, A ~ 0.01 and 7?j ~ 0.05, A = 0.05. The solid lines indicate
the upper limit of the compression ratio obtained from Eq. (2).
For light ion beams, the preheat ratio f^ can be expected to be lower than
10~4. In an actual case, the implosion efficiency can be varied with the beam energy
according to the particle beam range and the focusing size [2]. The hatched areain Fig. 1 shows the ignition with light ion beams, taking into account the above
considerations.
The condition required for a driver of megajoule level is to achieve a compression ratio as low as 200 with a 1 mm radius target. In this case, the nonuniformity of the implosion has to be within 3%. The Rayleigh-Taylor instabilityof the compressed core and the beam irradiance non-uniformity may not be very
serious at this low compression ratio and large target radius. The aspect ratio of thetarget radius to the wall is assumed to be 20 and the smoothness of the target is in
the sub-micrometre range.
The key issue of ICF with light ion beams is the achievement of these parameters.
The Mahobin target proposed by us [3] seems to be favourable for ignition in thisrespect. Figure 2 shows a schematic picture of the Mahobin target. The implosion
efficiency is estimated to be up to 8%. Sufficient radiation smoothing of the nonuniformity of the implosion can be expected according to a simple estimate. The
radius of the outer high-Z shell is 2.8 mm and the thickness of the radiator is
1.6 mm. The proton beam particle energy is about 5 MeV. The power density is1014 W/cm2 and the pulse length is 10 ns with a Gaussian waveform. The radiator ismade of metal foam and has a density of 0.6 g/cm3. The tamper and pusher are
made of solid gold, with a thickness of 20 jum and 5jum, respectively.

3.

FOCUSING EXPERIMENT WITH THE RADIAL FOCUS DIODE

3.1. Radial focus diodeExperiments were performed on a Reiden IV-H pulsed power machine. The
configuration of the Reiden IV-H is shown in Fig. 3. The parameters for maximum
output are a voltage of 6.4 MV, a current of 0.3 MA and a pulse duration of 20 ns
with plasma opening switches.

106

IMASAKI et al.
Marx.G.

ISC

PFL

TL
PPS

ITL

Diode

FIG. 3. Configuration of the Reiden IV-H pulsed power machine: Marx generator, intermediatestorage capacitor (ISC), pulse forming line (PFL), transformer line (TL), prepulse switches (PPS),
impedance transformer line (ITL).

FIG. 4. Experimental apparatus of the radial focus diode.

Figure 4 shows the radial focusing diode. Assuming an enhancement factor ofeight, the diode A-K gap was set to be 7.2 mm for V = 2 MV. The coaxial output
geometry of the Reiden VI-H machine was converted to a radial geometry by crosscurrent feed. The upper and lower half of Fig. 4 show the cross-section of thediode at different angles. The anode ring was connected to the inner high voltage
electrode of the coaxial line by four spoke holders. The cathode electrode wasconnected to the ground by another four spoke holders.
The diode setting was obtained by adjusting the cathode electrode with anaccuracy of ± 15 fim. The magnetic field coils installed in the stainless steel cathodeelectrode were powered by a 400 p¥ capacitor bank charged up to 20 kV. The
insulating magnetic field was applied 60 ;us before the diode pulse. On this timescale the field diffusion into the aluminium anode was very small. The shape of

IAEA-CN-47/B-H-5

107

appliedmagnetic field

FIG. 5. Aberration of the focused beam measured by a shadow box for an anode radius of
38 mm. The black circles are data points taken at different azimuthal positions. The broken
line indicates the calculated beam deflection angle due to the self-magnetic field.

the stainless steel coil housing had a significant effect on the magnetic fieldstructure. The applied magnetic field structure was calculated by the magnetic
diffusion code ORION-FIELD [4].
We used a surface flash-over anode made of aluminium, with 1 mm deep
grooves filled with paraffin. The grooves were 0.5 mm wide and the distance
between them was 0.5 mm. The anode area was 28 cm2.
3.2. Beam trajectory control
The ion beam trajectory may be disturbed in two regions, namely the diodegap and the transport path to the target. We tried to improve the beam trajectoryin the diode gap at the first step.
The beam trajectories passing through the diode were measured with ashadow box placed 2 mm inside the cathode edges. Figure 5 shows the beamaberration as a function of the radial position on the anode surface for an anodecurvature radius of 38 mm.
In this case the term aberration means the angle between the beam trajectory
and the direction of the ballistic focus to the diode centre. The positive angle
indicates an over-focus. The black circles are data points obtained at different
radial positions. The broken line indicates the calculated beam deflection angle
due to the self-magnetic field in the diode at peak diode power.
The results show that the beam aberration is larger than the self-magnetic field
deflection. The aberration had a minimum (maximum negative aberration) atz = 2-3 mm and increased with increasing z. In the outer region of the diode, apositive aberration was obtained. This aberration may be due to the non-spherical

108

IMASAKI et al.
A9Aberration(mrad)

self-magnetic field effectZ(mm)FIG. 6. Schematic picture of the electron virtual cathode.

(b)

(a)
7
6

corrected
anode

5

z "
(rrrri)rrectedanode

29.5
R (rmO

30.0. 7. Improvement of focusability by anode curvature correction.

IAEA-CN-47/B-II-5

109

shape of the electron virtual cathode and/or the non-uniform expansion of theanode plasma in the diode as shown in Fig. 6. This aberration can be corrected bychanging the shape of the virtual cathode and the shape of the anode surface.
We performed aberration control by correcting the anode curvature (seeFig. 7). The slope corresponding to the estimated beam emitting point on the anodesurface was corrected by the difference between the ideal trajectory and the real
trajectory. The corrected anode had a non-spherical surface. Using the corrected
anode, with the same trajectory measurement, the focus size changed from 5 mm to
3 mm (Fig. 7(b)). The improved focus size could be explained by the divergenceangle due to the plasma instabilities in the diode. This fact implies that the correction of the anode curvature was very effective in improving the focusing.

4.

TARGET INVESTIGATIONInteraction experiments for the Mahobin target have been performed.
Figure 8 shows a shadowgraph of the layered target. The target is composed of a
90 nm polyethylene foil sandwiched between rear and front copper foils of 10 jrni
and 20 /um thickness, respectively.
The power densities on the copper foil were monitored using the 6SCu(d,n)6SZn
reaction. The threshold of this reaction was 2.17 MeV. A scintillator (NE 102)
and a photomultiplier tube (Amperex 2020) were used for the neutron monitor
with a 6 m flight length.
The energy spectrum of the proton beam was estimated from the neutron
waveform. The proton energy onto the target was 2.9 MeV; it corresponded fairly
well with the estimated voltage on the anode.
The beam current on the target was estimated from the neutron number andfrom carbon activation. The neutron number was monitored with dysprosium
activators with a paraffin thermalizer. The carbon activation box was placed aroundthe target. Five meshes with 30% transparency were used to avoid vaporization of
the activated carbon surface.
The current waveform was monitored with charge collectors, a B-probeanda Rogowski coil placed around the target. With this system the power density on
the target could be calculated. From the diode configuration and the damage
pattern of the solid thick-plate target, the spread of the incident angle was estimatedto be 33°. The current density on the target was 35 kA/cm2. The power density
was estimated to be 1 X 10" W/cm2 and the pulse length was 20 ns.
The acceleration pressure of the target was estimated from the behaviour of
the target. The estimated pressure was 0.5 Mbar in this case. When the same target
was irradiated with a 1 MeV proton beam at a power density of 5 X 1010 W/cm2
the pressure was 0.08 Mbar. In this case the ion beam was stopped in the first foil
and therefore there was no effect of the Cannonball shape of the target.

no

IMASAKI et al.

0 ns

100 ns

800 ns1200 ns(Burnthrough)

i—i

1cm

FIG. 8. Motion of the layered target.

deposition was calculated on the,pressure was about 1 Mbar 20 ns
f^
The experiments and simulation result
scaling obtained with the
^
^

^ £

a urn-

focused withrear surface was 106 c m / , -Fig. 9. The ablationthef o r t h e ablation targetinT h e results for^

P ea k111

IAEA-CN-47/B-II-5

to
JO

5
S)
L.

3
in
41

c
o

10
3MeV
simulations

1
Layered•target

re

>

^

o
o

o
4)

LU

0.1

n1
10 =

... } Single•••' ^ J f o i l t a r g e t_4)*y4£\ 1MeV
^i"
.10

i
^11

MeV simulation
experiment
I

I
12

10"
10"
10
Beam Intensity (W/cm4)

10

f3

10

FIG. 9. Scalings and results for the various targets obtained in experiments and computer
simulations.

obtained was 5-6 times higher than in the case of the 1 MeV proton beam irradiation.
This fact shows the high hydrodynamic efficiency of the tamper target.
Because of the low power density in our experiment, the radiation effect was
not very important. However, the simulation shows that the radiation effect will
be important when the power density reaches more than 1013 W/cm2.
The temperatures of the front, body and rear surfaces of various targets weremonitored using PIN diodes and X-ray diodes. The surface temperature was twotimes higher than the temperature of the body. This fact may be explained by
the contamination of the carbon ions in the proton beams and the deposition profile of the ion beam.
There were two peaks in the temperature on the rear surface. When the thickness of the foils was increased, the time difference between the two peaks became
larger. The first peak may correspond to the shock wave and the second peakto electron thermal conduction.

5.

SUMMARY

Beam focusing and beam-target interaction have been investigated with aradial focus diode and with a high bright diode. The aberration of the ion beamin the diode is mainly due to the self-magnetic field and the configuration of the

112

IMASAKI et al.

virtual cathode. The beam focusability has been improved by a modification of
the anode curvature.
The effect of the Cannonball shape of the target was studied using a 3 MeV
proton beam. A five times higher pressure in Cannonball targets than in ablation
targets was observed. The deposition profile of the ion beam seemed suitable forthis target. The computer simulations indicated advantages of this scheme for
ignition experiments.

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

[6]

IMASAKI, K., et al., Int. Rep ILE-8412-PR, Institute of Laser Engineering, Osaka
University (1984).
IMASAKI, K., et al., in Ion Sources and Ion Assisted Technology (Proc. 8th Symp. Tokyo,
1984), Institute of Electrical Engineers of Japan, Tokyo (1984) 143.
IMASAKI, K., etal., J.Phys. Soc. Jpn. 50(1981)3847.
OZAKI, T., et al., Tech. Rep. 34, Osaka University (1984) 71.
NISHIHARA, K., et al., in Proc. Japan-US Seminar on Theory and Application of MultiIonized Plasma Produced by Laser and Particle Beams, Nara, 1982, Rev. Laser Engng 10,
Special Issue (1982) 183.
MIYAMOTO, S., et al., Jpn. J. Appl. Phys. 21 (1982) L-83.

IAEA-CN-47/B-m-l

X-RAY DIAGNOSTIC DEVELOPMENTS FORLASER-MATTER INTERACTION EXPERIMENTS
J.L. BOURGADE, P. COMBIS, J. LAUNSPACH,
J. de MASCUREAU, A. MENS, J.M. MEXMAIN,
A. SALERES, R. SAUNEUF
Commissariat a l'energie atomique,Centre d'etudes de Limeil-Valenton,
Villeneuve-Saint-Georges,
FrancePresented by J. Coutant and M. Decroisette

Abstract
X-RAY DIAGNOSTIC DEVELOPMENTS FOR LASER-MATTER INTERACTION
EXPERIMENTS.
The paper reports on progress in the development of new X-ray diagnostic instruments
required for laser-matter interaction experiments, with emphasis on the range of soft andhard X-ray emission between 100 eV and 100 keV. For high spatial/temporal X-ray measurements, a soft X-ray streak camera with lamellar electron optics has been developed. Thiscamera is used with a cylindrical mirror for diagnosis in the range 100 eV to a few keV.
Recently, a new optimized prototype has been tested and higher spatial/temporal resolution
has been measured. For soft X-ray imaging techniques (from 50 eV to a few keV) a Wolter
microscope has been developed and tested, and a detection system has been built around asmall image converter tube with an X-ray photocathode and a phosphor screen. The image
on this screen is read by a charge-coupled device (CCD) working in the visible spectral range.
For higher energy X-ray imaging, a simple pinhole camera is used which is connected to anelectronic image readout, using a CCD array that provides two-dimensional spatial information in the 2-10 keV spectral range. For spectroscopic measurements, an X-ray spectrometer is used, with a curved T1A phosphate crystal and a linear array of 4096 photodiodes
with automatic 12 bit digitization (range 1.3-28 keV). For accurate energy balance measurements in the keV spectral range, a number of quasi-linear adiabatic X-ray calorimeters andbolometers have been built for which absolute electrical calibration is possible. The paper
also presents some preliminary results on X-ray calibration with a pulsed soft X-ray source
working with various gases.

113

114
1.

BOURGADE et al.

SOFT X-RAY IMAGING

X-ray streak cameras are a basic tool for investigation of laser driven plasmas.
Our best X-ray camera (100 eV - 10 keV) is the P650X; it is constructed aroundan image converter tube with lamellar electron optics. The main characteristicsof this instrument are as follows:— Useful slit length: 12 mm;— Screen dimension: 24 mm along spatial axis,
40 mm along time axis;— Spatial resolution with an image intensifier (ITT Bendix F4113, useful
diameter 40 mm): 17 line pairs per millimetre (Cp-mm" 1 ) over the whole
screen at 35% contrast, in dynamic mode;— Temporal resolution: 3-5 ps for a sweep speed of 2 X 109 cm • s"1.
With a thin (500 A) gold photocathode and a cylindrical gold mirror, wehave developed a streaked soft X-ray imaging device in the range 100 eV toa few keV, with 20-fold magnification. The spatial resolution is limited by thegold mirror to around 6 //m. Figure 1 shows the experimental set-up and Fig. 2
shows the results obtained with a gold target [ 1 ].
To improve and extend the sensitivity of the electron optics, we haverecently tested the new P750X R.T.C.1 prototype shown in Fig. 3 [2]. In

slitalignment device^line of sight pipe andvacuum system

X-ray image plane

gate valves.

cylindrical goldor glass mirror 9=2°or5(with vertical generatinglines)
X-ray film holder/filter inserter
pellicle
soft X-ray
streak camera
P650Xwithbi-lamellar optics. 16 Ip • mm"'.3-5psFIG. 1. Experimental set-up of the streaked soft-X-ray imaging device.

R.T.C., La Radiotechnique Compelec, 130, Avenue Ledru Rollin,
F-75540 Paris Cedex 11, France.

115

IAEA-CN-47/B-m-l

FIG. 2. Example of recorded streak image and corresponding calibration of the sweep speed.
Laser wavelength: \ = 1.06 pm; incident energy: Et= 740 J; FWHM laser pulse: At = 900 ps,
gold target: 200 jim in diameter; 200 \im in length; gold X-ray mirror; grazing incidence
angle: 6 = 2°; cut-off energy: hvc = 2 keV; magnification: 20.

photocathodeaccelerator

blanking

electrodequadrupolar lens
spatial focusing
temporalelectronbeam

deflection
line

streakimage.

FIG. 3. P750X image converter tube with lamellar optics.

116

BOURGADE et al.

static mode, using a palladium photocathode on a sapphire window and ultraviolet light, we measured the following performance without an intensifier:— Magnification along the spatial axis (photocathode to screen): 2— Resolution along the spatial axis: 32 Cp-mm"1 on the photocathode (~32 /zm)
16 Cp-mrrT1 over the whole screen (~62 jum)
— Resolution along the time axis on the screen (~60 fim).
Figure 4(a) shows a photograph of a resolution chart (25 Cp-mm"1) and
Fig. 4(b) shows a microdensitometer trace. The corresponding temporal resolution
is in the range of 2-3 ps. These results are in good agreement with calculated
theoretical values. A camera is under construction.
For two-dimensional images in the spectral range 50 eV to a few keV, wetested a Wolter microscope working in the soft X-ray range. For analysis of theimage, we used a modified R.T.C. image converter tube with an X-ray photocathode (500 A gold or Csl layer on a 3000 A thick plastic foil) and an output
fibre plate on which a P20 phosphor screen is deposited. Figure 5 shows the experimental arrangement on the target chamber. The main characteristics of the Wolter
microscope are: magnification: 10; microscope resolution: 40 Ep-mm"1, resolutionof the visible image: 13.3 £p-mm" 1 .
The 10-fold magnified image appearing on the phosphor screen is read by acharge-coupled device (CCD) working in the visible spectral range. The angular
position of the image converter tube is determined in order to separate the soft
X-ray image from direct X-ray emission [3].
In the 1-10 keV spectral range a pinhole camera (Fig. 6) is used with a CCD
recorder operating under vacuum. The image of the plasma through the pinhole
is recorded on the CCD with an adjustable magnification [4]. The television
camera produced by Thomson-CSF2 is in three parts. The CCD recorder operating
under vacuum in the chamber is connected to the head of the miniature camera bymeans of a 29-pin vacuum-proof connector. The CCD is driven by a television
camera produced by Thomson-CSF LER3; the circuits have been optimized inorder to obtain a very high dynamic range on the video signal without coolingof the CCD. We measured a dark noise of 0.7 mV peak to peak, for a saturationlevel of 700 mV, but the linear response in amplitude is only about 50-100,
depending on whether or not CCD polarization voltages are set for maximum
sensitivity. The immediate processing of the image is then performed by an
8 bit image memory developed by Numelec 4 .
Thomson-CSF Boulogne DTE, 38, rue Vauthier, B.P. 305,
F-92102 Boulogne, France.
Thomson-CSF LER, Avenue Belle Fontaine,
F-35510 Cesson Sevigne, Rennes, France.
Numelec, 1, place de lagare, La Verriere, B.P. 38,
F-78320 Le Mesnil Saint-Denis, France.

117

FIG. 4. Static measurements with ultra-violet light.
(a) Photograph of resolution chart [magnification: 10; 25 Hp-mrn'1);(b) microdensitometer trace.

optic libre
place
image
converter
tube

insulating adapter
bellows
Wolter microscope
plasmaf
fluorescentscreendetector
C film or CCD

FIG. 5. Experimental set-up of the whole appartus mounted on the interaction chamber.

BOURGADE et al.

118

Pinhole chamber CCD recorderpreamplifier
P|asmao-ringr r nCDPinhole?

to mainelectronics

atmosphere

FIG. 6. Pinhole camera with CCD recorder.

atmospheric_pressurepreamplifier
linear arrays r
4096 photodiodes
temperature
regulation —Peltier cellclock frequency 10 kHz
fibre optic cable 30 m
sweep controller
synchronizer.memory
oscilloscope display
RS 232 interface
graphics display
copierIEEE 488 interface

computerfiltervideo multiplexer
amplifier
12 bit digitizing

FIG. 7. Data acquisition system using linear arrays.

2.

X-RAY SPECTROSCOPIC AND ENERGY BALANCE MEASUREMENTS

For spectroscopic measurements, a great number of data are recorded on
the film. To obtain the results very quickly, we developed real-time acquisition
systems on an X-ray spectrometer, using a curved T1A phosphate crystal. This
spectrometer is used in the range from 1.3 to 2.8 keV, with cooled (5°C) linear
array detectors of 4096 photodiodes and a 12 bit digitization. Figure 7 shows
the data acquisition system using linear arrays. Figure 8 shows an example ofa spectrum obtained from an aluminium plane target at 0.53 nm, with
1.6 X 10 15 W-cm" 2 laser irradiation.
For accurate measurements of the energy balance, a new type of isoperibolic calorimeter has been developed and used, either with an NG1 glass absorberto obtain a broadband (0.3-2 //m) light measurement or with aluminium windows
0.5 mm thick for ions and X-rays. Figure 9 shows a schematic of the calorimeter,with a small thermopile of high sensitivity and a low heat capacity for the absorbing

IAEA-CN-47/B-m-l

119

FIG. 8. Aluminium spectra obtained from the X-ray spectrometer using linear arrays ofphotodiodes. Aluminium plane target; laser intensity: 1.6 X 10s W- cm~2, \ = 0.53 fim.

FIG. 9. Broadband isoperibolic calorimeter.
1 - absorbing material, 2 - heater equalizer, 3 - resistive layer, 4 — thermopile, 5 - cold source.

BOURGADE et al.

120

X-ray Him

X-ray PIN dlodo

_.

\7—-

10 punip_ 50 \im Mg
fluorescent target

fluorescentradiation
M g K a 1,254 keV

X-ray **ay
CCD X
mm PIN diodeFIG. 10. Experimental set-up for direct and fluorescent mode calibration with pulsed
soft X-ray source.

and reference material. With a resistive layer, absolute electrical calibration ispossible and a sensitivity of 2.68 mV-J"1 -cm"2 has been determined. Absolute
X-ray measurements have also been performed with a special X-ray bolometer
made of a 5 fim thick platinum grid. A sensitivity of 74 mV- mJ" 1 • cm"2 hasbeen determined by electrical calibration [5].

3.

X-RAY CALIBRATION

X-ray calibration is needed for many detectors. We used mainly continuous
X-ray sources, operating at 1.8 keV and 4-5 keV, and a pulsed source obtainedby modifying a classical coaxial plasma gun. Figure 10 shows the experimental
set-up of the pulsed X-ray source around the chamber. The energy radiated into
X-rays is absolutely measured with the X-ray bolometer. The X-ray energy
emitted from a neon gas discharge is about 20 J • ATI~X . The major part of the
X-ray emission is located between the He-like limit of free and bound neon ions
(1.2keV)and the K-edge energy of the filter (1.3 keV - 1.56 keV - 1.84 keV).

IAEA-CN-47/B-in-l
i

1

—,

121
B

.

•
•

/

/

E 500

O 400
/

$

" /

<d 300

H
O

•

/ /

•

direct calibration 1.4 koV•

fluorescent radiation 1.73 keV

iffluorescent radiation 1.25 keV

A continuous X-ray source 1.8 keV

>

r

1

2

3

4

5

PHOTON FLUX (10 8 photons-cm"2)

FIG. 11. X-ray calibration ofCCD arrays {Thomson TH 7882).

Direct measurements can be made, but calibration is also possible by using thefluorescent mode. Figure 11 presents a comparison between the various methods
for X-ray calibration of CCD arrays [6].

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

LAUNSPACH, J., et al., in High Speed Photography (Proc. Mtg. Strasbourg, 1984),
Society of Photo-Optical Instrumentation Engineers, SPIE 491 (1984) 685.
GIRARD, A., et al. (LEP, Paris), NOEL, J., et al (Hyperelec, Paris), SPIE 491 (1984) 58.
SAUNEUF, R., paper presented at 26th Meeting of the Americn Physical Society,
Boston, 1984.
CAVAILLER, C, et al., SPIE 491 (1984) 442.
BOURGADE, J.L., et al., C.R. Seances Acad. Sci., Ser. 2, 301 13 (1985).
LAUNSPACH, J., et al., "Direct X-ray response of charge-coupled devices and photodiode linear arrays", paper presented at 30th Tech. Symp. Optical and Optoelectronic
Applied Sciences and Engineering, San Diego, CA, 1986.

IAEA-CN-47/B-III-2

EXPERIMENTAL STUDIES OFRADIATION DRIVEN LASER FUSION
H.Z. SHEN, H.S. PENG, Z.J. ZHENG, J.G. YANG
Southwest Institute of Nuclear Physics and Chemistry,
Chengdu
S.J. WANG
Shanghai Institute of Optics and Fine Mechanics,
Shanghai
China

Abstract
EXPERIMENTAL STUDIES OF RADIATION DRIVEN LASER FUSION.
In a study of target inlet closure characteristics, plasma expansion processes wereobserved both temporally and spatially. The measured closure velocity can be given by
Ct = 10 I 0 ' 3 , where I is laser power density. Furthermore, the X-ray energy spectra ofplane and hohlraum gold targets were recorded; their equivalent peak radiation temperatures
were 115 eV and 98 eV. The results show that hohlraum targets are advantageous for laser
energy absorption and X-ray conversion.

1.

INTRODUCTION

Radiation driven compression is considered to be a very promising approach
to inertial confinement fusion (ICF) because of the uniformity and high efficiency
achieved [1 ]. To study basic phenomena, such as laser energy injection efficiency
into hohlraum targets, X-ray conversion efficiency and radiation temperature, we
conducted laser-target interaction experiments.
Plane pinhole targets were used to simulate experiments of laser injection
into hohlraum targets. Also the effects of laser energy, pulse duration, inlet
diameter and power density deposition at the pinhole edge on laser energy
injection efficiency were investigated [2, 3]. To study the characteristics oftarget inlet closure, the plasma expansion velocities were measured; they canbe given analytically by C t = 10 I 03 , where C t is the closure velocity in jum per
0.1 ns and I is the laser power density at the inlet edge in 10 14 W/cm 2 . Also a0-like X-ray spatial distribution was observed, which implies a new X-ray
conversion process.
Radiation spectra and energy balances were used for a comparison ofplane targets and hohlraum targets. A higher laser absorption efficiency anda broadened X-ray temporal spectrum of the hohlraum targets were found.
123

SHEN et al.

124
100

ff

80
; 60
40

(
20

t

1
f

i

E = 7 ± 3J(a) T = 3 0 0 i 100 ps(b) T = 700 ± 100 ps(C) T = 1000 ps

/A

50

100

150

D (Aim)FIG. 1. Energy injection versus hole diameter.

2.

APPARATUS

The experiments were conducted on a Nd:glass laser facility (0.1 TW,
0.2-3 ns, 1.06 ^m, 10 13 -10 15 W/cm2, focal spot diameter 40-70 jum). Plane
gold targets (25 /im thick), disk pinhole targets (10 iim thick, hole diameter
70-150 fim) and hohlraum targets (10 jum thick, spherical shell, 200 jum indiameter, with injection and diagnostic holes) were used with a target aiming
accuracy of 10 jam.
The energy, duration, spatial and temporal distribution, focal spot, andsignal-to-noise ratio of the laser beam were monitored with light calorimeters,
a streak camera, an optical array camera and photodiodes.
The X-ray energy spectra were recorded with a ten-channel Dante spectrometer, the X-ray temporal spectra with an X-ray streak camera, the X-ray spatial
distribution with pinhole cameras, and the X-ray flux with windowless X-ray
diodes. Plasma and light calorimeters and silicon photodiodes were employed
for energy balance measurements. In addition, in the case of plane pinholetargets, the incident and injected energies as well as the temporal distributions
were measured with light calorimeters and an optical streak camera in the energy
injection experiments.

3.

RESULTS AND DISCUSSION

The integrated laser energy injection efficiency T, as a function of laser
energy E, laser duration r, beam focal spot radius Ro and target hole radius R,

125

IAEA-CN-47/B-III-2

1.0

Injected/ laser peak<
Incident'
laser peak\- 0.5Vt (X187ps)

FIG. 2. Instantaneous injection efficiency-

studied with different parameters, The results (for E = 100 J) can besummarized approximately by

where E is in joules, R and Ro are in micrometres and r is in 0.1 ns [4], A
typical family of injection curves is shown in Fig. 1. The injection efficiency
is dominated mainly by the laser power density I deposited at the target inletedge. When I exceeded 1012 W/cm 2 , the plasma produced at the edge
expanded towards the hole and blocked the incident laser beam. Also, theinstantaneous injection efficiency T(t) was obtained by comparing andnormalizing the incident and injected laser waveforms and energies (see Fig. 2).
Assuming that the spatial distribution of the laser beam is Gaussian, we
express the function by
T ( t ) = 1 -exp((R-C t -t)/Ro) 2
where C t is the plasma expansion velocity. Then, C t can be derived from Fig. 2and scaled as C t = 10 I0-3, where C t is in fim per 0.1 ns and I is in 1014 W/cm 2 .
The integrated X-ray spatial distributions emitted from the pinholetargets were recorded and 0-like images with a bright spot in the middle wereobserved. Figure 3 shows a typical X-ray density distribution scanned acrossan image. The peak in the middle corresponds to the bright spot. It is easy
to find a possible explanation. If the edge of the hole is uniformly irradiatedby a laser beam with sufficient energy, the plasma expands inwards. When theplasma reaches the centre of the hole, a part of the kinetic energy is converted
to interior energy, resulting in higher temperature and higher density. Therefore,

SHEN et al.

126

E = 6.9J
T = 207 ps

J

D = 140/jm

/

I

J1/

u

T = 55%\

ywWkv*»

111*86~

F/G. 5. Spatial X-ray distribution of a pinhole target.

10 15

Hohlraum target
E = 4OJ
T = 800 ps>

Plane target
E = 21JT = 900 ps

V

10 14

10 13

0.2

0.4

0.6
keV

0.8

1.0

1.2

FIG. 4. Radiation energy spectra.

Hohlraum target

8

10

tins)
FIG. 5. Temporal X-ray spectra.

12

IAEA-CN-47/B-m-2

127

more X-rays can be emitted from the denser plasma. However, more theoreticalwork is needed to obtain a confirmation for the X-ray production mechanism.
Figure 4 shows low energy X-ray spectra, detected with the Dante spectrometer, for plane and hohlraum targets. The corresponding equivalent peaktemperatures are 115 eV and 98 eV. The spectral distribution is very similar
to a Planck distribution, but the two wings of each curve are slightly higher.
Temporal X-ray spectra recorded with windowless X-ray diodes are shown in
Fig. 5. The spectrum of the hohlraum target is somewhat longer than thatof the plane target, which implies a longer 'restrain' time and possibly a higher
X-ray conversion efficiency.
The energy balance measurements indicated that about 70% laser energy
was absorbed by the hohlraum target.

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

YAMANAKA, C, et al., in Plasma Physics and Controlled Nuclear Fusion Research
1984 (Proc. 10th Int. Conf. London, 1984), Vol. 3, IAEA, Vienna (1985) 3.
AUERBACH, J.M., et al., J. Appl. Phys. 18 (1979) 2496.
YAMADA, K., et al., J. Phys. Soc. Jpn. 51 (1980) 280.
ZHANG, J., GU, J.P., personal communication, 1986.

IAEA-CN-47/B-HI-3

RESEARCH ON HEAVY ION DRIVERS
FOR INERTIAL CONFINEMENT FUSION
IN THE UNITED STATES OF AMERICA*
C. CELATA, A. FALTENS, T.J. FESSENDEN,
D.L. JUDD, D. KEEFE, C.H. KIM,
L.J. LASLETT, E.P. LEE, L. SMITH,
M.G. TIEFENBACK, A.I. WARWICK
Lawrence Berkeley Laboratory,
University of California,
Berkeley, California,
United States of America

Abstract
RESEARCH ON HEAVY ION DRIVERS FOR INERTIAL CONFINEMENT FUSION IN THE
UNITED STATES OF AMERICA.
The U.S. study of high-energy multi-gap accelerators for the production of large currents ofheavy ions for inertial confinement fusion is centered on the single-pass induction linac method. The
large technology base associated with multi-gap accelerators for high-energy physics gives confidence
that high efficiency, high repetition rate and good availability can be achieved, and that the path fromscientific demonstration to commercial realization can be a smooth one. In an induction linac driver,
multiple (parallel) ion beams are accelerated through a sequence of pulsed transformers. Crucial to thedesign is the manipulation of electric fields to amplify the beam current during acceleration. A proof-ofprinciple induction linac experiment (MBE-4) is under way and has begun the first demonstration ofcurrent amplification, control of the bunch ends, and acceleration of multiple beams. A recently completed experiment, called the Single Beam Transport Experiment, has shown that one can now counton more freedom to design an alternating-gradient quadrupole focusing channel to transport much higher
ion beam currents than formerly believed possible. This result, together with the recent developmentof efficient sources for multiply charged ions, has rekindled interest in exploring drivers with heavy ionsof charge state greater than unity. A recent study on Heavy Ion Fusion Systems Assessment (HIFSA)
has shown that a substantial cost saving results from the use of multiply charged ions and that a remarkably broad range of options exist for viable power plant designs. The cost of a driver at 3-4 MJ could
be US $200 per joule or less, and the cost of electricity could be in the range of 50-55 million US $per kilowatt-hour. A next-generation experiment is needed to address the questions of high-power beamtransport by magnetic lenses and of focusing the ions on a spot of about 1-2 mm radius, in a stable,
reproducible way.

Work supported by the Office of Energy Research, Office of Program Analysis andOffice of Basic Energy Sciences, United States Department of Energy, underContract No. DE-ACO3-76SF0OO98.

129

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CELATA et al.

THE INDUCTION LINAC APPROACH
Inertial confinement fusion (ICF) requires very high power
irradiance, and energy deposited on the fusion target,which are nearly
independent of the driver type. For ions, the depth of deposition mustbe small (typically ~ 0.1 gm/cm2 in a stopper material) to produce thehigh fusion yields required for an economically attractive power
plant. The range condition can be met in principle by any ion species,
accelerated sufficiently to match the range-energy relation.as well as
by short wave-length (< 300nm) photons. For the heavy-ion driver
approach to ICF, two conventional but very high current accelerator
technologies are being explored. These are the rf linac/storage ringsystem now studied in W.Germany, USSR and Oapan, and theinduction linac approach of the USA. For both accelerator types the
combined considerations of space charge limits and range in dense
matter lead to the use of heavy ions (A *» 200) of high kinetic
energy («* 10 GeV).
A typical set of final beam parameters suitable for a powerreactor [1] is given in Table I. It must be emphasized that cost
tradeoffs among the many components of a complete power plantallow a broad range of system parameters (such as repetition rate) tobe considered with minor effect on the final cost of electricity.
An induction linac driver is now envisioned as a multiple beamlettransport lattice consisting of N closely packed parallel channels.
Surrounding the lattice are massive induction cores of ferromagnetic
material and associated pulser circuitry which apply a succession of
long-duration, high-voltage pulses to the N parallel beamlets. A
multiple beam source of heavy ions operates at 2-3 MV, producing thenet charge per pulse required to achieve the desired pellet gain.
Initial current (and therefore initial pulse length) are determined bytransport limits in the lattice at low energy. The use of a large
number of electrostatic quadrupole channels (N ~ 16 - 64) appears
to be the least expensive option at low energies (below ~ 50 MV). Thisis followed by a lower number of superconducting magnetic channels(N ~ 4-16) for the rest of the accelerator. Combining of beams may
therefore be required at this transition. Furthermore, some splittingof beams may be required after acceleration to stay within current
limits in the final focus system.
The rationale for the use of multiple beams is that it increases
the net charge which can be accelerated by a given cross-section ofcore at a fixed accelerating gradient. Alternatively, a given amountof charge can be accelerated more rapidly with multiple beams since
the pulse length is shortened and a core cross-section of specified
volt-seconds per meter flux swing can supply an increased gradient.
However, an increase in the number of beamlets increases the cost anddimensions of the transport lattice and also increases the cost of thecore for given volt-sec product since a larger core volume is required.

IAEA-CN-47/B-IH-3

131

TABLE I. EXAMPLE DRIVER PARAMETERS
Pulse Energy

5.0 MJ

Particle Energy

lOGeV

Particle TypeB* (A = 209)

Pulse Power250 TW

Pulse Length

20 ns

Rep. Rate per Reactor

5 Hz

Number of Beams per Reactor

20

Net Pulse Charge

500 yC

Emittance (unnormalized)

3xl0" 5 m-r

Momentum Width± 1%

Spot Radius

4 mm

For a core of given cross-section (a volt-seconds/m), the volume offerromagnetic material increases as its inside diameter is increased.
Hence there is a tradeoff between transport and acceleration costs,
with an optimum at some finite number of beamlets. The
determination of this optimum configuration is a complex problem,
depending on projected costs of magnets, cores, insulators, energy
storage, pulsers and fabrication. The induction linac design code
LIACEP [2] is used for this purpose.
The choice of superconducting magnets for the bulk of the linac
is mandated by the requirement of system efficiency; this must be atleast ~ 10% in an ICF driver and ideally > 20% to avoid large
circulating power fractions, which result in a high cost of electricity(COE). Induction cores are most likely to be constructed from thinlaminations of amorphous iron, which is the preferred material,due toits excellent electrical characteristics and flux swing. At a projected
cost of ~ A $/lb (insulated and wound) this is a major cost item for the
first 2-4 GV of a typical linac. At higher voltage the cost of pulsers
and fabrication of the high gradient column with insulators dominate.
THE MULTIPLE BEAM EXPERIMENT, MBE-4
MBE-4 is a four-beam induction linac under construction andpreliminary operation at LBL that models much of the accelerator
physics of the electrostatically focused section of a considerably
longer induction accelerator [3]: Four parallel Cs+ beams extracted
from thermionic alumino-silicate sources in a 2.5 \xs pulse and injected

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CELATA et al.

into the linac at 20D keV. Acceleration to approximately one MeV will
be achieved when the experiment is complete in the summer of 1987.
The current in each of the four beams will increase from 10 to 40 mA
during acceleration, due to both increase in beam speed and shorteningof the bunch length. As a percentage of beam energy, theacceleration voltages in MBE-4 are much larger than will be used in adriver. Therefore, the consequences of errors in acceleration voltages
will be more apparent and more easily assessed. Four beams are used
to investigate potential effects caused by beam-beam coupling and to
get practical experience in difficulties associated with accelerating
and transversely controlling parallel beams. In examining the scaling
with injection energy and with quadrupole size, we have been careful
to preserve space charge domination of the beams both transversely
and longitudinally. Measured in terms of initial bunch lengths, MBE-4
is comparable in length to the electrostatically focused portion of adriver.
The voltage waveforms on the first accelerator gaps are nearly
triangular so as to impart an axial velocity shear or tilt to the beam byaccelerating the tail of the beam more than the head. After the tailof the beam has entered the accelerator, the head of the beam can be
accelerated as well and the waveforms on the downstream gaps can beessentially flat [A].
To
date, MBE-4
hasaccelerated
fourparallelspace-charge-dominated cesium ion beams from 0.2 MV to .54 MV,
with current amplification of approximately 2.7. The experiments are
in excellent agreement with our theoretical acceleration model. The
acceleration schedule is one in which the current waveforms grow inamplitude and decrease in pulse duration in a self-similar or
self-replicating way with acceleration distance. Aside from small
fluctuations, generated by 2-3% errors in the acceleration waveforms,
the experimental current waveforms are self-replicating. The
acceleration errors are mostly corrected by "trim" pulsers located atevery fourth accelerating gap. These are also used to control spacecharge spreading of the bunch ends. At large currents, rapid current
fluctuations are damped by space charge repulsion.
Experiments have demonstrated only a very weak transverse
electrostatic repulsive interaction among the beams in the source
area, which is easily corrected.

SINGLE BEAM TRANSPORT EXPERIMENT
A recently completed experiment, called the Single BeamTransport Experiment (SBTE) [5], has shown that we now can have thefreedom to design an alternating-gradient quadrupole focussing
channel to transport ion-beam currents much higher than formerly

IAEA-CN-47/B-III-3

133

believed possible. SBTE employs a ID mA, 120 keV Cs+ beamtransported by 87 electrostatic quadrupoles. In the presence of spacecharge, the transverse betatron oscillation frequency of a particle, u,
is lower than the value for little or no space-charge, &>0, and is given
approximately by the expression222
w = w - <«>/2
where <jp is the beam plasma frequency. Instabilities have previously
been predicted in certain ranges in (to,«oo) space and could result inloss of beam current or damage to the optical quality. The SBTE hasestablished that there is significantly more latitude to choose thefocussing parameters, u> and u 0 , than was previously suspected; inparticular, values of M/UQ as low as D.I can give stable transport. A
beam with this low a value is "space-charge dominated", and the
Debye length is very much less than the beam radius. These resultsfrom SBTE have rekindled interest in exploring drivers with heavy ionsof charge state, q, greater than unity, which, in general, will require
operation at lower values of «/<oo.

HEAVY ION FUSION SYSTEMS ASSESSMENT STUDY
The Heavy Ion Fusion Systems Assessment (HIFSA) study [6] wasconducted during 1984-86 with the specific objective of evaluating theprospects of using induction linac drivers to generate economicalelectrical power from inertial confinement fusion. Principal
contributors were LBL, L.LNL, LANL, McDonald-Douglas Astronautics
Co., U. of Wisconsin, the U.S. Department of Energy and EPRI. The
study used algorithmic models of representative components of afusion system to identify favored areas in the multidimensionalparameter space. The resulting cost-of-electricity projections arecomparable to those from other (magnetic) fusion scenarios, at a plantsize of 1000 MWe. These results hold over a large area of parameter
space, but depend especially on making large savings in the cost of theaccelerator by using ions with charge-to-mass ratio about three times
higher than has been usually assumed. The feasibility of actuallyrealizing such savings has been shown: (1) by experiments showing
better than previously assumed transport stability for space charge
dominated beams, and (2) by theoretical predictions that the final
transport to the target pellet in the expected environment of a reactor
chamber may be sufficiently resistant to streaming instabilities.
Neutralization in the chamber is required for the higher current pulsesthat result from the use of the higher charge-to-mass ratio beams.
The HIFSA study was organized to deal with the specific premisethat fusion in general, and the HIF approach to ICF in particular,
appears to be so costly and requires scaling to such large power plants

134

CELATA et al.

that it has not been possible to design a program that would be
attractive to the electric utility industry. The objective of the study
was to perform an assessment of heavy ion inertial fusion systems
based on induction accelerators, including representative reactor
systems, beam focusing and final transport, target design, and system
integration. Emphasis was given to systems for electric power
production and to design innovations and parameter ranges which offer
credible promise of reducing system size and cost. Effort
concentrated on system and subsystem conceptual design and analysis,
including cost/performance models for studying and exhibiting major
system parameter variations. Identification of needed R&D was
included.
Two important computational tools were developed for the
study: The linac optimization program LIACEP was extensively
rewritten, and the system program ICCOMO was written to permit
examination of large areas of commercial plant parameter space tofind local optima.
Probably the most important technical results of the study camefrom reexamining previously suggested cost-saving ideas [7], These
ideas, modified by newer experimental results, make it possible toenvision very significant cost reductions by (especially) using higher
charge-to-mass ratios. Most of the study was done for q = + 3,
A = 130.
Some important conclusions are:
(1)
(2)
(3)
(4)
(5)

COE is insensitive to repetition rate in the optimum range
3-9 pps.
Symmetric targets, which may use the beam energy more
efficiently, do not result in lower COE because of the increased
cost of the transport system.
High gain targets (G > 100) are only of moderate utility since
they increase the cost of the reactor vessel and their effect canbe partly realized by higher rep. rate.
Major reductions in the cost of induction linacs result from using
higher charge-to-mass ratios and multiple beams.
With the higher charge-to-mass ratios it is certainly necessary to
invoke neutralization during final transport. Work by Stroud [8]gives confidence that streaming instabilities will not destroy theemittance during transport through the target chamber.
ILSE:

A PROPOSED INDUCTION LINAC SYSTEMSEXPERIMENT

Experiments to date, outlined above, lend confidence thatseveral of the conditions and manipulations required in a fusion driver

IAEA-CN-47/B-HI-3

4MU
25 R

2 MU
is n
I us
INJECTORo135

O.V JJS

BEND

COMBINE
ELECTRIC
FOCUS

MflGNETIC FOCUS
flCCELEHBTOR

flCCEL.

I - 16 beams

4 beams
DRIFT-COMPRESSION
1 beamFIG. 1. Schematic of a fixture induction linac systems experiment.

can be accomplished. Nonetheless, several other manipulations remain
to be studied experimentally and will require ions of higher velocity.
Thus we are proposing an experiment to study the physics ofacceleration, combining and focusing ion beams at a level of current
and power considerably exceeding thatof MBE-4. In the preliminary
design, 16 parallel beams of ions (C + , Al + , or Al ++ ) produced from a
2 MV injector [9] will be accelerated to several MV and combinedtransversely. The four resulting beams are then further accelerated to~ 10 MV and the growth in transverse and longitudinal emittance isdetermined for comparison with theory. The apparatus will then be
used to study the problems associated with focusing ion beams to asmall spot. ILSE is designed to address the following physics andengineering objectives:
(1)
(2)
(3)
(4)
(5)

Examine the physics of transversely combining space-chargedominated ion beams.
Explore the transition from an electrostatic to a magnetic beamtransport system.
Examine the physics of bending ion beams with intense spacecharge and with time dependent energy.
Study the physics of drift-compression current amplification.
Explore the focusing of intense ion beams to a 1-2 millimeterspot.

In addition, ILSE will advance the technology of heavy ion
multiple-beam induction linacs much closer to that needed for adriver. Magnetic focusing elements are used after the point wherebeams are combined to the final focal spot. These desired features
motivate the choice of 10 MV of acceleration and a relatively lightion, which simulate a heavier ion at higher kinetic energy. A block
diagram of a conceptual design for ILSE is presented in Fig. 1.

136

CELATA et al.

REFERENCES
[1]
[2]
[3]

[4]
[5]

[6]
[7]

[8]
[9]

HIBALL-H, Rep. FPA-84-4, Fusion Power Associates (1984).
FALTENS, A., HOYER, E., KEEFE, D., LASLETT, L.J., Rep. LBL-8357, Lawrence
Berkeley Laboratory, California University (1979).
FESSENDEN, T.J., JUDD, D.L., KEEFE, D., KIM, C.H., LASLETT, L.J., SMITH, L.,
WARWICK, A.I., Rep. LBL-21348, Lawrence Berkeley Laboratory, California University(1986).
KIM, C , SMITH, L., Part. Accel. 18 (1985) 101.
CHUPP, W., FALTENS, A., HARTWIG, E.C., KEEFE, D., KIM, C.H., LASLETT, L.J.,
NEMETZ, R., PIKE, C , ROSENBLUM, S.S., SHILOH, J., SMITH, L., TIEFENBACK, M.,
VANECEK, D., Rep. LBL-15132, Lawrence Berkeley Laboratory, California University(1983).
DUDZIAK, D.J., HERRMANNSFELDT, W.B., "Heavy ion fusion systems assessment
study", Proc. Heavy Ion Fusion Symp., Washington, DC, May 1986.
FALTENS, A., HOYER, E., KEEFE, D., "A three-megajoule heavy ion fusion driver", HighPower Electron and Ion Beam Research and Technology (Proc. 4th Int. Top. Conf. Palaiseau,
France, 1986).
STROUD, P., "Heavy ion beam transport in ICF reactor chambers", Proc. Heavy Ion Fusion
Symp., Washington, DC, May 1986.
BALLARD, E.O., MEYER, E.A., OONA, H., RIEPE, K.B., RUTKOWSKI, H.L.,
SHURTER, R.P., VAN HAAFTEN , F.W., WILSON, D.C., HUMPHRIES, S., Jr., "Design
status of heavy ion injector program", High Power Particle Beams (Proc. 6th Int. Conf. Kobe,1986), Laser Society of Japan (1986).

IAEA-CN-47/B-in-4

MICROWAVE SIMULATIONOF LASER-PLASMA INTERACTIONS
C.J. PAWLEY, N.C. LUHMANN, Jr.
University of California,
Los Angeles, California
W.L. KRUER
Lawrence Livermore National Laboratory,
Livermore, California
United States of America
Abstract
MICROWAVE SIMULATION OF LASER-PLASMA INTERACTIONS.
The paper presents a detailed comparison between experiment and computer simulation of thenonlinear saturation of ion acoustic waves produced by the optical mixing of two intense electromagnetic
waves. For ne/n,, < 0.2, the saturation is found both in experiment and in simulation to be due
primarily to ion tail heating. For n ^ = 0.3-0.5, the first quantitative direct measurements of pumpdepletion reveal that it is as important as ion tail heating in determining saturation.

1. INTRODUCTION
We have employed microwave scale experiments to study laser-plasmainteractions1"3 in plasmas where diagnostics with the required time and spatial resolution are available. Herein, saturation of the stimulated Brillouinscattering (SBS) instability has been studied by exciting the ion waves by
the optical mixing of two electromagnetic waves when the frequency andwavenumber differences (coemi — coem2 = Aco, k ^ i - k g ^ = Ak) satisfy theplasma dispersion relation ep(Aco,Ak) = o^2>4>5^. In addition to interest in
this process from both an ICF as well as a basic plasma physics point ofview, there have also been proposals to employ optical mixing for a variety
of applications, including plasma diagnostics, beat-wave heating, tokamak
current drive and charged particle acceleration ^ .
The experimental results have been compared to a variety of theoretical and numerical models. At sufficiently low power that nonlinear saturation processes are not important, the results are well described by the solution of the coupled two-fluid Maxwell equations when both collisional andkinetic damping are accounted for^. For higher incident powers, twoapproaches have been employed. The first is to employ a 1-D particle ion
137

138

PAWLEY et al.

and fluid electron code where actual experimental effects including ion neutral collisions are incorporated. In this approach the electron density isgiven by n e = n ^ exp —^- + a

where a is the ponderomotive potential.

Currently, a is taken to be constant so that pump depletion effects whichbecome important for ng/nc> 0.2 are neglected. The second numerical code
employs a semi-empirical approach which utilizes energy balance to accountfor ion tail formation at the resonant phase velocity. This solves for the
spatial ion and electromagnetic wave distributions in the time asymptotic
steady-state and is an extension of the work reported by Heikkinen et al.^.
The coupled Vlasov-Maxwell equations are solved, with ion wave harmonic
generation, collisions, and tail formation included, together with variations in
the electromagnetic pump strengths (i.e. pump depletion).

2. EXPERIMENTAL ARRANGEMENT
Two electromagnetic waves (cOj/271, v^Hn = 3.3 GHz) with parallel
polarization propagate antiparallel to each other along the axis of a fieldfree helium plasma (2m length, 75 cm diameter, T e = 3-4 eV, rYeITi =20).dnewith —and ——> 10m. A collective millimeter wave
ne dz
[Te dzThomson scattering system permits fluctuations as small as 3xlO6 cm 3 tobe observed (n/ne = 10"4) during the high power microwave pulses.

3. EXPERIMENTAL RESULTS
As shown in Fig. 1, both Langmuir probe and millimeter wave scattering measurements indicate that the ion wave amplitude saturates at = 5-6%
for an 0.15 nc density plasma, even for power levels considerably higher
than employed in Ref. [2]. This level is significantly below that predictedfrom simple linear damping. The density dependence of the saturated amplitude was also investigated and, as shown in the inset to Fig. 1, the amplitude was found to decrease dramatically as the density approaches 0.5 nc.
For both of the above cases the theoretical curves will be discussed later.
Faced with the above saturation observations, our approach has been to
examine as many of the candidates as possible and to thereby identify the
primary saturation process(es). The first is simply inverse Bremsstrahlung
absorption of the incident microwaves due to the neutral background, caus-

139

IAEA-CN-47/B-III-4

SIMPLELINEAR
PREDICTIONVSPATIAL
CODE.PREDICTION--•— _

S
•

5
4

•S* 3
K
2

SIMULATIONQA EXPERIMENTSo

1

2 FLUID CODE
0.2

(NO TAIL)
i

20

o 5

40

60
(P., P 2 )

80
%

0.3

0.4

i n"/nc

i

100

120

(kW)

FIG. I. Dependence of ion wave fundamental on microwave power. The circles are from scattering,
the triangles from probes, the crosses from 1-D simulations; the - • - curve is from computer calculations using linear damping and thecurve includes ion tail formation.
Inset: Dependence of ion wave amplitude on plasma density as determined from scattering data at thecenter of the experimental chamber (7 m from the highest frequency horn) for PHF = 100 kW,
Pjj. = 70 kW. The solid curve is the computer prediction, including ion tail formation and harmonic
generation.

ing an increase in T e and therefore a decrease in vo/vte = eE/mcovte. However, we have investigated this effect previously^1'3} and find that at low
densities (< 0.1 nc) it is negligible. Even for ne/nc = 0.5 the maximum
change is only from 0.23 to 0.16, which is not sufficient to explain theobserved ion wave amplitude. The primary effect is to change the resonant
ion wave frequency. However, we have partially compensated for this shiftby employing the final resonance value for the microwave difference frequency. An additional effect of inverse Bremsstrahlung is to suppress harmonic generation because the change in the driven frequency increases withthe harmonic number.
A second possible saturation mechanism involves ion acoustic wave
harmonic generation P\ Here, the optical mixing generated ion wave nonlinearly generates a harmonic as it grows in amplitude. The harmonics do
not couple to the electromagnetic waves and therefore represent a loss
mechanism to the driven fundamental. Experimentally, we have observed
harmonic generation both with Langmuir probes and scattering, with sample
scattering signal data displayed in Figs 2(a) and 2(b). Here, Gs = 4°(8°)
corresponds to scattering from the fundamental (harmonic). The harmonic

PAWLEY et al.

140
(a)

Ib)

f-—r.f. on—\-r.f. o n ^ |

10/is
H

8°
6°
4°
2°

wvvvVVA ¥ VWW^

30

(c)

g 20

20

40

60

80

100

(P1P2)Y= (kW)FIG. 2. (a) and (b) Time history of scattering signals for (P,PJ* = 14 kW and Af = 174 kHz (a)and {Pf^= 100 kW and Af = 210 kHz (b), where njnc = 0.15. The effect of inverse Bremsstrahlung absorption is obvious in (b), as the ion wave frequency is near 170 kHz at the beginning of
the rf pulse but close to 210 kHz at the end of the rf pulse.
(c) Dependence of fl2ffii on incident microwave power (computer predictions). The solid curvescontain only the second harmonic term and the dashed curves contain both the second and third
harmonic terms. Circles and triangles are data points.

content at 6S = 8° is obvious. The reason for the overlap (i.e. detectingsome fundamental at 8S = 8°) is due to the fact that the angular distributionof scattered radiation is given by a Gaussian. Although harmonic generation
is observed, we find that it does not explain the observed saturation.
Specifically, as shown in Fig. 2(c), we find that as the input level isincreased, the second harmonic also saturates., therefore eliminating it as asole mechanism for clamping the fundamental. Additionally, the third harmonic amplitude is comparable to the experimental noise level and does notincrease in level, eliminating the possibility of increased higher harmonics.
However, we find that the inclusion of harmonic generation in our theoretical calculations permits a more detailed quantitative comparison between
theory and experiment. Similar behaviour is observed in the 1-D simulations with both the fundamental and second harmonic saturating, as shownin Fig. 3. As in experiments,the ratio ffj/nj is observed to become constantalthough the value is at the extreme limit of the experimental observations(see Fig. 4).

141

IAEA-CN-47/B-m-48n,r

Jlp I*t• 2 5.8%

V[ft. V A,

I

/

1
/

Sn2

/
/
2

3

t

o i i a t = 180

CO:,t = 180

2000

2000

5. Time history of ion wave fundamental and second harmonic obtained in 1-D simulations, using= 84 kW, k\De » 0.03, T,/Te = 0.04 and damping is 0.07.

Amplitude dependent frequency shifts due to ion trapping^7! or the
fluid dispersive terms (Korteweg-de Vries equation)^1 have also been proposed as saturation mechanisms. However, we do not observe this.
Specifically, the observed change in ion wave frequency is completely
explained by the change in Te due to inverse Bremsstrahlung absorption.
Another saturation mechanism is profile modification &\ However, measurements have shown that the modifications are completely negligible.
The remaining saturation mechanism at low densitites (ne/nc < 0.2) is
ion tail formation. Here it should be noted that for our experimental parameters the ion tail can be formed by the trapping of the ion wave only if thefinite harmonic content of the ion waves is included. The harmonic content
steepens the wavefronts, causing an increase in the wave electric field at thesteepened front. Such a tail formation is observed both in the simulations
and the experiment. In the experiment we have performed both scatteringand high resolution microchannel plate ion energy analyzer measurements todetermine the ion tail density. The ion tail density was determined from thescattering data in the following way. First, it was found that the angular
distribution of the scattered radiation broadens with increasing power, as
illustrated in the inset to Fig. 5 for the case of an 0.15 nc plasma, where thesolid curves are Gaussian fits to the scattering data which have been normalized at the same peak value to explicitly show the broadening. Afterunfolding the known (i.e. measured) instrument function, the dependence of142

PAWLEY et al.

0.5 -6 30 cm FROM HIGHEST FREQUENCY HORN
0 50 cm FROM HIGHEST FREQUENCY HORN
• SIMULATION

0.4

f

0.3 -

•

•
O

0.2

"•

"

n A° ° O

O

0

*A

$

OO

0.1
0

I

l

1

I

20

40

60

80

(p

i

i

i

i

100 120 140 160
1/a

e Ph)

(kW)

FIG. 4. Comparison of experimentand 1-D simulation ratios o

the width of the wavenumber distribution on microwave power is obtained.
If one then assumes that the width is related to the wave damping (i.e. Imk), the dependence of the damping on input power is thereby obtained.
Using the above inferred value for the total damping and subtractingout the linear damping, one obtains the nonlinear damping from which one
can infer the tail density. This inferred value is shown in Fig. 5 together
with the measured ion tail density obtained by energy analyzer and themeasured ion wave amplitude. Here the dashed and solid curves are thepredictions of the computer model described earlier; the agreement is seen
to be quite good. For densities ne/nc < 0.2, we therefore find that ion tailheating provides the saturation mechanism which results in the observed =5-6% ion wave amplitude. In addition, the formation of an ion tail strongly
inhibits harmonic growth since the Landau damping contribution from thetail is proportional to co.
Referring back to Fig. 1 we see that the predicted ion wave amplitude
from the 1-D simulations does not exhibit a strong clamping as observed
experimentally but rather continues to slowly increase with increasing
power. This may be due to the way in which particle collisions weretreated; in the simulations, all particles were assumed to experience the same
collision frequency. However, preliminary investigations seem to indicatethat the trend is in the wrong direction to explain the discrepancy.
The decrease in ion wave amplitude with increasing plasma density(see Fig. 1 inset) cannot be explained simply by ion tail formation. Hereone must also account for pump depletion due to the change in action of thepumps (see Refs [4 and 5]). Including this in our computer calculations
results in the solid curve shown in the Fig. 1 inset which is in good agreement with the experimental observations. A further check on this conten-

143

IAEA-CN-47/B-IH-4

12 i-

j

60

80

100

120

140

160

(P-,P 2 ) y2 (kW)FIG. 5. Dependence of ion tail density and ion wave fundamental on incident microwave power. The
open and solid circles are from scattering, triangles are from the energy analyzer and curves are fromthe computer.
Inset: Normalized dependence of scattered power on scattering angle for various power levels. Eachcurve is a Gaussian fitted to the data; the harmonics have been removed by filtering. The instrument
function has not been removed from these curves.

tion is found by calculating the theoretically predicted decrease in the highfrequency pump amplitude in one transit of the chamber and comparing itto the measured transmission, where good agreement is obtained after
correcting for the effective portion of the em waves actually involved in the
mixing process.

4. SUMMARY
We have observed the nonlinear saturation of ion acoustic waves
driven by optical mixing. At low densities (ne/nc < 0.25) the dominantsaturation mechanism is ion tail formation, while at higher densities(ne/nc = 0.3-0.5) pump depletion also plays a major role. Good agreement
was obtained between experiment and 1-D simulations for ne/nc = 0.1. Ion
wave harmonic generation is observed but found not to be a primary saturation mechanism, in contrast to some previously reported CO 2 laser experiments. It is of further interest to note that our low density optical mixing
saturation results were in complete agreement with our previous microwave

144

PAYVLEY et al.

SBS studies'3^ lending further strength to our contention that the opticalmixing results provide insight into SBS saturation.
Current work is focussed on the use of optical mixing to study thesaturation of both forward and backward stimulated Raman scattering.
Related work focusses on ExB acceleration of electrons at the critical layerof a plasma, with a DC magnetic field applied transverse to the density gradient and electric field of the P-polarized microwave beam.

ACKNOWLEDGEMENTS
The authors wish to gratefully acknowledge helpful conversations with
E.M. Campbell and K. Estabrook. The work was supported by theLawrence Livermore National Laboratory Laser Fusion Program throughU.S. DOE Contract W7405-ENG-48.

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

HUEY, H., et al., in Laser Interaction and Related Plasma Phenomena, Vol. 6, Plenum Press,
New York (1984) 605.
PAWLEY, C.J., HUEY, H.E., LUHMANN, N.C., Jr., Phys. Rev. Lett. 49 (1982) 877.
MASE, A., et al., in Plasma Physics and Controlled Nuclear Fusion Research 1980 (Proc.
8th Int. Conf. Brussels, 1980), Vol. 2, IAEA, Vienna (1981) 745.
COHEN, B.I., Comments Plasma Phys. Controll. Fusion 8 (1984) 197.
COHEN, B.I., et al., Phys. Fluids 18 (1975) 470.
HEIKKINEN, J.A., KARTTUNEN, S.J., SALOMAA, R.R.E, Phys. Fluids 27 (1984) 707.
KRUER, W.L., ESTABROOK, K.G., in Laser Interaction and Related Plasma Phenomena,Vol. 5, Plenum Press, New York (1979) 783.
CASSANOVA, M., et al., Phys. Rev. Lett. 54 (1985) 2230.
KRUER, W.L., VALEO, E.J., ESTABROOK, K.G., Phys. Rev. Lett. 35 (1975) 1076.

IAEA-CN-47/B-HI-5

PROGRESS IN
INERTIAL CONFINEMENT RESEARCHAT KMS FUSION, INC.*
J.T. LARSENKMS Fusion, Inc.,
Ann Arbor, Michigan,
United States of America
Abstract
PROGRESS IN INERTIAL CONFINEMENT RESEARCH AT KMS FUSION, INC.
KMS Fusion (KMSF) is active in the areas of ICF target development and fabrication, theoreticallaser-plasma interaction physics, computational physics and high power laser experiments on sphericaland planar targets. KMSF has developed fabrication techniques to produce high quality polyvinyl
alcohol (PVA) targets for high compression experiments, which are used primarily with D2 fuel.
KMSF also develops other polymers which can resist degradation from the tritium decay and havepromise for holding DT gas for an extended period. A cryogenic target apparatus, similar to the one
in use at KMSF, has been built and delivered to the University of Rochester to enable them to conductcryogenic target experiments. Unique experiments on the hydrodynamic behavior of planar exploding
foil targets have been performed. Four two-dimensional images of the expanding plasma can be madein the course of a single experiment with a multi-frame holographic interferometer. Major experiments
were conducted in 1985-86 to study the physics of preheat control in cryogenic fuel implosions. For
more than ten years, KMSF has been developing the techniques for producing implosions with preheat
levels in the 1-3 eV range. This requires short wavelength laser light, shaped laser pulses, uniform
spherical illumination, polymer shells with cryogenic fuel layers, and X-ray backlighting to record the
implosion history. Compressed fuel densities approaching 40 g/cm 3 have been achieved with less than
200 J of laser light.

LOW PREHEAT IMPLOSION EXPERIMENTS
In 1976, KMSF began to experiment with implosions driven
with frequency doubled 1.06 pm laser light, polymer fuel capsules, and cryogenic fuel configurations. Several series ofexperiments during the ten years perfected each of these
ingredients independently. In 1982 we demonstrated an initial
success in combining all these features.
The purpose of the experiment series was to demonstrate
improved fuel compression by carefully controlling the temperature the fuel was brought to prior to performing PdV work. It
is energetically advantageous to keep the fuel near its Fermi
Report prepared for the United States Department of Energy under Contract
No. DE-AC-08-82DP40152.

145

146

LARSENw

TargetAspheric Lens+

Corrector Window
Double Bounce
Cavity Mirror

FIG. 1. Uniform spherical illumination is achieved from two laser beams with a double bounce
configuration. Each 18.6 cm laser beam passes through anf/1.39 lens and into a two-mirror system.
A reflection from each surface puts the marginal ray at an angle of70° to the axis of rotation. A slight(about 20 urn) shift of the focal point behind the target's center provides uniform coverage. All
diagnostic instruments and target handling access is through the 5 cm annular gap between the twomirrors.

degenerate temperature. Preheat energy can be deposited in thefuel by three mechanisms: shock, electron deposition, and, to alesser extent, radiative processes. It is also advantageous tostart with as high a fuel density as possible, which is accomplished by forming a thin frozen layer on the inside of theshell. The integrity of this layer is maintained only if it
can be kept relatively cool and free of excessive preheat.
Numerical simulations and analytic model calculations show thata 1 to 3 eV temperature is sufficient to achieve compressedfuel densities several hundred times that of liquid DT with
only a few hundred joules of laser light.
Uniform spherical illumination was achieved with a new setof ellipsoidal mirrors designed to be more resistant to damage
than the previous TBIS system. This new system uses a doublebounce (OBIS) configuration from the mirrors in conjunction
with aspheric focussing lenses. Figure 1 shows the trajectoryof the marginal ray through the system. Careful positioning of
the two mirrors and lenses relative to the target position
allow the marginal rays to slightly overlap thus providing
uniform illumination over the whole of the target's surface.
DBIS increases the first bounce area on the mirror, relative tothat of the TBIS, so that the energy density is less. In
addition, AR coatings were optimized for 527 nm rather than the
dual wavelength coatings chosen for TBIS. Care was taken tominimize the debris from the target and support stalkreaching the mirror surfaces.

147

IAEA-CN-47/B-m-5

100
200
RADIUS Gmi)

300

FIG. 2. A measure of the illumination uniformity is provided by the symmetry of the ablated plasma.
Multi-frame holographic interferometry produces fringe patterns whose circularity and concentricity
show adequate uniformity. Numerical (Abel) inversion of the fringe pattern at different azimuthal
positions (12:00, 1:30, and 3:00) show similar density profiles.

The large solid angle occupied by the DBIS mirrors
severely restricts the diagnostic instrument access to thetarget. All instruments and target handling equipment must
pass through the annular gap between the two mirrors. The
cryogenic target positioner, cryogenic shroud, heating laser,and characterization interferometer take up a significant fraction of the gap. The X-ray backlighter diagnostic consists of148

LARSENo8
Q_

TIME -

FIG. 3. A bright, laser generated source of aluminum X-ray lines is produced behind the imploding
target. Temporally resolved transmission of the X-rays is recorded in one dimension by a pinhole/ streak
camera combination. Transmission through the core ceases at an interface radius of about 20 \im. The
implosion time is about 720 ps for a shaped laser pulse.

t = 720 ps

t = 540 ps

-fog
-50

0
+50
POSITION 0*m)

-50

0
+50
POSITION 0<m)

FIG. 4. Densitometer scans across the picture at several times provide the X-ray transmission as afunction of radial position. Transmission through the core ceases at a radius of about 20 /im. The
trajectory of the outer surface of the compressed core can be followed into stagnation, which occurs
at a radius of about 11 fim. Comparison of detailed numerical simulations (dashed lines) show excellent
agreement with the experimental profiles (solid lines).

IAEA-CN-47/B-m-S

149

the backlighter target and positioner, alignment telescope, and
X-ray streak camera. Other instruments competing for position
include pinhole cameras, holographic interferometer, and plasma
calorimeters.
A measure of the uniformity of illumination is provided by
the two-dimensional reconstructed image from the holographic
interferometer. Each frame, made with a 20 ps exposure, isdigitized and mathematically (Abel) inverted to provide aradial plasma density profile. Figure 2 shows one such frame
and three radial plasma profiles.
One of the key ingredients for a low preheat implosion isthe control of the shock wave generated by the laser pulseinteracting with the target. A laser pulse shaped in time,
roughly as t , accomplishes this. The other major source ofpreheat is from hot electron production which is minimized by
the use of 527 nm light. Approximately 160 to 180 joules oflight was incident on the target.
Low atomic number shells are preferred for these low preheat implosions, primarily for the compatibility with the Y-ray
backlighter diagnostic. Target capsules had a nominal 120 pm
o.d. and a wall thickness ranging from 3 to 8 ym. The cryogenic fuel was formed by the fast re-freeze technique resulting
in a solid fuel layer 1 to 2 pm thick. The uniformity andquality of this layer was verified interferometrically prior toeach shot.
The primary implosion diagnostic is an Y-ray radiograph of
the imploding shell as recorded by a streak camera. Because
the unablated shell is of low atomic number material andrelatively cold, there is little y-ray self-emission. An
aluminum V-ray source placed adjacent to the target emits line
radiation when illuminated by a third laser beam. X-rays fromthis target are absorbed by the inward moving portion of theshell, providing an image such as is shown in figure 3.
Interpretation of the history of the implosion depends
upon a careful analysis of these images. Densitometer scans,as a function of position, for several times during the courseof the implosion follow the position of the fuel-pusher interface until the Y-ray absorptivity causes the core region to
become opaque. The position of the outer portion of the compressed material can be followed to the culmination of the
implosion. Comparison of these experimental profiles to thosegenerated by the numerical simulation code show excellent
agreement, as is seen in figure 4. From this data (as well asother experimental quantities) the compressed fuel density canbe inferred. A conservative value of 35 to 40 g/cnr has beenachieved.

150

LARSEN
Breakup s X-ray emission Irom compressed core
100r

10

15
20
ASPECT RATIO

25

FIG. 5. Time integrated X-ray pinhole images occasionally show emission from the core location. A
statistical analysis of these occurrences shows a higher incidence for targets with larger shell aspectratios. These data may be the first direct evidence of a hydrodynamic instability in imploding
ICF targets.

Not all experiments ended with a uniformly compressed
core. Some shots showed significant X-ray self-emission fromthe core region in the time-integrated pinhole picture. A
statistical analysis of the occurrence of this feature as afunction of the shell's initial aspect ratio is shown infigure 5. No information was obtained as to the mechanism
responsible for this "shell breakup"; however, it is tempting
to speculate the Rayleigh-Taylor instability may be involved.

IAEA-CN-47/B-IH-S

151

During 1987 we will extend these measurements on low preheat implosions using cryogenic fuel and shaped laser pulses.
A brighter, higher energy (titanium target) backlighter source
will be used in conjunction with higher resolution imaging
devices, and possibly including a two-dimensional X-ray framing
camera.

IAEA-CN-47/B-HI-6

PROGRESS IN
INERTIAL CONFINEMENT FUSION PHYSICSAND COMPUTATIONAL CAPABILITIES AT DENIM
G. VELARDE, J.M. ARAGONES, J.A. GAGO, L. GAMEZ,
M.C. GONZALEZ, J.J. HONRUBIA, J.M. MARTINEZ-VAL,
E. MINGUEZ, J.L. OCANA, R. OTERO, J.M. PERLADO,
J.M. SANTOLAYA, J. SANZ, J.F. SERRANO, P.M. VELARDE
Instituto de Fusi6n Nuclear (DENIM),
Escuela T6cnica Superior de Ingenieros Industrials,
Universidad Politecnica de Madrid,
Madrid, Spain

Abstract
PROGRESS IN INERTIAL CONFINEMENT FUSION PHYSICS AND COMPUTIONALCAPABILITIES AT DENIM.
The contribution to inertial confinement fusion (ICF) at the Institute of Nuclear Fusion(DEN IM) of the Polytechnical University of Madrid has centred on the target physics area toimprove the description of the different processes involved. Moreover, target-blanket coupling
and new methodologies for materials activation and damage have been developed. The energy
deposition by laser or ion beams in the plasma, radiation transport, atomic physics, charged
particles, suprathermal and thermal electron transport were the main topics. Laser-plasmainteraction is studied with two-dimensional ray tracing that takes into account absorption andscattering, and the source of suprathermal electrons from resonance absorption and Raman
scattering is computed. Particle-in-cell simulations have been added to the present capabilities
at DENIM and some results on stimulated Raman scattering and microinstabilities in ion beamplasma interaction have been obtained. The tabular DENIM Atomic Library has been generated using newly developed atomic physics computational capabilities on the basis of average
atom models with different quantum treatments and a steady-state non-local thermodynamic
equilibrium (NLTE) model. The effect of the transport of knock-on deuterons and tritonsby neutron scattering has been incorporated recently. Furthermore, a detailed calculation of
the isotopic evolution by neutron reactions, for hydrogen and lithium isotopes, has beencombined with knock-on transport procedures for calculating the suprathermal fusion along
the ion relaxation path. Some results are presented on target analysis, considering laser orion beams with different pulse tailoring. Numerical simulations have been performed to study
the influence of aspect ratios and masses on the efficiency of the implosion and burnup of
ICF targets.

1.

SIMULATION MODELS

Inertial confinement fusion (ICF) physics and computational developments
at DENIM are either closely related to or have been incorporated in the NORCLA
code for the one-dimensional simulation of plane and spherical ICF targets.

153

154

VELARDE et al.

ION BEAMSRAY TRACING
INV. BREMSS.
RES. ABSORP.
RAD. PRESS.

LINEARSTOPPING
POWER

SUPRATHERMALELECTRONSENERGY DEPOSITION
TO ELECTRONS ANDIONSMOMENTUM TRANSFNORMAHYDRODYNAMIC EQUATIONS
EOS.Z
ANALYTICAL
EOS
MODIF. SAHAIONIZATIONMASS CONSERVATIONMOMENTUM CONSERVATIONENERGY:IONSELECTRONSRADIATION

r.v.p.Te.Ti.Z

MULTIGROUPRADIATION
DIFFUSION

DENIM ATOMIC TABLESEOSIONIZATIONOPACITIESCONDUCTIVITIESASTROPHYS.
LIBRARY(LAND

SESAMETABLES(LANL)

THOMAS-FERMIMODELSADELADIRAC + HFS

FIG. 1. Simulation models for hydrodynamics in ICF plasmas at DENIM.

Figures 1 and 2 show the improved simulation models included in the NORMA
code [1,2].
The NORMA segment can compute the linear stopping of external ion beams
using analytical stopping powers along radial tracks, with space dependent densities
and temperatures at selected time steps.
The laser-plasma interaction is considered at two simulation levels: ray
tracing coupled to hydrodynamics and particle-in-cell (PIC) simulation. The
ray tracing algorithm has proved to be flexible, cheap in computer time and155

IAEA-CN-47/B-IH-6

CLARA
PAR. + RES.
ABSORPTION

LASER
BEAMS

/p,Te,Ti,Th,Sh

SUPRATHERMALELECTRONS\THERMAL ELECTRONSSELF-CONSIST. E
COULOMB
LIGHT ORHEAVY IONS

ION
BEAMS

FOKKERPLANCKTRANSPORTCOLLECTIVE

ENERGYDEPOSITIONTO
ELECTRONS+ IONSAND
MOMENTUM
TRANSFER

COULOMB+SHORT RANGE
ALPHAS±

P.Ti

NORMA • — I

T+T

KNOCK-ON
IONS

ISOTOPICEVOLUTION

NEUTRONSBOLTZMANN
TRANSPORT\
ELASTIC+ N-2N
INELASTIC + N-yr,v,p,Te,TitZ\\

P,Te,Ti/

TRANSITIONS\^

f-f.b-f.b-b

ENDF/BLIBRARY

RADIATION
Y +X

NJOY

PAIRS + PHOTO. +COMPTON+THOMSON + ABSORPT.

MULTIGROUP
N + Y XS

FIG. 2. Simulation models for particle transport in ICF plasmas at DENIM.

easy to implement; it includes current corrections for non-Maxwellian electron
distributions, approximate factors for stimulated Brillouin scattering (SBS) andstimulated Raman scattering (SRS), and radiation pressure projection in theradial direction. The PIC WAVE code [3] was used for more detailed simulations of laser interaction. In particular, the growth and saturation of SRS was
simulated [1,4] and its importance as a mechanism in hot electron production
shown (Fig. 3). Also, the WAVE code was used for simulating a neutralized beam
impinging on a uniform plasma slab [1, 5].

156

VELARDE et al.

- 1

r
FIG
G. 3. Distribution function of the overall simulation box at t = 1015/W0, £gradie] •188c/W0,
T = 1.5 keV, I = 10i6 W/cm2. Two hot electron temperatures can be clearly seen.

The NORMA hydro solver has been extended to a fully implicit solutionof the energy equations for three temperatures (electrons, ions and radiation).
Radiation diffusion can be treated in a Grey approximation, with Rosselandmean opacities. The multigroup radiation diffusion set of subroutines has been
completed recently in a quasi-implicit coupling with the energy equations, andbenchmarks are being completed [1,2, 6].
Atomic data (equation of state (EOS), ionization, conductivities andopacities) are available from the DENIM Atomic Tables which have been largely
generated using a new family of atomic codes [ 1, 7-9]. The PANDORA codeuses the screened hydrogenic average atom model, while the ADELA codesolves the Dirac self-consistent Hartree-Fock-Slater (HFS) average atom.
The GEMINIS code computes the opacities.
A Thomas-Fermi code computes tabulated EOS and ionizations for elements
in any T-p grid. This model provides also the mean ionization and the averageionization potential for partially ionized atoms in plasmas with any degeneracy;these data are used to calculate ion stopping powers in the bound and freeelectrons of partially ionized plasmas. The model has been extended to calculations of the linear dielectric function in plasmas with any degree of ionizationand degeneracy.
The time dependent Fokker-Planck transport equation for charged particles
is solved by an original advanced finite-element method [9]. Numerical benchmarks have shown that the method is very efficient, i.e. its accuracy with ten
energy intervals is equal to or better than its accuracy with 40-50 intervals inthe diamond scheme in energy or with about 100 intervals in the multigroupmethod. This advantage is kept for all charged particle transport problems
where the first-order differential terms in energy (continuous slowing down andE-field forces) are dominant. This is the case for ICF applications, including

IAEA-CN-47/B-IH-6

157

external ion beams (light or heavy ions), alpha particles, knock-on ions andsuprathermal electrons [10].
The transport of external light or heavy ion beams, with any given spreadin velocity and angle of incidence (because of the finite beam radius) and time
pulse shape, is calculated with the Finite Element Method (FEM) solution in
space/energy/angle of the Fokker-Planck transport equation on a moving
Lagrangian mesh coupled to the hydrodynamics.
The transport of alphas born by fusion D-T reactions, with the emission
spectrum and source intensity in terms of the space/time dependent ion
temperatures and densities, is calculated with the FEM solution in space/energy(with S N in angle) of the transport equation, including the Fokker-Planck
collision terms (stopping powers and deflection) and the Boltzmann terms forlarge-angle encounters [10].
The transport of knock-on deuterons and tritons by neutron scattering has
been recently implemented and combined with the neutron transport calculation[1, 2], The kinematic laws for elastic and inelastic neutron scattering provide
the energy angle dependent source of the recoil nuclei, which are transportedby the same procedure as the alpha particles. The CLARA code is being extended
to include simultaneous transport of neutrons, alphas and knock-on ions, andcalculates not only the energy deposition and momentum transfer but also thetransport of mass (thermal ions) and the isotopic evolution by neutron reactions.
The detailed calculation of the isotopic evolution by neutron reactions, forhydrogen and lithium isotopes, has been completed recently and is being coupledwith the knock-on transport procedures for calculating the suprathermal fusions
along the ion relaxation path [11].
2.

ANALYSIS OF DIRECTLY DRIVEN TARGETS

Computer simulations have been carried out on heavy and light ion targets
and on directly driven laser targets. The main purpose of these parametricalstudies was to obtain a better understanding of the influence of pulse shape,
masses and aspect ratio on high gain performance. The published studies[1,2, 10, 12-15] related to this subject present more details about the characteristics of the energy deposition, implosion and burnup phases, and show the rangeof variation of driver energy for efficient power production. A summary of theresults for light and heavy ions is presented below. The results for laser driven
targets are considered to be of a preliminary nature.
Some other related work has been performed in order to characterize atarget with a well defined burnup parameter which permits a simple couplingof the neutronic response of the ICF blanket with the spectra of neutrons leaking
from the target [ 16]. New simulation models and related work have beendeveloped in order to take into account the activation of fusion materials [17],and a tentative scheme of materials damage has been proposed recently [18].

158

VELARDE et al.

2.1. Light ion beam targets

Our goal is to find optimized requirements for achieving a significant
energy gain. Two targets with different aspect ratios (low, high) and masses
have been considered [2, 15]. We have considered an external tamper of lead
(96; 96 mg) to minimize radiation leakage; an absorber of aluminium (99; 113 mg)
to stop the ions; an internal tamper of lead (6; 19 mg) to avoid radiation preheatingof the fuel; a thin shell of aluminium (0.5; 0.6 mg) to mitigate the Rayleigh-Taylor
instability; and a layer of cryogenic deuterium-tritium (2; 2 mg).
The ion beam pulse has a maximum power of 400 TW, FWHM = 14 ns,
centred at 15 ns, with a total energy of 5 MJ. The proton energy (with a 10%
energy spread), the radius of the beam and the intensity are:- Target 1: 6 MeV, 0.3 cm, 200 TW-cm"2- Target 2: 4 MeV, 0.4 cm, 90 TW- cm"2
The maximum power has been chosen in order to reach reasonable valuesof hydrodynamic efficiency. For instance, with 700 TW this efficiency isgreater (0.166) than in the case selected, with 400 TW (0.126). The beamfocusing increases the effective range in such a way that the thickness of theabsorber should be increased (5% for completely focused beams) in order tomaintain the same temperature conditions in the payload.
In case 1 the total gain obtained was 80, while in case 2 the total gain was
90. The results are presented in Fig. 4, together with data for heavy ion targets
and data from Lindl and Mark [19] for short-wavelength lasers. Our simulations
indicate a threshold near 2.2-2.5 MJ/mg to obtain a substantial energy gain forlarge aspect ratio targets such as those required for current pulse power diodefocusing. We consider this result somewhat conservative, mainly because of thehigh values of radiation leakage obtained in our simulations.
2.2. Heavy ion beam targets
The study of heavy ion beam targets has been carried out with strong effort,
with calculations for reference targets (HIBALL like), parametrization of fuelmass and target aspect ratio, and equivalent simulation of the heavy ion beams
tuned in voltage (energy) [1, 14].
Ion beams of 10 GeV Bi+ were assumed for the driver with power leadsof 720 TW, and a prepulse of 2.4 TW and 10 ns. The duration of the main pulse
was varied to obtain different total energies. The fuel masses were changed from
1 mg to 4 mg and a preheating lead shield of 15 mg was fixed. The masses of thetamper and ablator were modified to obtain different aspect ratios. The gain
versus beam energy for the cases of 1 mg and 2 mg is represented in Fig. 4.
Some very interesting results have been obtained in the simulation of theeffect of variable energy of the ions impinging on a target [2, 14]. In this case,

159

IAEA-CN-47/B-ni-6
f IDEAL
140 t

120

•

100 —

/ NC>RCLA/ v;VRIABLE ION ENERGYNORCLA
01'TIMIZED SINGLE SHELL, ,\ ^LIGHT ION BEAM
J><vSINGLE SHELL

•
*

/

Img

/

80

60

2mg

NORCLA^ S ^HIBALL TARGETS
^

7

40

— ^ ^ L I N D L S MARKjLLNL (1984)

lmg

20
•

/

3

4

BEAM ENERGY

FIG. 4. Energy gain versus beam energy for heavy and light ions and short-wavelength lasers.

a more realistic peak power (500 TW) was used. The effect was simulated byadjustment of the deposition profile to a given pattern, which is shown to be aneffective method of improving the dynamics of the target under compression,
permitting a drastic reduction of the mass of the internal shielding tamper.
The gain is shown in Fig. 4 for I mg of fuel mass. It can be concluded that theuse of heavy ion beams constantly tuned in voltage (energy) gives promising
results and leads to improvements.

REFERENCES[ I ] VELARDE, G., et al., Analysis of directly driven ICF targets, Laser Part. Beams 4(1986)349.
[2] VELARDE, G., et al., "ICF physics research at DENIM", Emerging Nuclear Energy Systems(Proc. 4th Int. Conf. Madrid, 1986), World Scientific Publishing, Singapore (1987).

160

VELARDE et al.

[3] FORSLUND, D.W., Rep. LA-UR-85-415, Los Alamos National Lab., NM (1985).
[4] VELARDE, G., et al., "Models for laser and ion beam plasma interaction in ablative
driven implosions", Laser Interaction and Related Plasma Phenomena (Proc. 7th Int.
Workshop Monterey, CA, 1985), Abstracts.
[5] VELARDE, G., et al., Bull. Am. Phys. Soc. 30 9 (1985) 1504.
[6] VELARDE, G., et al., "Calculational model of radiation transport in ICF", Atomic
Physics for Ion Fusion (Proc. 2nd Int. Workshop Oxford, 1984), Vol. 2 (1984) 569.
[7] VELARDE, G., et al., "Atomic processes calculations in hot dense plasmas using
average atom models", Radiative Properties of Hot Dense Matter (Proc. 3rd Int. Conf./
Workshop Williamsburg, VA, 1985), to be published.
[8] VELARDE, G., et al., Bull. Am. Phys. Soc. 30 6(1985) 1123.
[9] HONRUBIA, J.J., ARAGONES, J.M., Nucl. Sci. Eng. 93(1986) 386.
[10] VELARDE, G., et al., "Ion beam and fusion products transport in ICF targets",
Atomic Physics for Ion Fusion (Proc. 2nd Int. Workshop Oxford, 1984), Vol. 2 (1984)617.
[11] VELARDE, G., et al., "Lithium resonance self-shielding: Effect of tritium breedingin fusion blankets", Nuclear Data for Basic and Applied Science (Proc. Int. Conf.
Santa Fe, NM, 1985), Vol. 1 (1985) 195.
[12] VELARDE, G., et al., Trans. Am. Nucl. Soc. 49 (1985) 118.
[13] VELARDE, G., et al., Bull. Am. Phys. Soc. 30 9 (1985) 1850.
[14] VELARDE, G., et al., "Heavy ion beam transport and interaction with ICF targets",
Heavy Ion Fusion (Proc. Int. Symp. Washington, DC, 1986), to be published.
[15] VELARDE, G., et al., "Simulations of light ion beams and inertial confinement
fusion physics", Light Particle Beams (Proc. Int. Conf. Kobe, 1986), to be published.
[16] VELARDE, G., et al., Fusion Technol. 8, 1, 2B(1985) 1850.
[17] HERNAN, P., PERLADO, J.M., SANTOLAYA, J.M., SANZ, J., Trans. Am. Nucl. Soc.
52 (1986) 307. Also Fusion Technol. (Nov. 1986).
[18] PERLADO, J.M., HERNAN, P., SANTOLAYA, J.M., SANZ, J., "Comparative analysis
of ICF blankets; Data and methodology requirements", Emerging Nuclear Energy
Systems (Proc. 4th Int. Conf. Madrid, 1986), World Scientific Publishing, Singapore(1987).
[19] LINDL, J.D., MARK, J.W-K., Laser Part. Beams 3 1 (1985)37.

IAEA-CN-47/B-m-7

RECENT EXPERIMENTAL RESULTS
OF CRYOGENIC PULSED ION DIODES
K. KASUYA, K. HORIOKA, H. YONEDA, K. MITOBE,
S. KATO 1 , M. HARADA 1 , S. MORIMOTO 2 ,M. MORIOKA 2 , T. OBAMA 2 , T. TAIRA 2 , K. NAKAMURA 3
Department of Energy Sciences,
The Graduate School at Nagatsuta,
Tokyo Institute of Technology,
Nagatsuta, Midori-ku, Yokohama,Japan

Abstract
RECENT EXPERIMENTAL RESULTS OF CRYOGENIC PULSED ION DIODES.
A new machine for the extraction of pulsed ion beams from a cryogenic anode has comeinto operation. The anode is cooled by a refrigerator with helium gas circulation (15 W, 20 K).
The vacuum chamber is evacuated with a turbo-molecular pump to obtain a clean vacuum.
The source materials are puffed on the anode and frozen on its surface in the vacuum chamber.
Diode and beam characteristics have been examined with two kinds of ice on the anode. The
extracted ion currents have been measured by a large diameter biased ion collector. As an
example, a total ion current of 120 A and an ion current density of 5 A/cm2 on the anodesurface have been obtained at a diode voltage of about 300 kV with N2 ice. Without ice, themeasured ion signal is reduced to less than 1/10, i.e. a beam purity of more than 90% could beobtained. Time integrated Thomson parabola tracks have been analysed to ensure beam purity.

1.

INTRODUCTION

In 1981, a cryogenic pulsed ion diode was proposed by some of the present
authors [1]. This diode should be used to overcome difficulties experienced
with the conventional diode operated at room temperature. Since our cryogenic
diode could use frozen dielectric materials on the anode surfaces as ion sources,
pure ion beams [2] could be extracted from the uniform anode plasmas [3, 4].
Moreover, the source materials could be supplied easily through the gas puff
valves without breaking the vacuum. The present paper describes briefly themain results of recent experiments with cryogenic ion diodes at the Tokyo Instituteof Technology. Our final goal is the development of cryogenic pulsed ion sources
with high purity, high brightness and high repetition.
1

Nissin Denki Co., Ltd, Kyoto 615, Japan.
Ishikawajima-Harima Heavy Industries Co., Ltd, Yokohama 235, Japan.
3
Aisin Seiki Co., Ltd, Kariya 448, Japan.
2

161

162

KASUYA et al

CRYO-COOLERn

MARX GENERATOR

PFL

1 MV
3.75 kJ( Max. )TL

H, 5JJ.60ns

fjl(Mis"'rr^.i;'Afli i: 'iifii ;> ^i : r

0 . 2 Hz

15W-20K

DIODE

Tja CHAMBER

ITnr

J
r
HAG.
CONTR )L

f\TURB )-MOLECULAR

PUMP 500 L/s

MAIN
CONTROL

1
FIG. 1. Schematic diagram of the newly constructed experimental apparatus.

2.

EXPERIMENTAL APPARATUS

Figure 1 shows the newly constructed apparatus, which has a repetitive
capability of 0.2 Hz. A 5S2/60 ns coaxial water-filled pulse-forming line (PFL)
charged by a 1 MV/3.75 kJ Marx generator is used to apply a high voltage pulseon the diode (anode-cathode) gap. Figure 2 shows a schematic diagram of thediode section. A magnetically insulated extraction diode (radial magnetic field)[2-4] is used. The anode is cooled by a Stirling cycle refrigerator, with arefrigeration capacity of 15 W at 20 K, or by a liquid helium cooling system [4].
The vacuum chamber is evacuated with a turbo-molecular pump to avoid oil
contamination on the low temperature anode surface. In our experiments, thesource materials are puffed onto the anode and frozen on its surface in thevacuum chamber. The high voltage applied at the diode gap induces flash-overalong the dielectric surface. The flash-over plasma works as the ion source.

3.

EXPERIMENTAL RESULTS

The temperature history of the anode surface was measured by an
Au/0.07 Fe/Chromel thermocouple. The temperature of the refrigerator cryohead
was monitored by a couple of Pt-Co resistor thermometers. These results havebeen reported in Ref. [5]. The anode surface was cooled from room temperature
to about 20 K within 120 minutes.
Various ion species, such as N2 and Ar, were frozen on the cryogenic anodeand tested as ion sources. The extracted ion currents Ij were measured by abiased ion collector of large diameter, placed 20 cm from the anode surface.

IAEA-CN-47/B-III-7

1! I ! I163

1 f! I I l I * I I

FIG. 2. Schematic drawing of the cryogenic diode section.
CC is cryocooler, TF1 and TF2 are transfer tubes, TMP is turbo-molecular pumpand GIV is gas injection valve.

200
T(ns)

400

FIG. 3. Typical waveforms of pulse-line voltage Vp and total ion current Ii versus time.

Typical oscillotraces of the ion current extracted from the N2 puffed anode andfrom the bare anode (without gas puff) are shown in Fig. 3, with typical waveformsof the pulse-line voltage Vp . A total ion current of about 120 A O 5 A/cm2 on
the anode surface) was measured at a diode voltage of about 300 kV for the casewith a puffed anode. Without puffing (N2 gas), the measured ion signal was
reduced to less than 1/10. This means that more than 90% of the extracted beam
current can be attributed to the puffed material (N 2 ).

KASUYA et al.

164

0 •

100

200

300

E/Z (keV)

100

200

300

E (keV)

FIG. 4. Beam purity measured with a particle track detector.
(a) Number of tracks (N+ and N**J versus energy/charge.
(b) Reproducibility check ofN* particle counts with three successive beam shots.

(b)

(a)
N++(190keV)

N + (H0kaV)

T (200 ns/div.)Ht(230 keV) C+(190 keV)

T (200 ns/div.)

FIG. 5. Time of flight signals measured with biased ion beam collector
at three different locations: (a) for Ni ice, (b) for CHA ice.

IAEA-CN-47/B-ffl-7

165

2.0

r-—160K

t

9*

1.0
i

1
1 90K

n
0

SO
100
Diode time (ns)FIG. 6. Electron number density Ne of anode plasma versus diode time,measured spectroscopically.

To ensure beam purity, a CR-39 (plastic) particle track detector was used
together with a Thomson parabola analyser. The number of particle tracks on
the CR-39 detector was counted with an optical microscope by viewing separate
small parts of the parabola. The results are shown in Fig. 4, where the ordinate
gives the relative number of counts (N) per unit beam energy (AE) and theabscissa gives: (a) the energy (E) over the particle charge state (Z), and (b) the
energy (E) alone. The ratio N'VN** is 58%/42% and the beam purity is very
high with the frozen N2 anode, as shown in Fig. 4(a). The reproducibility ofthe purity measurements is shown in Fig. 4(b), giving only the N+ results forthree shots.
The time of flight of the ion beams extracted from the N 2 ice and theCH4 ice was also measured with the biased ion beam collector. Examples ofthe signals are shown in Fig. 5 for ion currents measured at three different
locations. The distance between the anode and the ion collector is indicatedby L (40, 60 and 80 cm). The monitor signals with the voltage pickup at thediode position are shown at the bottom of Fig. 5. The separation between H+
and C+ for the case of CH4 ice is clearer than the separation between N+ and N4*
for the case of N2 ice.
The effect of the anode temperature on the extracted beam characteristics
has been examined previously, and it has been found that the uniformity anddensity of the anode plasma increased with temperature [6, 7]. In our recent
experiments we measured spectroscopically the electron densities of the anodeplasmas produced from H2 O ice. The light emission from the anode plasma
was guided to a 50 cm monochromator by an optical fibre to avoid electromagnetic noise. The electron temperature was estimated from the hydrogen
Ha/H|3 line ratio, and the electron number density was determined from the
Stark broadening of the H a line. Figure 6 shows these results as a function of
the diode time, with the anode temperature as a parameter. The electron
number density Ne increases rapidly at higher anode temperature.

166

4.

KASUYA et al.

CONCLUSIONS

Pulsed ion beams of about 120 A were extracted from N2 puffed cryogenic
sources, using our newly constructed apparatus with the Stirling cycle coolingsystem and the turbo-molecular evacuation system. More than 90% of theextracted current could be attributed to puffed N2 gas. Measurements of thequantitative purity of the extracted ion species were performed with a time
integrated track detector.

ACKNOWLEDGEMENTS

The support by a Grant-in aid for Scientific Research from the Japanese
Ministry of Education, Science and Culture is acknowledged by the authors.
They also thank Prof. C. Yamanaka (Osaka University) and Prof. H. Takuma(University of Electro-Communications, Cho-fu, Tokyo) for their encouragement,
and H. Sumitani for his technical assistance.

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

[6]
[7]

KASUYA, K., HORIOKA, K., TAKAHASHI, T..URAI, A., FUJIKAWA, M., Appl.
Phys. Lett. 39(1981)887.
HORIOKA, K., TAKAHASHI, T., KASUYA, K., MIZUI, J., TAZIMA, T., Jpn. J.
Appl. Phys. 23 (1984) L374.
KASUYA, K., HORIOKA, K., TAKAHASHI, T., YONEDA, H., KUWABARA, K.,IEEE Trans. Plasma Sci., PS-13 (1985) 327.
TAKAHASHI, T., HORIOKA, K., YONEDA, H., KASUYA, K., Appl. Phys. Lett.
46(1985)249.
KASUYA, K., HORIOKA, K., YONEDA, H., OHBAYASHI, K., MITOBE, K., in
Ion Sources and Ion-Assisted Technology (Proc. 10th Symp. Tokyo, 1986), Instituteof Electrical Engineers of Japan, Tokyo (1986) 61.
YONEDA, H., HORIOKA, K., OHBAYASHI, K., KASUYA, K., Appl. Phys. Lett. 48(1986) 1196.
HORIOKA, K., YONEDA, H., OHBAYASHI, K., ISHIZAKI, K., KASUYA, K.,
Res. Rep. 769, Institute of Plasma Physics, Japan (1986) 221.

IAEA-CN-47/B-m-8

EXPERIMENTAL STUDY
OF UNIFORM ACCELERATIONWITH UV LASER LIGHT
K. KOYAMA, I. MATSUSHIMA, Y. OWADANO,
M. TANIMOTO, T. KASAI, A. YAOITA,
I. OKUDA, Y. MATSUMOTO, F. NEMOTO,
T. TOMIE, S. KOMEIJI, M. YANOElectrotechnical Laboratory,
Tsukuba Science City,
Ibaraki, Japan

Abstract
EXPERIMENTAL STUDY OF UNIFORM ACCELERATION WITH UV LASER LIGHT.
Non-uniformity smoothing and the properties of randomly phase-shifted irradiation with UV
lasers (0.27 /an) have been investigated experimentally. Non-uniformity smoothing of the accelerated
target measured by X-ray backlighting is in reasonable agreement with the improved 'cloudy day model'
in the power density region 5 x 1013 W/cm2 to 2 x 1014 W/cm2 and in the modulation wavelength
range 25-70 (jm created by multi-spot irradiation. Randomly phase-shifted irradiation, including
various modulation wavelengths, has been applied to study non-uniformity smoothing in a wider rangeof wavelengths by comparing the amplitudes of the Fourier components of UV and green laser intensity
distribution and of the X-ray radiation distribution. In the experiments with randomly phase-shifted UV
irradiation, local heating of the target has been found by observing the ion energy spectrum, the shocktransit time across the target, the burnthrough of the target and the X-ray backlighting shadow.

1.

INTRODUCTION

Experimental and theoretical evidence has been accumulated concerning theadvantages of a short-wavelength laser light in laser-plasma interaction processes,
such as efficient energy deposition, reduction of suprathermal electrons and highablation pressure [1, 2]. An important problem in laser fusion is the uniformity of
the ablation pressure, which has to be better than a few per cent in order to achieve
high gain [3]. In the case of direct-drive laser fusion, the asymmetry of the pressure
at the ablation surface is mainly due to non-uniformity of the laser illumination
when hydrodynamic instabilities do not exist. An investigation of non-uniformity
smoothing with a short-wavelength laser (0.35 pm wavelength) has been performed
previously [4].
In this paper, we present an experimental study of non-uniformity smoothing
and the properties of randomly phase-shifted irradiation [5] for a
0.27 nm wavelength laser using X-ray backlighting, Fourier analysis and other167

KOYAMA et al.

168

PARALLEL PLATE
FOCUS

FIG. 1. Schematic drawing of the focusing optics.

methods. We investigate the smoothing effect and the improvement of the acceleration uniformity for the quarter-micrometre wavelength region. Later, compression
experiments will be performed with the KrF excimer laser which is under development [6].

2.

EXPERIMENTAL APPARATUS

Experiments with a single-beam neodymium laser delivering 20 J in 350 ps(1.06 ^m pulse) have been performed. After frequency quadrupling through two
potassium dihydrogen phosphate crystals, the 0.27 /xm laser energy is typically 7 J.
The beam is focused onto foil targets with an F/6 lens of 500 mm focal length. The
incident irradiance ranges from 5 X 1012 W/cm2 to 2 X 1014 W/cm2. The targets
are planar 1 mm wide aluminium foils. The thickness of the targets ranges from
5 /urn to 15 jtni.
To study non-uniformity smoothing, irradiance modulations are introduced
without energy loss by inserting a pair of quartz plates in the incident beam behindthe focusing lens, as shown in Fig. 1. The wavelength of the intensity modulationis changed by varying the angle of incidence to the quartz plates.
For the randomly phase-shifted irradiation experiments, a random phase
plate [5] is used. This consists of randomly distributed half-wavelength phase shifter
elements. The size of the aperture element is 1 mm X 1 mm for a 0.27 /*m laser.
An X-ray pinhole camera is routinely used to monitor X-ray emission from theplasma. Four 10 /xm dia. pinholes with different filters are placed in front of a solidstate image sensor (charge-coupled device). For a 25 fxm detector element and acamera magnification of eight, the spatial resolution of the pinhole camera is about
13 /xm. The filters used are made of 8 /xm and 30 /im thick beryllium and 5 /xm and9 /xm thick aluminium.
The X-ray backlighting technique is used to measure the accelerated foil
velocity. The backlighting source is produced by focusing 1.06 pm laser light with

IAEA-CN-47/B-IH-*
IRRADIATIONMODULATIONMAIN
PLASMA

169
X-RAY
SHADOW

100 ,um
FIG. 2. Irradiation modulation of the focused pattern and X-ray shadow of the accelerated target.

a pulse duration of 350 ps onto a copper target. X-ray shadowgraphs are recordedwith a 5 fim pinhole camera through a 6 pm thick aluminium filter. The spatial resolution of the pinhole camera is 10 /jm. The effective photon energy range of the backlighting source is 1.0-1.5 keV.
The transit time of a shock wave through a target is measured by an opticalstreak camera with a slit focused onto the rear surface of the target across the centreof the laser driven target region. The temporal resolution is 100 ps and the spatial
resolution is 15 jum. The measured wavelength of the rear surface emission ranges
from 650 run to 850 run.

3.

NON-UNIFORMITY SMOOTHING

Non-uniformity smoothing has been studied experimentally by changing theirradiation power density and the modulation wavelength I. From video signals of
focal spot images in vacuum, a spatial wavelength of 25 ±5 ^m to 70±5 pm and a
modulation rate
Tl

Umax

^min/'V^max' ^min/of 0.73 ±0.05 to 0.29+0.05 are deduced ( I r a and I,™ are the peak and trough of
the laser intensity). Figure 2 shows a typical example of the focused laser intensity
distribution and an X-ray shadowgraph of the accelerated target. Assuming that therear side velocity modulations are equal to the ablation pressure modulations,non-uniformity smoothing is studied by measuring the displacement of the shadow
from the initial position of the rear surface of the target. The modulation rate of therear side velocity is defined asTv =(.^max ~ ATUiJ'(.Ainax 'where Z^ and Z^ are the peak and trough of the shadow of the displaced rear surface. The displacement after extinction of the backlight is 1/20. From our previous

170

KOYAMA et al.
k~0.1/um-1

1.0

0.5

I

10

10"

UWcm 1 )FIG. 3. Irradiation power dependence of the smoothing factor F at different modulation wavenumbers,
k = 2ir/ (, for 12.5 pm thick aluminium targets.

results for ablation pressure scaling, Pa oc I0-7 [1], the following pressure modulation rates without lateral smoothing, yp, are derived: YP = 0.6 for 7, = 0.7 and
Yp = 0.2 for 7, = 0.3. The smoothing factor, defined as T = 7V/7P, is comparedwith the improved 'cloudy day model', which is derived from a treatment of thetransverse perturbation propagating between the critical surface and the ablation surface [7].
Figure 3 shows the irradiation power dependence of F at different modulation
wavenumbers (k = 2TT/Q for 12.5 /xm thick aluminium targets. The solid and brokenlines in Fig. 3 indicate the smoothing factors analytically evaluated for
k = 0.3 /iin"' and k = 0.1 /xm"1.
The X-ray backlighting data have to be treated cautiously because of theblurring of the boundary caused by the thermal expansion of the target. In our experiment, the estimated blurring is small (10 ^m) so that the velocity modulation can bemeasured; the density distribution is assumed to obey the rules of one-dimensional
theory for shock waves and rarefaction waves.

4.

RANDOMLY PHASE-SHIFTED IRRADIATION

An experiment with randomly phase-shifted irradiation has been performed byinserting a random phase plate in front of the focusing lens. The measured intensity
distribution for a 0.27 /zm laser at the focal plane is shown in Fig. 4. The focusedlaser pattern shows spikes of various intensity for the different wavelength components, with a minimum pitch of about 3 jtm. The Fourier spectrum of the focusedlaser pattern is approximately |I(k)| = V k m -k, where k is the wavenumber and k^ is the cut-off wavenumber. The zeroth-order diffraction square(250 urn x 250 //m) contains 80% of the laser energy. The mean power density

IAEA-CN-47/B-EU-8

150

200

250

171

300

350

Pixel Address

FIG. 4. Intensity distribution fora 0.27 \im laser produced by inserting a random phase plate in frontof the focusing lens.

varies from 2xlO 1 2 W/cm2 to 1.6X1013 W/cm2. The targets used in the experiment are 5 /xm thick and 15 mm wide aluminium foils.
4.1. Fourier analysis of the smoothing effect
The dependence of the smoothing effect on the wavenumbers of the spatiallymodulated laser intensities can be investigated by comparing the amplitudes of theFourier spectrum of the incident laser beam with those of the observed physical quantities, because the focused laser pattern consists of a wide range of modulation
wavenumbers, as mentioned above. To investigate the effect of thermal diffusion
around the cut-off density, the amplitudes of the Fourier components of the focusedlaser intensity are compared with the amplitudes of the X-ray emission distribution.
The experiment is performed with lasers of 0.27 jum and 0.53 fim wavelength. The
sensitive spectral range of the X-ray detector is 1.0-1.5 keV. The detectable maximum wavenumber of the modulation (0.5 ixm~[) is limited by the spatial resolutionof the X-ray pinhole camera.
To demonstrate the effect of thermal diffusion over the critical density layer,
the amplitudes are compared at the temperatures TL (derived from the laser intensity
profile IL) and Tx (derived from the X-ray intensity profile Ix obtained with the
X-ray pinhole camera). In this derivation, we assume a relation TL = I2/3, on thebasis of a simple analytic model prediction, and Tx » Ix1/3, derived from the detector sensitivity for the calculated plasma radiation.
Figure 5 shows the amplitudes of the Fourier components of TL and Tx for a
0.53 fim laser; the intensities of the two curves are normalized at the fundamental
wavenumber. The smoothing factor of the thermal diffusion, TT, is defined as theratio of the Fourier amplitudes of TL and Tx. The smoothing factors for 0.27 jtm
and 0.53 pm irradiations at a mean intensity of 1.3xlO 13 W/cm2 are plotted in

172

KOYAMA et al.

from laserfrom X-ray

0.2

0.4

0.6

Wavenumber, k m (j*m"FIG. 5. Amplitudes of Fourier components of derived temperatures for a 0.53 fim laser.

Fig. 6. With green laser (0.53 ^m) irradiation, it can be seen that there is a tendency
for modulation smoothing as the wavenumber is increased. On the other hand, with
UV light (0.27 /an) irradiation, the reduction of the modulation is relatively small
and insensitive to the modulation wavenumber. Therefore, relatively poor thermalsmoothing can be expected with UV laser irradiation; the poor S/N ratio of the video
signal obscures the smoothing factor at the higher wavenumbers.
4.2. Shock transit time measurements andion energy distribution
The modulation of the shock transit time and the laser burnthrough patterns
have been compared. The streak photograph of the rear surface luminosity shows a
few bright precursors 200 ps after the peak of the laser pulse. The burnthroughpatterns monitored with a UV television camera correspond well to the precursors
appearing in the streak camera photograph taken simultaneously.
The observed shock arrival time is comparable with the calculated value of
190 ps for a power density of 10M W/cm2 (the peak intensity of the randomly
phase-shifted focus); this value was obtained with a one-dimensional hydrodynamic
model of laser-driven shock evolution [8]. After decay of the target density below
the cut-off density for a 0.27 /an laser, the tail of the laser pulse shows a burnthroughpattern. This means that modulation smoothing of the very intense spikes is rather
poor. The spikes on the rear surface of the accelerated target are observed also in
the X-ray backlighting photographs.
In the case of the irradiation of the randomly phase-shifted UV laser using therandom phase plate (power density 1013 W/cm2), a high energy shelf (about
50 keV) is observed in an ion energy spectrum obtained by a time-of-flight measurement (see Fig. 7). On the other hand, a monotonic ion energy distribution is observedin the irradiation without the random phase plate, in spite of the higher power density

IAEA-CN-47/B-HI-8

173? 0.1-

0.01
Wavenumber, k m i

FIG. 6. Smoothing factors of thermal diffusion, YT , derived by Fourier analysis.

(7xlO 1 3 W/cm2). The local intensity of the randomly phase-shifted irradiation issimilar to that of the coherent irradiation, but there is an obvious difference in theenergy spectrum between the two irradiation schemes. This fact suggests the presenceof local heating in hot spots of the focus.
In the experiment with the randomly phase-shifted irradiation, only a spatial
incoherence is introduced. Therefore, spatially fixed particular points are heated upby the intense part of the focused beam throughout the duration of the laser pulse.
If the position of the hot spots would change rapidly and randomly, which could be
the case if both spatial and temporal incoherence were introduced, more uniform
heating and/or acceleration might be possible.

5.

EXCIMER LASER AMPLIFIER

An e-beam pumped KrF laser has been developed as a middle-stage amplifierof a 1 kJ system for laser fusion research. The laser consists of one Marx generator(1 MV, 11 kJ), two pulse-forming lines (46 (2, 100 ns) with laser triggered output
switches, two e-beam diodes (10x60 cm2), and a laser cell (20x20x60 cm3 active
volume).
Two e-beams are injected into the cell through a titanium anode and pressure
foils. So far, a 235 J (85 ns) laser pulse has been generated in an injection locked
oscillator mode. The overall efficiency is 2.0%.
Both spatial and temporal incoherence may be required to achieve uniform
heating and acceleration of the target. With the broadband KrF excimer laser system

174

KOYAMA et al.
with Random Phase Plate.without RPP

10100
ION ENERGY(keV)
FIG. 7. Time-of-flighl measurement of the ion energy spectrum. Irradiation power densities:
1&3 W/cm2 with the random phase plate and 7 x 1013 W/cm2 without the random phase plate.
Laser wavelength: 0.27 /im.

under development in our laboratory, it may be possible to experimentally prove theeffectiveness of this scheme.

6.

CONCLUSIONSThe experimental results for non-uniformity smoothing of 0.27 /wn lasers are
in reasonable agreement with the improved 'cloudy day model' [7] in the modulation
wavenumber range 0.1-0.3 ^m" 1 and the power density region 5 x 1013 W/cm2 to
2xlO 1 4 W/cm2. The velocity modulation becomes smoother as the laser intensity is
increased and the modulation wavelength is decreased. The power density required
to achieve uniform compression, F < 1 %, is estimated to be more than
5xlO 1 4 W/cm2 for k = 0.1 fan"1 and 5 x l 0 1 3 W/cm2 for k = 2.0 nm~\Concerning the smoothing effect of the thermal diffusion around the cut-off
density, with a 0.53 /xm laser there is a tendency for modulation smoothing as thewavenumber is increased. With a 0.27 /zm laser, the reduction of the modulation isrelatively small and insensitive to the modulation wavenumber.
The randomly phase-shifted irradiation causes local heating and produces a highenergy shelf in the ion energy distribution.

REFERENCES[1][2]

OHWADANO, Y., et al., in Plasma Physics and Controlled Nuclear Fusion Research 1982
(Proc. 9th Int. Conf. Baltimore, 1982), Vol. 1, IAEA, Vienna (1983) 125.
MAX, C.E., McKEE, C.F., MEAD, W.C., Phys. Fluids 23 8 (1980) 1620.

IAEA-CN-47/B-in-8
[3]
[4]
[5]
[6]
[7]
[8]

175

BODNER, S.E., J. Fusion Energy 13 (1981) 221.
BOCHER, J.L., et al., Phys. Rev. Lett. 52 10 (1984) 823.
KATO, Y., et al., Phys. Rev. Lett. 53 11 (1984) 1057.
OWADANO, Y., et al., in Laser and Electro-Optics (Tech. Digest Conf. San Francisco, 1986),
Optical Society of America, Washington, D.C. (1986) 298.
MANHEIMER, W.M., COLOMBANT, D.G., Phys. Fluids 25 9 (1982) 1644.
COTTET, F., ROMAIN, J.P., Phys. Rev., A 25 (1982) 576.

IAEA-CN-47/B-IH-9

INERTIAL CONFINEMENT FUSION RESEARCH
WITH INTENSE PULSED LIGHT ION BEAMS
K. YATSUI, Y. SHIMOTORI, Y. ARAKI,
K. MASUGATA, S. KAWATA, M. MURAYAMALaboratory of Beam Technology,
Technological University of Nagaoka,
Nagaoka, Niigata, Japan
Abstract
INERTIAL CONFINEMENT FUSION RESEARCH WITH INTENSE PULSED LIGHT ION
BEAMS.
The paper reviews research and development of an intense pulsed particle beam at the Technological University of Nagaoka. The activities in connection with the 'ETIGO Project' include bothexperimental and theoretical studies. (1) A new, self-magnetically insulated ion diode called the 'plasma
focus diode' has been developed; a significant enhancement of the ion current density on the anode surface is achieved with this diode compared with other ion diodes, under the same experimental conditions. (2) A Rutherford scattering pinhole camera technique has been used to measure the local
divergence angle and to determine its dependence on various parameters. (3) A new diagnostic technique
has been developed for the measurement of ablation processes in ion beam-target interaction; a strong
light from a z-discharged plasma is utilized as a backlight. (4) An inductive post-acceleration experimenthas been carried out on an ion beam highly charge and current neutralized by electrons. (5) A new pulsed
power machine, the ETIGO-H, has been constructed; it has a Marx generator with 160 kJ of stored
energy and its output energy at 3 MV is 70 kJ. (6) The development of a particle-in-cell diode simulation
code, ion beam focusing with an external magnetic or electric field, ion beam transport and pellet implosion have been studied theoretically with one- and three-dimensional codes.

1.

INTRODUCTION

Light ion beams (hereafter referred to as LIBs) seem to be the most promisingenergy drivers for inertial confinement fusion (ICF). The focusing of LIBs is the mostimportant and serious problem that has to be solved. At the Laboratory of BeamTechnology of the Technological University of Nagaoka, basic and systematic studies
have been carried out [1, 2]. The 'ETIGO Project' includes studies aiming at a clear
understanding of the physics of ion beam focusing, the development of new ion
diodes and new diagnostic techniques, research and development of an inductive
post-acceleration technique for a light ion beam highly charge and current neutralized
by electrons, the construction of the new pulsed power machine ETIGO-II, and thedevelopment of particle-in-cell computer simulation codes for a comparison withexperimental results. In this paper, recent progress in these activities is briefly
reported.
177

178

YATSUI et al.

Ion Beam—-H— t—

0

10

20 30 40z (mm)

FIG. 1. The self-magnetically insulated 'plasma focus diode',

2.

EXPERIMENTAL RESULTS

2.1. Development of the self-magnetically insulated'plasma focus diode'A new and simple ion diode of the self-magnetically insulated type has beendeveloped [3]. Since the shape of this diode is similar to that of a Mather-type plasma
focus device, we call it the 'plasma focus diode' (PFD). Figure 1 illustrates schematically the basic principles of the PFD. The anode and cathode are formed by a pairof cylindrical electrodes; a flashboard is attached to the anode surface. If the diodecurrent exceeds a certain critical value, the electrons tend to be pinched around the
top of the diode (z = 40 mm), resulting in an electron beam directed to the top of
the anode or in the z-direction. The ion beam, which is extracted from the flashboard
anode, passes through a perforated cathode and tends to be concentrated on the central axis.
The features of the PFD can be summarized as follows. (1) The axially symmetric configuration ensures axial symmetry of every field and quantity. (2) The electrons do not pinch on the active area of the ion source, which permits multiple-shot
operation without a change of the flashboard. (3) Because of the initial radiation ofelectrons onto the anode before the beginning of magnetic insulation, a high ion current density can be obtained. (4) The targets can be irradiated with a large solid angle.
The basic properties and characteristics of the PFD as obtained experimentallyare summarized as follows. (1) An ion current density on the anode of 720 A/cm2
is obtained at an inductively calibrated diode voltage of about 660 kV, a diode currentof about 142 kA, a pulse width of about 90 ns, an ion current of about 32 kA, andan efficiency of about 22%. The ion current density is significantly higher than thedensity obtained with other diodes at the same parameters and conditions. (2) The
electrons are pinched near the top of the anode; correspondingly, the number density

179

IAEA-CN-47/B-IH-9(a)Vj - 650 kV: 7 mm01
TJ

Vd* - 450 kV(const.)Itf -450kV(»)-650kV(o)B/B-

10I, ( k A )20

FIG. 2. Dependence of local divergence angle (4>) on (a) transverse magnetic field strength (B/BJ
and (b) ion current (Ij); the inductively calibrated diode voltage (V*d) has been kept constant.

of protons increases in the direction of the top of the diode. (3) The local divergenceangle and the angle of deviation from ideal trajectories have been found to decrease
with decreasing diameter of the cathode holes. A possible reason for this is that theelectric field and the electron sheath are significantly disturbed when the diameter ofthe cathode holes becomes comparable to the anode-cathode gap.
The application of the PFD could be extended by spherically arranging both theanode and the cathode so that a two-dimensional focus is obtained. With such a PFD
it might be possible to concentrate a very high power density onto the target placedat the centre.
2.2. Measurement and evaluation of the local divergence angle
A better understanding of the local divergence angle in several parameter spaces
is very important for an enhancement of the brightness of light ion beams. Therefore,
we have measured and evaluated the local divergence angle (</>) of a light ion beam
extracted from a dual-current magnetically insulated diode (MID) using a Rutherford
scattering pinhole camera technique [4].
Figure 2(a) shows the angle <j> plotted against the insulating magnetic fieldstrength normalized by the critical field (B/Be). In this experiment, we observed thatthe diode current (IJ and the electron current (Ie) decrease monotonically withincreasing magnetic field. As seen from Fig. 2(a), the angle 4> decreases with increasing B/Bc when V^ is kept constant. Also, when the value of B/Bc is kept constant,the angle <j> increases with increasing V^. Figure 2(b) shows the angle <j> as a function of ion current I; and it is clear from this plot that the angle <>/ does not dependon the diode voltage and that it increases with increasing Ij.
An energy resolved measurement of <f> has been carried out with a speciallydesigned energy spectrometer. The ion detector is a CR-39 film covered with 1.5 pm
thick Mylar to eliminate carbons. Figure 3 presents the density distribution of protons

180

YATSUI et al.
100.

100

650 keV

501
35( V. )

100
50
0

50
0

30
x ( mn

25

ZK1

35

30
x ( mm )25

550 keV

35

0
100r
50
0
(•/.)
100
50

30
x ( mm )30( mm )

25

30: ( mm )

25

35

(V.)

600 keV

(•/.)

100

A50 keV

50

25

0'

380 keV35

350 keV3530x { mm )

25

100

50
0

500 keV35

30( mm )

25

FIG. 3. Density distribution of protons in the energy range of 350—650 keV, obtained by an energy
resolved measurement of the local divergence angle.

in the energy range of 350—650 keV. It can be seen clearly that the angle <f> of thehigh energy component produced in the initial stage of the voltage pulse is large compared with the angle 0 in the low energy range.
Using a combination of plastic scintillator and image converter camera, we havecarried out a time resolved measurement of 4>. Figure 4 shows a typical example
obtained with this technique. The part seen to be 'white' in the photographs indicatesthe location of the beam irradiation. In particular, as seen from photograph No. 3,
many 'islands' appear. This seems to be due to the fact that the anode plasma does
not grow uniformly throughout the anode surface and that the ion beam is extracted
from many 'source' islands that are strongly localized. Therefore, inhomogeneity of
the anode plasma is important in the initial stage of beam production. With the passage of time, two circular patterns appear; this indicates the 'smoothing' of the anodeplasma (see photographs Nos. 4-6). After smoothing, however, a relatively large
divergence exists, which seems to indicate the appearance of some plasma instabilities. These results qualitatively agree with the fact that the high energy componentof the LIB has a large divergence, as mentioned previously.
2.3. Ablation pressure measurement of beam-target interaction
by the backlighting technique
To study the ablation process of LIB-target interaction, a new high speed streak
photographing technique has been developed in which the strong light produced from

181

IAEA-CN-47/B-IH-9

10mm

1 2 3 4 5 6 7 8 9 10 11 12 13 1415
1t; :: i: ; : i; !: ':: : '400" :: 'j ; ^
s

160_
V>l

800 • y*-• y ^ ^ . } ' " ' 80^
0

0

0

200
400
t (ns)

FIG. 4. Framing photographs of a time resolved measurement of the local divergence angle, andwaveforms of the inductively calibrated diode voltage (VJ) and diode current (IJ.

a z-discharged plasma is used as backlight [5]. The target materials are aluminium(5 and 15 /am thick) and Mylar (2 /xm thick), and the targets have a size of
50 mm X 10 mm.
Figure 5 shows typical high speed streak photographs obtained with thistechnique. The rear velocity (vr) can be estimated as follows:
vr = 1900 m/s (15 urn Al)
vr = 4600 m/s (5 /urn Al)
vr = 930 m/s (2 /am Mylar)at Vd* = 1010 kV (R = 14.1
at Vd* = 1010 kV (R = 14.1 /*m)at Vd* = 590 kV (R = 7.6

where R denotes the range of protons. For 15 pm Al (Fig. 5(a)), which corresponds
to a 'thick' target, the velocity change is at t = 1 fis, which seems to be due to thefact that the pusher begins to be heated and expanded by the thermal conduction fromthe ablator ('burnthrough').
The ablation pressure is evaluated as Pa = 1.6 X 103 bar, at P^ (beam power
density injected) = 2.4 GW/cm2 for the thick aluminium target (15 /tm). This low
value of the ablation pressure seems to be due to the fact that a considerable amountof the beam power is absorbed by ionization or excitation of the target. Finally, thehydrodynamic efficiency ij (kinetic energy of the pusher per ion beam energy
injected) can be roughly estimated as JJ ~l%.

182

YATSUI et al.
(a) Al(b) At (5jjm)

LIB(c) Mylar (2jJtn)

LIB

FIG. 5. Streak photographs with ablation pressure measurement using the backlighting technique,(a) and (b) V'd ~ 1010 kV, ld ~ 97kA, B/Bc ~ 2.5, (c) V'd ~ 590 kV, Id ~ 38 kA, B/Bc ~ 2.3.
Targets were placed at z = 160 mm downstream from the anode.

100 mmFIG. 6. Cross-sectional view of the inductive post-accelerator MALIA-I.

2.4. Inductive post-acceleration of a highly neutralized light ion beam
An inductive accelerator system has been successfully operated for postacceleration of a light ion beam that is highly charge and current neutralized byelectrons [6].
Figure 6 shows the cross-sectional view of the inductive post-accelerator
MALIA-I (Medium-Mass Atom Linear Induction Accelerator). It consists of an ion
diode and a post-accelerator. The ion diode is an applied-Br magnetically insulated
diode (MID) which is directly connected to a Marx generator (200 kV, 5 kJ). The
anode is annular and is equipped with a flashboard of polyethylene. The anodecathode gap is 15 mm. The insulating magnetic flux density is about 0.15 T at theIAEA-CN-47/B-IH-9

183

x (mm)x (mm)
FIG. 7. Thomson parabola spectrometer data: (a) before post-acceleration, (b) after post-acceleration.

centre of the anode. The post-accelerator is driven by a Blumlein pulse forming line
(PFL) with switching reversed. The central conductor of the PFL is directly connected to the post-accelerator, which has neither a pre-pulse switch nor a charging
inductor. In the charging phase of the PFL, reversed current (pre-pulse) flows
through the chamber, inside which ten pieces of ferrite cores are installed; these cores
are ready to reset to full swing (AB * 0.63 T). The cross-section of the ferrite coresis 250 cm2, giving Vr = 1.57 x 10"2 V-s. The post-acceleration gap, separated by
20 mm, is immersed in a transverse magnetic field (B = 0.095 T) for the insulationof electrons. Using the applied-B MID, we produced an ion beam with a diode voltage of 90 kV, Jj = 20 A/cm2, Ij = 4 kA and T = 750 ns; this beam has beenhighly charge and current neutralized.
Figure 7 presents data of a Thomson parabola spectrometer before and after
post-acceleration. Before post-acceleration, H + , H2+, C + and C 2 + are present. The
peak energy of H + , H2+ and C 2 + is about 90 keV/z. Applying an induction voltage,
V M , of 210 kV, with a pulse width of 50 ns at t = 0.9 /is after the start of thediode voltage, we carried out the post-acceleration experiment, the data of which are
shown in Fig. 7(b). The peak energy here is about 240 keV for H + , which is about
150 keV higher than that before post-acceleration.

184

YATSUI et al.
Marx Generator

Energy Storage Capacitor(ESC)

Impedance Conversion Line
(1CL)

Diode

~]m

Series Gap

Pulse Forming Line (PFL)

FIG. 8. Cross-sectional view of the new pulsed power generator ETIGO-II.

2.5. Construction of the new pulsed power machine ETIGO-II;
ion diode experiments
The new pulsed power machine ETIGO-II has recently been put in operation[1]. The construction of the new machine, which was officially approved by theGovernment, began in 1984-1985. Figure 8 shows a cross-sectional view of
ETIGO-II. The Marx generator, with a stored energy of 160 kJ, produces a high voltage pulse of 3 MV for about 1 /us. A 12-channel water gap switch is used to decrease
the inductance. Using an impedance conversion line, for conversion from 1.7 fl to
6.5 fi, a high voltage pulsed power output was achieved with a voltage of ~ 3 MV,
current of —460 kA, pulse width of ~50 ns, power of ~ 1.4 TW, and energy of~70 kJ. Using this machine, we carry out focusing experiments with a barrel shaped
diode or with the PFD (Section 2.1), as well as beam-target interaction experiments,
voltage enhancement experiments with a plasma opening switch, and research anddevelopment of several diagnostic techniques.

3.

THEORETICAL STUDIES

Theoretical studies have also been carried out; in particular, a 2.5-dimensional
particle-in-cell (PIC) computer simulation code, PCSKfK, has been successfully
developed in co-operation with the Kernforschungszentrum Karlsruhe (KfK) [7].
This code can be applied to multispecies ions and non-uniform meshes. In the subroutine of the particle generation, new particles are created by a Gaussian law
adapted to a half-space mesh in addition to an electrode. The velocity of each particle
is determined by using a Maxwell distribution function and random numbers. For theparticle pusher, the relativistic equation of motion is solved by a Bunemann scheme.
Poisson's equation and the static Ampere law are solved in order to obtain the electricand magnetic fields. To check the reliability of the PCSKfK, Child-Langmuir electron currents have been computed; the numerical errors have been found to be lessthan 2%.

IAEA-CN-47/B-IH-9185

Furthermore, theoretical investigations are being carried out on LIB focusing
by external magnetic or electric fields, LIB transport and pellet implosion, using oneand three-dimensional codes.

REFERENCES
[1]
[2]

[3]

[4]
[5]
[6]
[7]

YATSUl, K., et al., in High Power Particle Beams (Proc. 6th Int. Conf. Kobe, 1986), Instituteof Laser Engineering, Osaka University (1987) paper O-L-2.
YATSUl, K., et al., in ICF Research by Light Ion Beams (Proc. 2nd Int. Top. Mtg. Nagaoka,1986), Laboratory of Beam Technology, Technological University of Nagaoka(1986) 1.
SHIMOTORI, Y., et al., in Particle Beam Fusion and Related Problems (Proc. Top. Mtg.
Nagoya, 1985), Res. Rep. IPPJ-769 (2), Institute of Plasma Physics, Nagoya University(1986) 267.
TAKAHASHI, K., et al., ibid., p. 282.
IKEDA, M., et al., ibid., p. 305.
TANABE, T., et al., Pys. Rev. Lett. 56 (1986) 831.
KAWATA, S., et al., Primarbericht 14.04.01P44A, Kernforschungszentrum Karlsruhe (1986);
J. Phys. Soc. Jpn. 53 (1983) 3416.

IAEA-CN-47/B-m-10

RADIATION TRANSPORTIN LASER IRRADIATED TARGETS
A. NG, Y. GAZIT, F.P. ADAMS,
D. PARFENIUK, P. CELLIERS, L. DA SILVA
Physics Department,
University of British Columbia,
Vancouver, British Columbia,
CanadaAbstract
RADIATION TRANSPORT IN LASER IRRADIATED TARGETS.
The paper presents results of studies on X-ray transport in laser irradiated targets where the laser
driven shock plays an important role. X-ray preheat was observed in Cu and Mo targets from rearsurface luminous emission measurements. The observed preheat in Mo targets can be attributed to direct
X-ray shinethrough. The preheat in Cu targets is probably caused by radiation transport coupled withionization behind the shock front. Direct measurements of X-ray transmission through an irradiated Altarget and through an equivalent, cold foil were also made. The results suggest a significant increasein X-ray transport, due to a shift in the K-shell photoabsorption edge induced by shock compressionof the target.

1.

INTRODUCTION

In the irradiation of medium- to high-Z targets by intense, short wavelength
laser light, high ablation pressure [1-4] as well as efficient X-ray conversion [5-8]are obtained. Using a numerical model which treats hydrodynamics, ionization andradiation in a completely self-consistent manner, Duston et al. [9] have studied indetail photon emission and energy transport in laser heated foils. In particular,
ionization by radiative heating resulted in shifting of absorption edges to higher
photon energies. This in turn increased the energy transport, allowing the radiation
to slowly burn its way through the target. On the other hand, in the presence of astrong shock, the process of radiation transport through the dense target can also be
significantly modified owing to pressure ionization [10]. This may further lead to acoupling of the energy transport to the shock hydrodynamics.
In this paper, we present results of our study on X-ray transport in laser
irradiated targets where the ablatively driven shock wave may play an important role.
Energy deposition on the rear surfaces of high-Z targets (Cu and Mo) by shock andradiative heating was investigated from temporally and spectrally resolved measurements of the rear surface luminous emission. To assess the effect of shock compression on radiation transport, X-ray transmission through an irradiated target(aluminium) was compared with that through an equivalent, cold foil.
187

188

NG et al.
1000

100:

z
UJ

1
TIME (ns)

TIME (ns)FIG. 1. Intensity of rear surface emission (4700-6300 A) from (a) 20 fim Cu target and (b) 15 \xm Mo
target. Solid line: measured intensity; dashed line: calculated intensity for shock heating
alone; dot/dashed line: calculated intensity for heating by X-ray shinethrough.

2.

X-RAY PREHEAT MEASUREMENTS IN COPPER AND MOLYBDENUM
TARGETSFor X-ray preheat studies in high-Z targets, Cu (Z = 29) foils, 20 pm thick,and Mo (Z = 42) foils, 15 /xm thick, were irradiated with a 0.53 ^m, ~ 2 ns
(FWHM) laser pulse from an Nd-glass laser. The laser beam was focused onto thetarget front side with f/10 optics at an incident angle of 10° off target normal. At
focus, the intensity distribution was nearly Gaussian, with 90% of the incident laser
energy contained in a spot of 80 /xm diameter (yielding an averaged irradiance $90)
and 60% energy in a 40 /*m diameter spot (with the corresponding $#,). A time
integrated image of the focal spot revealed hot spots of < 10 pm in diameter.
However, no significant hot spot was observed in the shock front [11] or the preheat
front [12] as it emerged from the target rear surface.
Heating of the target rear surface was observed in luminous emission spectrum
measurements. Details of the diagnostics have been described in Refs [13, 14].
Figure 1 shows the observed luminosity in the spectral band of 4700-6300 A
as a function of time for the Cu and Mo targets, at an absorbed irradiance of$60 = 1.3 X 10'3 W/cm2. The risetimes (corresponding to one decade increase in
the luminosity) were ~ 0.8 ns for the Cu target and ~ 1 ns for the Mo target — much
longer than the risetime (essentially detection system limited) of ~ 100 ps observedin lower-Z aluminium (Z = 13) targets at similar irradiances [15]. The very sharp
turn-on of the luminosity in the Al targets is characteristic of shock heating. For the
Cu and Mo targets, if the observed rear surface heating were caused by the shock,an extremely thick shock front (~10/jm) would be inferred from the measured
luminosity risetimes and shock speeds (~10 6 cm/s).

189

IAEA-CN-47/B-m-10
2.0

-1

1
TIME (ns)

-3

-2

-1

0

1

TIME (ns)FIG. 2. Rear surface temperatures for (a) 20 pm Cu target and (b) 15 p.m Mo target. Open circles:
measurements; dashed line: calculated temperature for shock heating alone; dot/dashed line:
calculated temperature for heating by X-ray shinethrough.

By comparing the measured emission spectrum with black-body curves, thetemperature of the target rear surface can be determined; the results are presented
in Fig. 2. The effective temporal resolution of 200 ps was governed by the time
integration window used to obtain the instantaneous emission spectrum. Also shownin the figure are the shock heated temperatures calculated from one-dimensionalhydrodynamic simulations. Evidently, heating of the target rear surfaces could not
be accounted for by shock heating alone and there appeared to be significant preheat
ahead of the shock front.
Possible energy transport mechanisms effecting preheat of the target rearsurface include fast electrons, photon transport from directly behind the shock front,
and X-ray transport from the coronal plasma on the target front side. Even at an
irradiance of $$(, = 8 X 1013 W/cm2, X-ray continuum measurements have shownthat < 0.1 % of the absorbed laser energy was deposited in a hot electron componentof about 2 keV [15], which would have a very limited range (~ 0.1 /xm) in the target.
The photons emitted from the shock heated material at ~ 1 eV would be
predominantly in the ultraviolet, with a very short range (<0.1 jtm) in the solid. Fora shock speed of 106 cm/s, they could only cause heating ahead of the shock front
with a risetime in the luminosity of < 10 ps.
On the other hand, a simple model for X-ray transport through the target isdirect shinethrough where heating of the target rear surface is caused by photons witha mean free path equal to the target thickness. The necessary conversion efficiencies
for such photons to produce the observed heating were ~ 1 % (~ 5 keV) and ~ 0.3 %(~ 15 keV) for the Cu and Mo targets, respectively. Results of the temporal profilesof the rear surface luminosity and temperature calculated from the 1-D hydrodynamic
simulations with and without such direct X-ray heating are shown in Figs 1 and 2.
These results suggest that direct X-ray shinethrough may account for the preheat in
the Mo targets but not in the Cu targets. For the Cu targets the radiative preheat can

190

NG et al.

FIG. 3. X-ray signal on (a) the front side and (b) the rear side of a (25 ±1) nm Al target. The front
signal was time delayed by 50 ps relative to the rear signal.

be satisfactorily modelled by a wave-type transport induced by a change in theopacity of the shocked target that is due to pressure ionization. Details of this model
will be published elsewhere.

3.

X-RAY TRANSMISSION THROUGH ALUMINIUM TARGETS

In this experiment, planar Al targets (17-40 /*m) were irradiated with a
0.53 fim laser pulse similar to that described above. X-rays transmitted through theirradiated target were monitored using a high-speed X-ray diode (with an aluminizedpolycarbonate filter to eliminate visible/UV radiation) placed on the target rear side.
Also, X-ray transmission through an equivalent, cold foil was measured with asimilar diode (with an aluminized polycarbonate filter) placed on the target front side;
it had an additional Al filter of the same type as the target used. Figure 3 shows
examples of the front-side and rear-side X-ray intensities. The front-side X-ray signal
closely resembled the incident laser pulse. In contrast, the rear-side X-ray diodeshowed a strongly increased X-ray transmission through the target.
When the X-ray measurements were performed, we also monitored the arrivalof the laser driven shock wave at the target rear surface, using an optical reflectivity
diagnostic [16]. For the measurements presented in Fig. 3(a) the shock break-out
occurred 0.4 ns before the peak of the front-side X-ray emission. The very large
increase in the rear-side X-ray signal following the shock break-out was attributed
to enhanced X-ray shinethrough resulting from target decompression by thetwo-dimensional rarefaction wave launched at the target rear surface at shock breakout [15]. More importantly, a significantly enhanced X-ray transmission through theirradiated target was evident at the time of shock break-out. Figure 4 shows the X-ray
intensities transmitted through the irradiated targets and the equivalent cold foil, at
the time of shock break-out. Enhanced X-ray transmission was found for targets up
to 30 fim thick. This was attributed to increased transmission of the Al He a andlAEA-CN-47/B-ffl-lO

191

10
20
ALUMINUM THICKNESS

FIG. 4. Transmitted X-ray intensity at the time of shock arrival at the target rear surface.
Solid circles: measured intensity through the irradiated target; open circles: measured intensity throughan equivalent cold target.
Dashed lines: calculated intensity, assuming that the fraction of total X-ray energy residing in the Hea
and intercombination lines, fx, is 5% (a) or 15% (b).
Solid lines: calculated intensity, assuming that fx = 5% and the mass absorption coefficient for the
Hea and intercombination lines, )i/p, is 320 cm2/g (c) or 420 cm2/g (d).
Dot/dashed lines: calculated intensity, assuming that fx = 15% and ii/p = 480 cm2/g (e) or
iilp = 540 cm2/g (f).

intercombination (IC) lines at ~1.6keV resulting from a blue shift in theK-absorption edge caused by pressure ionization in the shocked target. A simple
model, taking into account the change in the K-shell binding energy due to ionization,
electron degeneracy and plasma screening effects, indicated that the K-edge will beblue-shifted past the He a and IC lines for an ionization of 4, comparable to that
expected for the twofold compressed aluminium in this experiment.

4.

CONCLUSION

X-ray preheat has been observed in Cu and Mo targets. Radiation transport bydirect shinethrough can account only for the Mo target results. Significant increasein X-ray transmission has been found in shock compressed Al targets. This is
attributed to a shift in the K-shell photoabsorption edge caused by pressure ionizationof the shocked target.

REFERENCES[1] YAAKOBI, B., et al., Opt. Commun. 39 (1981) 175.
[2] TRAINOR, R.J., et al., Phys. Rev. Lett. 42 (1979) 1154.
[3] NG, A., et al., Appl. Phys. Lett. 45 (1984) 1046.

192
[4]
[5]
[6]
[7]
[8]
[9]
[10]
[11]
[12]
[13]
[14]
[15]
[16]

NG et al.
COTTET, F., et al., Phys. Rev. Lett. 52 (1984) 1884.
YAAKOBI, B., et al., Opt. Commun. 38 (1981) 196.
MEAD, W.C., et al., Phys. Fluids 26 (1983) 2316.
MATHEWS, D.L., et al., J. Appl. Phys. 54 (1983) 4260.
NISHIMURA, H., et al., Phys. Fluids 26 (1983) 1688.
DUSTON, D., et al., Phys. Rev., A 27 (1983) 1441.
ZIMMERMAN, G.B., MORE, R.M., J. Quant. Spectrosc. Radiat. Transfer 23 (1980) 517.
NG, A., et al., in Shock Waves in Condensed Matter, Plenum Press, New York (1986) 255.
PARFENIUK, D., PhD Thesis, University of British Columbia, Canada (1987).
NG, A., PARFENIUK, D., DA SILVA, L., Opt. Commun. 53 (1985) 389.
DA SILVA, L., et al., J. Appl. Phys. 58 (1985) 364.
NG, A., et al., Phys. Fluids 28 (1985) 2915.
PARFENIUK, D., et al., Opt. Commun. 56 (1986) 425.

Session G
INTOR
Chairman of Session
Session G-I

B.B. KADOMTSEVUSSRIAEA-CN-47/G-I-1

INTOR: INTRODUCTORY REMARKS
S. MORI1
INTOR Group2,
International Atomic Energy Agency,
ViennaAbstract
INTOR: INTRODUCTORY REMARKS.
The International Tokamak Reactor (INTOR) is defined and its major objectives in the present
phase (Phase Two A, Part ID.) are described.

The International Tokamak Reactor (INTOR) is viewed as the major experimentin the tokamak programme between the present generation large tokamaks and thegeneration of demonstration reactors (DEMOs). Role and concept of INTOR are
being evolved by the INTOR Workshop, in the context of providing the broad,general prerequisites for design and construction of the DEMOs.
The INTOR Workshop is a collaborative effort among Euratom, Japan, theUSA, and the USSR. It is conducted under the auspices of the International Atomic
Energy Agency (IAEA), in terms of reference defined by the International Fusion
Research Council (IFRC), an advisory body to the Director General of the IAEA.
The INTOR activity is being carried out in phases and is now in
Phase II A/Part 3 (1985-1987); the reports of the former phases are Refs [1-4]. A
brief summary of the administrative structure of the INTOR Workshop, its role in' Permanent address: Japan Atomic Energy Research Institute, 2-2-2 Uchisaiwa-cho,
Chiyoda-ku, Tokyo, Japan.
'

INTOR GROUP

EC

Japan

USA

USSR

G. Grieger
E. Cocorese
F. Engelmann
F. Farfaletti-Casali
M. Harrison
A. Knobloch
D. LegerE. Salpietro
P. Schiller
C. Vieider

S. Mori
N. Fujisawa
T. Honda
H. IidaM. Kasai
T. Kobayashi
T. Okazaki
T. Tsunematsu

W.M. Stacey
D. EhstC.A. Flanagan
C. Henning
B.G. Logan
M.K. Peng
D.E. Post
T.E. Shannon
D.L. Smith
P.T. Spampinato
J.M. Tarrh

B.B. Kadomtsev
V.I. Igitkhanov
B. Kolbasov
A. Kostenko
A. Kukushkin
R. Litunovskij
I.V. Mazul
V. Pistunovich
G. Shatalov
V.L. Vdovin
S. Yakunin

195

196

MORI

TABLE I. INTOR WORKSHOP OVERVIEW
Workshop session
Home institute activity"(weeks/sessions/participants)(man-years)
ZERO PHASE

November 1978December 1979

10/5/(4 x 4)

(15-20) x 4

PHASE ONEJanuary 1980July 1981

13/7/(8 x 4)

(20-40) x 4

PHASE TWO A/Part I

August 1981August 1983

12/7/(8 x 4)

(30-50) x 4

PHASE TWO A/Part II

January 1984September 1985

10/5/(8 x 4)

(30-50) x 4

PHASE TWO A/Part ffl

March 19861987

Rough estimate.

world fusion research, and its activity in each phase was given at the previous IAEA
Conference [5]. An overview of the Workshop is shown in Table I.
The major objectives of the present phase are to further improve the INTOR
concept by incorporating innovations to be developed on the INTOR timescale, toinvestigate critical issues in the concept, and to define R & D requirements in supportof the INTOR concept. An INTOR related Specialists' Meeting of the IAEA
(January 1986) has identified such innovations as could substantially improve thetokamak as a commercial reactor and/or as INTOR [6].
The following four presentations (papers IAEA-CN-47/G-I-2 to G-I-5) will
delineate the Workshop activity and specify research areas needed to define andimprove the INTOR concept. They are: Overview of the INTOR Workshop(W.M. Stacey, Jr.), Impurity and Particle Control (D.E. Post), Operational Limits
and Confinement (F. Engelmann), and Heating and Current Drive (N. Fujisawa).
These papers are based on the work of scientists and engineers in the EC, Japan,
the USA, and the USSR, who have contributed to the INTOR Workshop.

REFERENCES[1][2]

INTOR GROUP, International Tokamak Reactor: Zero Phase (Rep. Int. Workshop Vienna,1979), IAEA, Vienna (1980).
INTOR GROUP, International Tokamak Reactor: Phase One (Rep. Int. Workshop Vienna, 1980
and 1981), IAEA, Vienna (1982).

IAEA-CN-47/G-I-1
[3]
[4]
[5]

[6]

197

INTOR GROUP, International Tokamak Reactor: Phase Two A, Part I (Rep. Int. Workshop
Vienna, 1981-1983), IAEA, Vienna (1983).
INTOR GROUP, International Tokamak Reactor: Phase Two A, Part II (Rep. Int. Workshop
Vienna, 1984 and 1985), IAEA, Vienna (1986).
MORI, S., INTOR GROUP, INTOR: Introductory Remarks, in Plasma Physics and Controlled
Nuclear Fusion Research, 1984 (Proc. 10th Int. Conf. London, 1984), Vol. 3, IAEA, Vienna(1985) 187.
INTERNATIONAL ATOMIC ENERGY AGENCY, Tokamak Concept Innovations,
IAEA-TECDOC-373, IAEA, Vienna (1986).

IAEA-CN-47/G-I-2

INTOR: OVERVIEW OF THE INTOR WORKSHOP
W.M. STACEY, Jr. 1
INTOR Group2,
International Atomic Energy Agency,
ViennaAbstract
INTOR: OVERVIEW OF THE INTOR WORKSHOP.
The current INTOR design concept and the ongoing work in the INTOR Workshop are reviewed.
The ongoing work includes evaluation of innovations that could substantially improve tokamak reactors,
analysis of critical technical issues that affect the feasibility of the INTOR design concept and revisionof the INTOR conceptual design.

1.

INTRODUCTIONThe US, USSR, EC and Japan have collaborated since 1979,
through the IAEA-sponsored International Tokamak Reactor (INTOR)
Workshop, on a number of activities focused upon the next major
device (beyond the present generation of large tokamaks) in theworld tokamak program — an engineering test reactor. The INTOR
Workshop has proceeded in phases, first confirming the technicalfeasibility of such a device, then defining it's features througha conceptual design, and subsequently working on critical technicalissues that affect that design concept. Results of the recently
completed phase (1983-85) are published in Ref. 1.
During the present phase (1985-87), the INTOR Workshop isanalyzing six critical technical issues and examining a number ofinnovations that could substantially improve the tokamak as acommercial reactor concept. Scoping studies are being carried out
to evaluate the impact of these innovations on the INTOR designconcept. A revised INTOR conceptual design will be produced
during 1987.
2.

OBJECTIVES

The objectives of INTOR follow from consideration of itsrole in the fusion program and from an assessment of the technicalbasis which could exist within the next several years for its
design. INTOR is considered to be the step in the world fusion' Permanent address: Georgia Institute of Technology, Atlanta, GA, USA.
The members of the INTOR Group are identified in Paper No. IAEA-CN-47/G-I-1.

2

199

200

STACEY

program between the present generation of large tokamaks and a
tokamak DEMO. The programmatic objectives of INTOR are that itshould:(a) be the maximum reasonable step beyond the present
generation of large tokamaks (TFTR, JET, JT-60, T-15)
in the world fusion program(b) demonstrate the plasma performance required for the
tokamak DEMOs(c) test the development and integration into a reactor
system of those technologies required for the DEMOs(d) serve as a test facility for the blanket, tritium
production, materials and plasma engineering technology
development(e)test fusion reactor component reliability(f) test the maintainability of a fusion reactor(g) test the factors affecting the reliability, safety andenvironmental acceptability of a fusion reactor.
The technical objectives of INTOR have been developed tosupport the achievement of the programmatic objectives, while being
consistent with the anticipated technical basis for the design andconstruction of such an experiment to initially operate in the late
1990s. These technical objectives are given in Table 1.

3.

DESIGN CONCEPTThe original INTOR design parameters have been modified somewhat on the basis of the new knowledge of plasma physics and fusion
technology that has been gained during the course of the later
phases of the Workshop. The choice of parameters for the INTOR
concept is based, as before, upon anticipated advances beyond thepresently available data base. These parameters are given in
Table 2. A cross-section view of the INTOR design concept isshown in Fig. 1.
To ignite with the parameters given in Table 2, the energy
confinement time in INTOR must be 1.4 s. Energy confinement scalinglaws based upon Ohmic discharges, such as the neo-Alcator and T-ll
scaling laws, predict energy confinement times of 1 to 8 s. Scalinglaws derived from neutral beam heated discharges with extensive edge

IAEA-CN-47/G-I-2

201

TABLE 1. INTOR TECHNICAL OBJECTIVESA.

Reactor-relevant mode of operation
1. Ignited D-T plasma
2. Controlled burn pulse of >100s
3. Reactor-level particle and heat fluxes
4.
5.
6.

(P >1 MW.nf 2 )Optimized plasma performance
Duty cycle ^70%
Availability 25%

B.

Reactor-relevant technologies
1. Superconducting toroidal and poloidal
coils
2. Plasma composition control (e.g.,
divertor)
3. Plasma power balance control
4. Plasma heating and fuelling
5. Blanket heat removal and tritium
production
6. Tritium fuel cycle
7. Remote maintenance
8. Vacuum
9. Fusion power cycle

C.

Engineering test facility
1. Testing of tritium breeding andextraction
2. Testing of advanced blanket concepts
3. Materials testing (limited)
4. Plasma engineering testing
5. Electricity production ~5-10 MW(e)-2
6. Fluence — 3 MW.a.m

recycling (L-mode) scale favorably with plasma current and size for
INTOR, but the inverse dependence upon the total heating power
would lead to a prediction of an energy confinement time of about
one-half second in INTOR. Recent results from neutral beam heateddischarges in D-III and from ICRH heated discharges in PLT indicatethat the inverse dependence upon total heating power saturates aftera few megawatts, in which case the L-mode scaling laws extrapolate
to an energy confinement time for INTOR of several seconds. A morefavorable (relative to the L-mode) H-mode scaling is observed insome neutral-beam-heated experiments with divertors. The ASDEX
H-mode scaling extrapolates to an energy confinement time of about
4 s for INTOR.

202

STACEYTABLE 2. INTOR PARAMETERS

Chamber major radius, R
Plasma radius, aPlasma elongation, K
Burn average beta, (3Poloidal beta, j3j
Average ion temperature,
Average ion density,

4.9 m
1.2 m

1.6
4.9%

1.5
T

10 keVi

20 -3
1.4X10mn.

Energy confinement time, 7"E

1.4 s

Plasma current, IpField on plasma axis, B

8 MA

Safety factor (separatrix), q x

1.8Peak thermonuclear power, P ,585 MW(th)Neutron wall load, P

1.3 MW.m" 2

Burn time, Stage I/Stage II
Duty cycle, Stage I/Stage II
Max. availability
ICRF heating power at start-up
ECRH startup assist power
LHR current drive power
First wall material

5.5 T

100/150~60/~70%

25%
50 MW
10 MW
20 MW
H O , SS 316

Breeding blanket material

H 2 0 , SS 316

Breeder temperature
Breeding ratio
Annual tritium consumption at
25% availability
Tritium inventory
TF coil bore
TF coil conductor

410-800°C>0.6
6.1 kg.a
3.1-4.5 kg
6.3 m X 9.3 m
Nb Sn

TF coil maximum field
PF coil conductor

11 T
NbTi and/or Nb Sn

Stationary power loads
Pulsed energy storage
Inboard (shield) thickness
Outboard (breeding blanket andshield) thickness

200 MW
14 GJ
0.8 m
1.5 m

IAEA-CN-47/G-I-2

203"

FIG. 1. INTOR Phase Two design.

Recent experimental and theoretical results on the maximum
achievable value of beta can be characterized by the formulacaB

where C is a constant in the range 3-4. Choosing C = 4 leads tog c = 4.9% with I = 8 MA, B = 5.5 T and a = 1.2 m. With theseparameters, qj. = 1.8 and q^ = 2.4.

Studies indicate that 0.8-1.3%

in beta must be allowed for alpha particles and impurities. Thus,
a total beta of 4.9% corresponds to a D-T beta in the range
3.6-4.1%. The plasma parameters in Table 2 are predicted upon a
D-T beta of 4.1%.
The INTOR poloidal field coil system provides 112 V.s.
Assuming neo-Alcator scaling during the current ramp-up and Z=

1.5, this inductive capability would provide a 380 s burn at 6.4 MA
and a 150 s burn at 8.0 MA. A lower hybrid current drive systemis specified in the design for current ramp-up and .transformer
recharge. If this system works as hoped, INTOR will have thecapability of achieving burn times in excess of 1000 s.

204

STACEYA single-null poloidal divertor, with the chamber at thebottom, has been chosen for impurity control. Analyses indicatethat it is possible to magnetically form the divertor channels andto control the separatrix motion to within several centimeterswith coils external to the toroidal field coils. A relatively
short channel length is adequate because of the high-density modeof divertor operation. The reference plasma condition ischaracterized by a low plasma temperature £s20 e V ) , a high ion
20 -3
density (> 2 X 10m ) and a high ion flux at the divertor
plate. For the reference case a tungsten plate bonded to a copper
heat sink has been chosen as the best solution.
The mechanical configuration design was driven from the outset
by the requirement to provide sufficient access in order tofacilitate maintenance and assembly/disassembly. A semi-permanent
inboard, upper and lower shield forms the primary vacuum boundary.
Twenty-four removable torus sectors fit within this semi-permanentshield. These torus sectors are partially (outboard and upper)
the tritium-producing blankets and partially (inboard and lower)
the heat-removal shields. The final closure of the vacuum
boundary on the outboard is at the outer boundary of theremovable sectors. Once the vacuum boundary is cut, each torus
sector can be withdrawn horizontally with straight-line ortranslational motions through a 'window' between adjacent toroidal
field coils. The divertor channel is broken up into twenty-fourmodules which are removable between the toroidal field coils. The
single-null divertor was chosen over the double-null divertor inorder to achieve this more simply maintainable mechanicalconfiguration.
Semi-permanent, superconducting toroidal and poloidal fieldcoils will be enclosed in a common, semi-permanent cryostat, thus
completely separating the cold and warm structures. Only the lower
outer PF coils may have a separate cryostat for ease of maintenance.
All poloidal field coils will be external to the toroidal fieldcoils and super-conducting. Forced-flow conductor designs havebeen developed for the toroidal and poloidal field coils. Each ofthese conductor concepts is under active development, and a final
decision can await results from the development programs.
The rather demanding structural requirements for the toroidal
field coils are met by a combination of design strategies. Coilwedging, intercoil support structure and a bucking cylinder will be
used to handle in-plane and centering forces. Gussets, intercoil
support structure, a ring girder, the bucking cylinder and shear
ties will be used to handle out-of-plane forces and the over-turning
moment. A built-up laminated structure will be used.
A water-cooled, stainless-steel first wall with a panel-type
construction is specified. This first wall is expected to last the205

IAEA-CN-47/G-I-2

TABLE 3. INNOVATIONS CONTRIBUTED
Topic
Impurity control
Beta and confinement enhancement
Heating andcurrent drive
Advanced magnetics
Plasma engineering
Configuration andmaintenance
Advanced FW/B/S
Advanced materials
Compact reactors

EC Japan USA USSR Total
3
3
5
15
4
0

6

4

3

13

2
2
1

6
3
1

2
1

2
3
6

12
9
10

3
12

4
1

2
5
5
1

1
3

3
8
1

4

4

12
26
13
10

30

32

24

34

120

2
4

full lifetime of the device, provided that the melt layer which ispredicted to form on the inboard section during a plasma disruption
is stable. A tritium-producing blanket will be installed from theoutset of operation in order to reduce the operational cost. A
solid breeder (Li20) blanket that covers the outboard and uppersurfaces of the plasma chamber can produce more than 60% of thetritium consumed in INTOR. Adequate shielding is available forcomponent protection and to allow access 24 hours after shut-down.
4.

INNOVATIONSAn INTOR-related IAEA Specialists Meeting on Tokamak Concept
Innovations was held in January 1986.2 The purpose of this meeting
was to identify innovations that would significantly improve theprospects of tokamak development leading to an attractive endproduct — a viable tokamak fusion reactor. Emphasis was placed
upon innovations that would lead to substantial improvements in atokamak reactor, even if they involved a radical departure frompresent thinking. For the most part, such innovations are not yet
supported by the existing data base. Innovations were contributed
to the Meeting in nine categories, as indicated in Table 3.
The innovations contributed to the Meeting were evaluated on
the basis of substance and feasibility, and those innovations whichhave a high priority for further consideration were identified. It
is significant that 46 innovations were identified having a highpriority for further consideration, and 25 of these were judged to
be feasible on the time-scale of a next-generation experimental
reactor (i.e. the INTOR time-scale). This evaluation indicates that

206

STACEYa substantial improvement in the tokamak reactor concept, relative
to our present perceptions, may be realized in the future. Moreover, there appears to be a possibility to realize a significant
improvement in the INTOR concept.
The most promising innovations identified during the Meetingfell into a relatively small number of areas. A number of highpriority innovations were identified which could lead to a simpler
impurity control system, relative to the poloidal divertor. Amongthese were neutral beam driven impurity flow reversal and edge
ergodicity to a achieve a cool edge region that would make a pumped
limiter feasible, and helium removal via burial in in-situ deposited
surfaces.
Several high priority innovations were identified which could
lead to enhancement of beta. Among these were accessing the second
stability regime with or without plasma indentation, suppressionof sawteeth oscillations and current profile control.
Fast wave and neutral beam current dfive were identified aspromising alternatives to lower hybrid current drive for theachievement of steady state operation. Ion Bernstein wave heating
was identified as a promising alternative to ICRF.
The use of sub-MeV, negative ion based neutral beams for
impurity control, heating, current drive and burn control was given
a high priority for further consideration. A reactor that used asingle NBI system for those four purposes could be simpler than one
with ICRF heating, LHR current drive, a poloidal divertor for
impurity control and some other system for burn control.
Advances in superconducting magnet technology were felt tooffer substantial promise for improving the tokamak reactor concept.
Higher current density, higher field conductors would lead to morecompact coils. Forced flow, super-cooled helium conductors andfibre reinforced non-metallic structures were identified aspromising areas for technology advances.
Many promising concepts for breeding blankets were identified.
Among these, a self-cooled liquid metal blanket and a hightemperature He-cooled blanket appear to offer substantial
improvement relative to previous blanket concepts. The developmentof ferrltic/martensitic alloys in the near term and vanadium alloys
over the longer term would lead to extended blanket lifetimes. The
protection of components exposed directly to the plasma by means ofliquid metal droplets promises to extend the lifetime of the mostlimiting components (i.e. divertor plate or limiter).
Several possibilities for achieving a more compact tokamak
reactor were examined. It was concluded that improving upon the

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207

present line of moderate aspect ratio, moderate elongation,
superconducting reactor concepts offered more promise thanalternatives based upon very small aspects ratio, extreme elongation,or normal coils.
A proposal to locate MHD channels in the blanket introduced thepossibility of taking advantage of the unique characteristics offusion to reduce the cost of the entire energy conversion system.
In summary, the results of this Meeting led to threesignificant conclusions. There is a large potential for substantial
improvement in the tokamak reactor concept, relative to previous
perceptions; improving upon the current mainline tokamak
configuration is more promising than pursuing extreme configurations,and there is a possibility of utilizing the unique features of thefusion energy source to improve the overall reactor plus balanceof-plant system.

5.

CURRENT INTOR WORKSHOP ACTIVITIESThose innovations that were identified at the Specialists'
Meeting as having high priority for further consideration arepresently being evaluated by the INTOR Workshop, with emphasis
upon those which could be tested on the INTOR time scale. Scoping
studies are in progress to evaluate the possible impact of these
innovations upon the INTOR design concept.
Six technical issues are being studied currently in the INTOR
Workshop.
5.1

Impurity Control

The single-null poloidal divertor is the reference impurity
control scheme for INTOR. Studies are in progress to improveand validate models used to analyze the INTOR divertor
performance. In addition, five innovative schemes are being
considered: neutral beam induced flow reversal, ergodic edgelayer, radiating edge, helium burial and liquid metal protected
plates.
5.2

Beta and Confinement Enhancement

New experimental and theoretical results are kept under
scrutiny.
Methods are being studied for enhancing beta by:
optimizing profiles and shapes; accessing the second stability
regime; wall stabilization; and control of MHD activity.

208

STACEY
5.3

Heating and Current Drive

Work is ongoing to improve and validate the models used
for the reference ICRF heating, ECRH startup assist and LHR
current schemes in INTOR. In addition, fast wave and neutralbeam current drive and ion Bernstein wave heating are being
evaluated for possible use in INTOR.
5.4

Electromagnetics

Computer codes for transient electromagnetic calculations
are being improved and validated. The electromagnetic
requirements for plasma position, shape and disruption control
are being analyzed. High field-high current density TF
conductors, subcooled He-II superconducting coils and non- .
metallic structural materials are under evaluation as
innovations.
5.5

Configuration and Maintenance

A comparative study of the reference INTOR horizontal
maintenance scheme with a vertical maintenance scheme is inprogress. Other issues being evaluated are the application ofshape memory alloys, the use of ferromagnetic inserts for field
ripple reduction, PF coil redundancy, rapid replacement schemes
for the divertor and first wall, and the containment of tritium
and activated dust during maintenance.
5.6

First Wall and Blanket

First wall and breeding blanket design options for INTOR
are being evaluated. Several methods for protecting divertor
plates and the use of helium retaining surface materials arebeing studied. In addition, design constraints and guidelines
in these and related areas are being developed to guide theupcoming INTOR design concept revision.
In preparation for the design concept revision, a set ofphysics and engineering design constraints and guidelines arebeing developed.

ACKNOWLEDGEMENT
The work reported in this paper was carried out by a large number of scientists
and engineers in the USA, USSR, EC, and Japan.

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209

REFERENCES[1]

INTOR GROUP, International Tokamak Reactor, Phase Two A, Part II (Rep. Int. Workshop
Vienna, 1984 and 1985), IAEA, Vienna (1986); also Nucl. Fusion 25 (1985) 1791.
[2] INTERNATIONAL ATOMIC ENERGY AGENCY, Tokamak Concept Innovations,
IAEA-TECDOC-373, IAEA, Vienna (1986); also Nucl. Fusion 26 (1986) 375.

DISCUSSION
ON PAPERS IAEA-CN-47/G-I-1 AND G-I-2
R.W. CONN: Can the plasma current in the INTOR design be increased above
8 MA, and, if so, has the poloidal field system been designed to accommodate higher
current? It seems to me that the confinement capability of the present design isperhaps too small, given the ignition and burn objective of the machine.
G. GRIEGER: In principle, the plasma current can be made larger in INTOR,
but, in the present design, this would be at the expense of the voltseconds available
for currents driven inductively during the burn phase. However, it is expected that
progress in physics will narrow down the present range in predicted confinement
times to values compatible with the present INTOR parameters.

211

IAEA-CN-47/G-I-3

INTOR: IMPURITY AND PARTICLE CONTROL
D.E. POST1
INTOR Group2,
International Atomic Energy Agency,
ViennaAbstract
INTOR: IMPURITY AND PARTICLE CONTROL.
The impurity control studies for the past and present phases of INTOR have concentrated on thedevelopment of the design concept of an impurity control system for INTOR based upon a single null
poloidal divertor. The performance characteristics of the divertor have been analysed by the use ofdetailed computational models and consideration of the operational experience of poloidal divertors intokamak experiments. A viable engineering concept for the divertor system has been developed. Additionally, a pumped limiter system has been retained as a backup option. As a part of the present phaseof INTOR, a number of innovative impurity control schemes are also being studied. The major resultof all of the studies has been to increase the level of confidence in the divertor as the best available
system for impurity control for INTOR.

INTRODUCTION
The INTOR impurity control system must remove both the heat(124 MW) and particles (2 x 1 0 2 0 He atoms/sec) produced by thefusion reactions in INTOR [1]. The removal of the heat must be
accomplished without producing a large impurity content in theplasma and without damaging the first wall components.
The
removal of helium must be done with reasonably sized pumping
systems.
The two major systems considered for INTOR have beenpoloidal divertors and pumped limiters. A single null poloidal
divertor has been selected for the INTOR reference design
because of the possibility of operating the poloidal divertor ina "high recycling" regime in which a low-temperature, highdensity plasma is formed at the divertor plate by intense,
localized neutral recycling. The low-temperature plasma lowers
the sheath potential, which minimizes the sputtering of theplate, thus producing few impurities and leading to a longerosion lifetime for the plate. The pumped limiter was chosen
as a backup system due to the expectation that the plasma
temperature at the limiter would be high, leading to severe
erosion and impurity production problems.
A number of' Present address: Princeton Plasma Physics Laboratory, Princeton University,
Princeton, NJ, USA.
2
The members of the INTOR Group are identified in Paper No. IAEA-CN-47/G-I-1.

213

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POST

alternative systems have also been considered, such as bundle
divertors, but were judqed not as likely to be effective as apoloidal divertor or pumped limiter systemo
The major emphasis of the impurity control studies for
Phase IIA Part 2 of the INTOR was the development of detailed
design concept for the poloidal divertor system for INTOR. The
emphasis of the present phase has been continued refinement of
the divertor design includinq the new developments in divertor
experiments
andtheory andexaminationof a number ofinnovations for possible inclusion in the INTOR impurity control
system.

DIVERTOR PHYSICS
As stated above, the main attraction of a poloidal divertor
is the possibility of producing a low-temperature, high-density
plasma at the divertor plates due to intense, highly-localized
recycling.
The low plasma temperature produces a low sheathpotential.
Thus, the kinetic energy of ions striking thedivertor plate will be low, perhaps as low as 15-20 eV. There
are many materials which have sputtering thresholds above thisvalue, and which can be used for constructing the surface of thedivertor plate. Thus, the impurity production and plate erosion
rates can be very low, even with large energy and particle
fluxes.
The high plasma density leads to a high neutral atom
density in the divertor chamber, which minimizes the pumping
requirements for achieving the required helium exhaust.
The performance of the candidate divertor system has been
analyzed extensively with detailed models. These models include
the effects of the transport of particles and energy in the edge
plasma both along and across the flux surfaces in realistic twodimensional geometries based on flux surfaces computed using MHD
equilibrium codes.
The interaction of the plasma with therecycling neutral gas is computed self-consistently including
all of the major plasma-neutral collision processes, such as
charge exchange, atomic and molecular ionization, excitation,and recombination, and all of the major electron-molecule
collision effects.
These calculations indicate that the basic
INTOR divertor design should be able to operate in the "highrecycling" mode.
The performance characteristics of the INTOR divertor havealso been extrapolated from divertor experiments on ASDEX, PDX,PBX, AND D-III [2].
In each of these experiments, proper
localization of the neutral recycling has produced a "highrecycling" divertor, even with very large heat and particle
fluxes.
In the ASDEX AND PDX experiments, mechanical baffles
were used to localize the neutral recycling to the divertor
region. However, such baffles and larqe divertor chambers makea poloidal divertor very expensive in terms of space andpoloidal field system requirements.
In the D-III and PBXIAEA-CN-47/G-I-3

215

experiments, the plasma itself was used to baffle the recycling
neutralatoms
inageometricallyopen
configuration.
Measurements on D-III indicate that densities three or fourtimes the central plasma density could be produced near thetarget plate, and that temperatures in the 5-10 eV range could
be produced there even with substantial levels of neutral beam
heating.
Based on these results and the modeling studies, anopen, compact divertor design has been selected for INTOR (Fig.
1).
The experimental results also indicate that the impurity
levels in the experiments were reduced with proper operation of
the poloidal divertor in the "high-recycling" regime.
In addition, operation with a poloidal divertor appears tooffer the potential advantage of improved energy confinement in
the "H-mode" [3]. Operation in the H-mode produces improvementsin the confinement time over limiter operation (L-mode) offactors of up to two or more. Such operation would increase theiqnition margin
for
INTOR.
Although there are several
theoretical models for the confinement improvement due to the Hmode, none is generally accepted. The experimental recipes for
producing the H-mode include operation with a poloidal divertor
magnetic topology, and localization of recycling away from theplasma edge, especially on the large major radius side.
The
INTOR divertor is designed to localize the neutral recycling in
the area of the divertor chamber, both to produce a high"recycling" divertor and to maximize the potential for good "Hmode" operation.

DIVERTOR ENGINEERINGOne of the major fociof the Phase IIA Part 1 INTOR
studies was the consideration of a pumped limiter system for
INTOR [41. The major focus of the Phase IIA Part 2 workshop was
the development a preliminary conceptual design concept for the
INTOR poloidal divertor system.
One result of the detailed studies was that the divertor
chamber was shortened by roughly 40% in height from thepreliminary design of the Phase IIA Part 1 design [4] (Pig.
1).
This has the advantage of reducing the size of the vacuum
vessel and the poloidal field system requirements for thepoloidal divertor.
Both theoretical analyses of the divertor
design and experimental studies of operating divertors indicatethat the neutral recycling is localized very close to the plate,
and that a shorter divertor chamber would be adequate to provide
proper localization of the neutral recycling.
The design of the divertor plate is a heat sink covered bya protective armor. The armor is attached to the heat sink witha hiqh heat conductivity bond. The lifetime of the divertor isdetermined by the thermal fatigue of the composite divertor
plate and the degradation of its properties with radiation
damage.
The design studies included materials considerations,

216

POST

FIG. 1. Schematic illustration of INTOR divertor chamber.

fabrication
methods,
thermochemical
analyses,
disruption
analyses, lifetime predictions, and calculations for tritium
permeation through the divertor.
The reference divertor design consists of two actively
cooled plates angled so that the peak heat loads are reduced
from ~15 MW/m 2 normal to the separatrix to less than 5 M W / m .
The divertor plates consist of a structural heat sink to which
is bonded a high-Z plasma side material (Fig. 1 ) . The primary
candidate materials are copper alloys for the heat sink andtungsten for the plasma side material. The heat sink is cooledby water flowing in 0.6 cm diameter thin wall (0.1 cm) copper
tubes brazed onto the copper heat sink. The divertor surface isbroken up into 24 independent modules, placed side by side in
the toroidal direction. Each module consists of a large numberof two-material tiles with typical dimensions of 4 cm in theradial direction and about 3 cm in the toroidal direction which
are bonded to the cooling pipes.
The modules are part of adivertor cassette which can be removed for maintenance.
Detailed thermal stress analyses indicate that the maximum
allowable heat flux to the divertor plates is ~ 5 MW/m2. Helium

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217

and liquid lithium were examined as alternate coolants to water,
but were found to be inadequate for the divertor plate.
Stainless steel was chosen as the vacuum vessel material.

PUMPED LIMITER PHYSICS AND ENGINEERING
Studies were also carried out for a pumped limitersystem.
The major drawback of the pumped limiter is the
expectation that the plasma temperature at the limiter will behigh (~150 eV), resulting in a high sputtering rate. The pumped
limiter design is for a concave, double-edge bottom limiter.
The concept uses a removable limiter module with one module ineach of the 12 reactor sectors. The blade consists of a watercooled substrate covered by protective tiles.
The preferredmaterial for the limiter is carbon, but beryllium also appears
attractive.
Using a refractory metal such as tungsten on theleading edge also appeared attractive, since the temperature at
the leading edqe (~50 eV) is lower than at the point the limiterfirst contacts the plasma (~150 eV), and sputtering may not be asevere a problem.
While the analyses indicate that limiters may have large
impurity production rates, the actual performance of present
limiter systems using graphite tiles on TFTR and JET has beenquite good, with Z ff's 4 2 being produced routinely with highdensity operation and high power heating. If such good impurity
control results continue as the level of auxiliary heating
increases to ~30-40 MW, then pumped limiters will bereconsidered for the INTOR reference impurity control system.

INNOVATIONSFive possible innovative impurity control schemes arepresently being considered for the INTOR impurity control
system. They are:(1) active control of impurity transport,(2) removal of the plasma energy losses through edge
radiation,(3) use of an ergodic limiter,(4) burial techniques for He pumping, and(5) liquid metal coated divertor plates.
The proposal for active control of impurity transport isbased on the observation that, in many tokamak experiments (e.g.
[4]), the injection of directed momentum into the plasma canlead to enhanced outward transport of the impurities.
These
results have been analyzed with a theory [6] which, extrapolated
to INTOR conditions, leads to the expectation that injection
powers in the range of 50 MW should be adequate to keep thecentral INTOR plasma relatively clean. The potential impact on
INTOR of an impurity control system based on such a technique

218

POST

including the impact of a high energy, negative-ion-based
neutral beam system and driven operation is being assessed. The
major questions involve the credibility of such a method, andthe impact and feasibility of operating INTOR as a steady-state,neutral-beam-driven experiment.
A second proposal is to rely upon intense impurity
radiation from the plasma edge to reduce the heat flux via
charged particles to the limiter or divertor. There have been anumber of theoretical studies (e.g. [7]) of this issue, but they
have all either relied upon very high impurity concentrations or
impurity transport models which were felt to be unrealistic.
One conclusion of these studies was that medium-Z impurities,
such as iron, would be needed because the emissivities of thelower-Z impurities, such as carbon, are too low, and thetemperature at which they are completely ionized is too low. It
was also felt that such a radiating layer might be thermallyunstable. The recent observations of "detached" plasmas on DITE
and TFTR [8,9] in which all of the plasma energy is radiated to
the wall show
that, at least with low levels of heating,radiating edges do exist and are stable. The utility of suchphenomena in either reducing the peak heat loads or allowing achange in the INTOR impurity control system is being studied.
The major
unresolved question is the feasibility of such ascheme for the high power levels needed for INTOR.
Another
issue is whether such a scheme will be competitive with otherschemes, such as poloidal divertors, in terms of the plasma
volume required.
The third innovation proposed is the use of an ergodic
limiter [10-15] to reduce the peak heat loads and to cool theplasma edge.
The basic idea is use helical coils near theplasma edge to produce magnetic islands in the plasma edge. Dueto the resonant nature of the coils with the plasma current, thecurrents required in the coils would be modest. The mixing of anumber of different islands would produce an ergodic region at
the plasma edge where the transport perpendicular to the flux
surfaces would be very large. There have been three types of
ergodic layer proposals:(1) stationary ergodic limiter (e.g. [10-13]),(2) rotating ergodic limiters [14], and(3) helical divertors [15],The first two of these proposals rely upon enhanced cross-field
diffusion and energy conduction at the plasma edge to increase
the radial energy and density decay length.
In the extreme
limit, this would allow a uniform dispersal of the plasma energy
and particles on the wall.
It is also argued that such anergodic layer would also produce an edge radiative layer which
would aid in spreading out the heat uniformly and in cooling theedge [11]. Experiments on TEXT [16] have shown that, whileislands at the edge can be produced, they tend to form localized
hot spots, thus concentrating the heat flux.
Rotating theislands either by oscillating the plasma current or the helicalcoil currents would spread out the heat more evenly over the

IAEA-CN-47/G-I-3

219

wall [14]. The third proposal is for helical divertors [15]which would guide the islands into small divertor chambers
arranged on the vacuum vessel wall parallel to the resonant
islands. These divertor chambers would be small and quite close
to the plasma.
They could also be very "closed," and thus
enhance chances for the H-mode.
Each of these schemes would
also try to enhance the recycling at the plasma edge to bring
down the edge temperature.
The major issues associated with these proposals are thedegree of ergodization produced in the plasma edge, thefeasibility of making an ergodic edge with a reasonable coil set
and current requirements, the penetration of the helicalperturbations to the plasma center and the subsequent effect on
the central transport, and the successful demonstration both of
the feasibility of producing ergodic edge layers on tokamak
experiments and their utility in impurity control for thoseexperiments.
The fourth proposal is to bury the helium ash in some
portion of the surface of the vacuum vessel [17].
The
attraction of such a scheme is that removal of the helium witheither a divertor or a pumped limiter requires 100,000-2 00,000
A/sec of pumping, with large pumping ducts and penetrations of
the vacuum vessel, blanket, and shield.
The actual helium
production rate is small compared to the number of atoms in thetokamak wall (2 x i o 2 0 He atoms/sec vs 1 o' atoms), and could bepumped by burial or implantation quite easily. The basic idea
is to trap the helium ions or neutral atoms by ion implantation
as they strike the target plate as ions or are reflected as
neutrals from the target plate to strike a pumping surface near
the target plate.
The optimum material to be used for thepumping surface would be vanadium, nickel, or another materials
which have a low mobility for helium and a high mobility forhydrogen. The helium would stay buried, and the hydrogen would
diffuse to the surface, recombine, and leave the surface. To
prevent helium saturation of the pumping surface, new surface
layers would have to be laid down periodically.
The fifth proposal is to use liquid metal layers on thedivertor plates
to protect the plates and extend
theirlifetime. The liquid would be a low melting point, low-Z metal,
such as lithium, which would flow across the divertor plate, andprotect the divertor plate. While this proposal has a number of
attractive engineeringfeatures, the uncertainties in the
behavior of the liquid metal layer under the large heat andparticle fluxes expected during the operation of INTOR arelarge. The plasma-metal layer interaction is the major focus of
the present studies for this proposal.

FUTURE STUDIESINTOR

The confidence level for the poloidal divertor system forhas been enhanced by the recent studies, and by the

220

POST

success of poloidal divertor experiments. The "high-recycling"
regime is relatively well understood theoretically.
The
experimental performance of poloidal divertor9 has also been
excellent.
There are, however, theoretical studies thatindicate that the present highly inclined divertor plates for
INTOR may not be optimum for high recycling operation [18], An
additional consideration not yet included in the models is therole of impurity radiation in the divertor chamber, which ispresent in the experiments, and should reduce the peak heatloads. Additional studies are underway for this effect. Other
unresolved
issues still remain for divertors.
Improved
confinement has been observed with divertor operation, but thescaling of the improved confinement and its behavior with longpulse operation needs to be established.
The coupling of ICRF
waves to divertor plasmas also needs experimental study.
The
impact of disruptions on divertor plates warrants further
study.
There are also indications that central peaking of
impurities may occur during H-mode operation.
Another important area for theoretical and experimental
investigation is the transport of impurities along the fieldlines in a diverted plasma, including the effects of sputteringby helium ions and unthermalized hot ions from the main plasmaedge region.
The issues with limiters include the success (or lack ofsuccess) in providing impurity control in such large, high power
tokamaks as JET, TPTR, JT-60, and TORE-Supra.
Recent results
are quite promising. There also exists some possibility of theoperation of a "high-recycling" limiter [19],Resolution of many of these issues is part of the ongoingexperimental programs of many tokamaks.
Experiments withdivertors prototypical of INTOR are underway on D-III-D and will
be doneon ASDEX-upgradeand Alcator C-MOD.
Divertor
experiments are part of the programs on JT-60, JET, ASDEX, and
PBX, and divertors are part of the base designs of CIT, NET,
PER, TIBER, and OTR.
Theoretical work will take the directionof using models with better and more realistic geometries,
better boundary conditions, impurity transport, and better
coupling of the nonlinear interaction of the plasma and neutral
gas.
There will be continued emphasis on the comparison ofexperiments and theory.
Pumped limiter experiments will be carried out on TEXTOR(ALT-II), TORE-Supra, and D-III-D and analyzed to further theunderstanding of the operation of pumped limiters.
These theoretical and experimental studies will lead to thecontinued evolution of designs for the INTOR impurity control
system.
ACKNOWLEDGMENT
The INTOR impurity control group would like to acknowledge
the contributions of the many scientists in each country who

IAEA-CN-47/G-I-3

221

contributed to the studies for the INTOR Phase IIA Part 2 andPart 3 workshops.

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

INTOR, International Tokamak Reactor, Phase Two A, Part II (Rep. Int. Workshop Vienna,
1984 and 1985), IAEA, Vienna (1986).
LACKNER, K., KIELHACKER, M., J. Nucl. Mater. 128&129 (1984) 368.
WAGNER, F., et al., Phys. Rev. Lett. 49 (1983) 1408.
INTOR, International Tokamak Reactor, Phase Two A, Part I (Rep. Int. Workshop Vienna,
1981-1983), IAEA, Vienna (1983).
ISLER, R.C., et al., Nucl. Fusion 23 (1983) 1017.
STACEY, W., SIGMAR, D., Nucl. Fusion 25 (1985) 463.
GIBSON, A., WATKINS, M. in Controlled Fusion and Plasma Physics (Proc. 8th Europ. Conf.
Prague 1977), Vol. 1 (1977)31.
McCRACKEN, G., et al., "A study of marfes and detached plasmas during helium discharges
in DITE," to appear in J. Nucl. Mater. (1986).
STRACHAN, J., et al., Formation of Detached Plasmas, Rep. No. PPPL-2379, Princeton
Plasma Physics Laboratory, Princeton, NJ (1986), to appear in J. Nucl. Mater. (1986).
KAWAMURA, T., et al., J. Nucl. Mater. 111&112 (1982) 268.
FENEBERG, W., WOLF, G., Nucl. Fusion 21 (1981) 669.
OHYABU, N., Nucl. Fusion 21 (1981) 519.
IGITKHANOV, Yu., USSR Contribution to INTOR (1984).
TAZIMA, T., SUGIHARA, M., Rep. JAERI-M 8390. Japan Atomic Energy Research Institute,Tokyo (1979).
KARGER, F., LACKNER, K., Phys. Lett. 61A (1977) 385.
OHYABU, N., et al., Nucl. Fusion 25 (1985) 1684.
BROOKS, J., MATTAS, R., J. Nucl. Mater. 121 (1984) 392.
PETRAVIC, M., et al., to appear in J. Nucl. Mater.
SINGER, C , BRAAMS, B., Fusion Technol. 9 (1986) 320.

DISCUSSION
R. W. CONN: Two-dimensional fluid calculations of flows in diverted plasmas
show, in high recycling cases, that reversed flow out of the divertor can occur on
the flux lines near the separatrix. This would carry any impurities generated in thedivertor back into the main chamber. I believe a number of calculations show this,including your own. What are the implications of this for the INTOR design, and will
design changes have to be made?
D.E. POST: The general approach for INTOR had been to minimize the impurity production by use of a high recycling divertor. Experimentally, the impurities
produced at the plate appear not to get into the main plasma. Theoretical studies of
the impurity transport in the scrape-off layer are underway. Experimental results withdivertors often do show impurity problems which are attributed to improvements inthe particle confinement time. On balance, however, present assessments of divertors
for INTOR, including the results presented at this meeting, strongly support the
choice of a divertor for INTOR.

IAEA-CN-47/G-I-4

INTOR: OPERATIONAL LIMITSAND CONFINEMENT
F. ENGELMANN1
INTOR Group2,
International Atomic Energy Agency,
ViennaAbstract
INTOR: OPERATIONAL LIMITS AND CONFINEMENT.
A summary is given of the work being done at the INTOR Workshop in the areas of operational
limits and confinement of tokamak plasmas. This includes a continuous assessment of the database as
well as consideration of innovative concepts.

1. INTRODUCTION
In Phase Two A, Part III, of the INTOR Workshop a working
group was formed to assess the status of tokamak physics inthe areas of operational limits (i.e. limits to the plasmabeta, plasma density and safety factor) and confinement andalso to analyse possibilities for improvements in these areas.
The latter point includes consideration of innovative concepts
whose development might require more time than is available
for a next-step device like INTOR.
With respect to the status of tokamak physics, theworking group can draw on the results of the data baseassessment performed during Phase Two A, Part II, of the INTOR
Workshop [i]; see also [2]. New results obtained in tokamak
experiments and theory are evaluated as they become available.
Of course, the pace of progress here is set by that of thetokamak programme. As far as innovative concepts areconcerned, the working group has defined a number of tasks,
mainly in the field of beta limits in tokamaks, which are now
being investigated, to evaluate their potential.
In the following a report on the present- status of thework is given.

1

2

Permanent address: FOM Instituut voor Plasmafysica Rynhuizen, Nieuwegein,
Netherlands.
The members of the INTOR Group are identified in Paper No. IAEA-CN-47/G-I-1.

223

224

ENGELMANN

2 . BETA LIMITS

A considerable number of tokamak experiments in a wideparameter range have led to the conclusion that there is anoperational limit to the plasma beta, 3 m a x , depending on the
discharge parameters according to) - C I(MA)/a(m) B(T)(1)

where I is the plasma current, a is the plasma minor radius, B
is the confining magnetic field and C is in the range
C <= 2.8 to 3.5

(2)

[i].
This
finding
is generally
consistent with thepredictions of ideal MHD stability analyses for circular andmoderately D-shaped plasmas. In fact, fitting the theoreticalresults for the stability limit with respect to both ideal
ballooning and low-n kink modes also yields the scaling (1)with
C = 3 to H

(3)for optimized profiles [i]. There is no explicit proof,
however, that the operational limit encountered in experiments
is actually due to an ideal MHD mode becoming unstable.
While no dependence of C on the confinement regime or on
the heating method applied was found in experiments, recent
results on ASDEX [3] showing a time dependence of the beta
limit on the time scale of current diffusion confirm that C
depends on the current density profile j(r). The beta limit
sometimes is disruptive and sometimes manifests itself as adegradation of confinement. The reason for this difference is
not yet understood.
Emphasis is at present being put on improving thetheoretical analysis of the impact of the plasma shape and theplasma profiles on ideal MHD beta limits. Specifically, work
is under way to extend the theoretical studies of ideal MHD
beta limits to larger plasma elongation and triangularity,
including indented plasma cross-sections. In addition, thepotential, for enhancing beta, of flat current profiles with asafety factor on axis, q0, appreciably above 1 as well as ofoperation at q0 < 1 is being assessed. In this study attention
is being paid to the possibility of operating in the second
stability regime of ideal ballooning modes.
Preliminary results indicate that the scaling (1) with C
in the range (3) is valid for the beta limit with respect toboth ballooning modes and the n=1 kink mode forIAEA-CN-47/G-I-4
elongations
triangularities
aspect ratios

K < 1.8
6 < 0.5
1.5 < A < 4.

225

and

This applies to current profiles with q 0 both somewhat
above and somewhat below 1 as long as the safety factor at theplasma edge, qa, is larger than 2. For q 0 > 1, the optimum
profiles are characterised by low shear in a wide central
range and maximum shear at the plasma edge, while the pressure
gradient is only large where shear is low. This result tends
to confirm earlier conjectures [^]- Experimentation withplasmas having profiles of such a kind is presently limited toshort pulses in screw-pinch devices. For q 0 < 1, finite shear
and vanishing pressure gradient are required within the q = 1
surface to stabilise the internal m = 1/n = 1 kink mode, while
shear at the edge can be modest. In this case the improvement
in beta relies on the possibility of reducing q a along with
q 0 . An important question is thus whether in this situation
the sawtooth properties and the stability of the m = 2 tearing
mode, critical for the appearance of disruptions (see Sect.
3), are favourable.
For close to circular plasma shapes an improvement of C
by 50 % above the aforementioned value is possible [5J.
However, since in this case I/aB remains comparatively small,such shapes remain unattractive from the point of view ofplasma beta limits.
More extreme plasma shapes with larger elongation K = 2
andtriangularity (6 = 0.6 to 0.8) and some indentation(i< 0.1), are now being studied. Recently "crescent" shapes,
have been predicted to enhance the beta limit with respect toideal ballooning modes to values corresponding to C = 5 to 6[6]. The stability limits for the n = 1 kink modes are, however, not yet explored and a conducting shell on the outboard
side of the plasma might be necessary to achieve stable plasma
operation at high beta. Of course, shape control is also moredemanding in this case. Work was initiated to provide more
complete data on the ideal MHD beta limits for these shapes.
The second stability regime of ideal ballooning modes is
attractive not only because it might permit operation at
higher plasma beta, but also because it allows a larger aspectratio and lower plasma current. However, instability of
externalkink
andresistive
modes
is
a
concern.
Theoretically, the regime was found to be accessible forstrongly indented plasma shapes (i :> 0.2) [1,7] as well as
for circular and D-shaped plasmas, provided that qoJS1-5[8,9]. This implies quite demanding control requirements for
either the plasma shape or the current density profile. An
experimental demonstration of the actual existence of such aregime would be essential to provide a basis for judging thepossibilities that it may offer.

226

ENGELMANN

3- LIMITS TO DENSITY AND SAFETY FACTOR
The operational limit to the plasma density encounteredin tokamak experiments is usually expressed in terms of
Murakami-Hugill parameterM • n e (10 19 m~3) R(m)/B(T)(H)

n e being the average electron density. The maximum value of M
that is achievable can be written in the form [1 ]Mmax " «'<»

(5)

where q is the safety factor at the plasma edge and g is acoefficient that
(i)
(ii)
(iil)
(iv)

(v)

is about constant for q > 3, at least in ohmic
discharges [1];tends to decrease around small integer values ofq, in particular for q =• 2 [i, 10 ];decreases with increasing impurity content of theplasma [ 1 ];increases with increasing additional heating
power P, typically like P 1 / 2 but when additional
heating is applied, there appears to be also a
dependence on B [1, 11 ];increases with the central peaking of the density
profile, which can be generated by, for example,
pellet injection [12, 13].

However, there are quantitative differences in different
experiments which are not covered by these dependences. The
safety factor q is sometimes identified with the inverse
rotational transform and sometimes used in "cylindricalapproximation" then representing a measure of the current
density.
Inaddition,a generally
accepteddetailed
explanation of the phenomena causing the density limit isstill lacking.
The highest value of M so far achieved is 17. It wasreached in Doublet-Ill in a discharge with Ti gettering andneutral beam heating at q = 2 [ 1 ^ ].
Not considering special cases like devices with a closely
fitting shell around the plasma, tokamak operation is limited
to safety factors at the plasma edge ofboth for limiter and divertor discharges [1,10].
The density and q limits are generally disruptive. MHD
activity at the plasma edge with the m=2 tearing mode playing

IAEA-CN-47/G-I-4

227

an important part precedes plasma disruption. At the density
limit the power balance at the plasma edge also is an
important factor. In particular for plasmas with anappreciable impurity content and for not too small values ofq, the density limit can be related to radiation losses
becoming equal to the heating power, and a cool plasma edgelayer with or without poloidal asymmetries can form
[15 - 18].
From this behaviour it can be hoped that controlling the
m=2 mode may not only prevent hard disruptions, but also makes
it possible to extend the accessible density range in clean
discharges with intense heating. Low shear around the q = 2
surface is beneficial.

1. CONFINEMENT
Knowledge of the confinement properties of tokamak
plasmas is rapidly increasing. As a general trend, confinement
is degraded in tokamaks, as compared to purely ohmicallyheated discharges, when additional heating is applied, except
for very low plasma densities [i]. This, however, does notnecessarily mean that there is actually a fundamental
difference between the impact of ohmic heating and otherheating schemes on plasma confinement. Clearly, there aredifferent confinement regimes when additional heating isapplied,which
arerelatedtodifferent
profilecharacteristics, e.g.
(i) the "low confinement" or L regime, present in"normal" limiter discharges and in divertor discharges withlow heating power and/or high edge radiation in which theelectron temperature profiles are fairly peaked;(ii) the "high confinement" or H regime found in divertor
discharges, in particular at higher heating power, characterised by a high edge temperature and a broad temperature
profile;(iii) a regime in which the density profile is peaked andrecycling is kept low in limiter discharges and in which also
improved confinement properties are observed [19].
Also in high-density ohmic discharges peaking of the density
profile by pellet injection creates an alternative plasmastate with better confinement properties [12,13 3 Power losses are generally dominated by electron heat
conduction. This appears to be comparatively low in thecentral region of the discharge, where the magnetic shear issmall. Normally energy transport here is, to a large extent,caused by sawtooth collapsing; if sawteeth are suppressed orjust before sawtooth collapse, often strongly peaked

228

ENGELMANN

temperature profiles are observed (see, for example, refs.
[19,20]). Further outside, in the so-called confinement
range, the shape of the electron temperature profile isdifficulttochange("profileconsistency"[21]),
particularly in large devices and for higher heating power[22 - 27].
To make further progress, a clearer picture of theconditions leading to the various confinement regimes and abetter understanding of the transport mechanisms prevailing inthem is needed. In particular, the role of the current anddensity profiles, the importance of the plasma edge conditions(temperature, density, shear), the way in which plasma heating
affects confinement and the impact of plasma size must be
elucidated to identify the key parameters that govern
confinement in tokamaks. A specific point important in thiscontext
is to separate clearly high-beta effects from otherparameter dependences, especially from those due to theelectron temperature or temperature gradient.

5. PROFILE OPTIMISATION AND CONTROL
Profile control is expected to be a means for optimising
tokamak plasma performance. It may be a necessary or at leastadesirable tool for achieving operation at high beta andhigh density, for controlling disruptions as well as forenhancing confinement. However, the requirements to be
fulfilled for these various aims are not definitely clear andcertainly to some extent contradictory.
To reach high beta operation, it would appear thatkeeping qo somewhat above 1 and providing low shear in a wide
central range of the plasma, which allows a moderately wide,
centrally peaked pressure profile, remains a promising option.
This tends to contradict the need for low shear and high edge
temperature to minimise the MHD activity at the plasma edge
and, concomitantly, the probability for the occurrence ofdisruptions. A trade-off will therefore be necessary.
Operation at qo < 1 is of interest in this context, although
it necessarily implies that qa is small likewise.
For good confinement one can anticipate that a high edge
temperature is needed, that sawteeth have to be controlled andthat low shear is beneficial. This again points to equilibria
similar to those required for reaching high plasma beta, but
good thermal insulation of the plasma edge must also beensured. Critical issues in this picture are that somesawtooth activity may be needed to avoid impurity accumulationand that high-density operation and disruption control must be
achieved at the same time. Moreover, in a reactor, for power
exhaust a high edge plasma density appears to be necessary toIAEA-CN-47/G-I-4

229

allow divertor operation in the high-recycling regime. On theother hand, optimisation of confinement is criticallyimportant only to reach ignition, hence at modest beta, whileduring operation at high beta and with high nuclear heatingsome confinement degradation is not only acceptable, but isactually needed to ensure steady-state burn.
Active control of the current profile, at least if needed
for times longer than the skin time, requires non-inductive
current drive. Density profile control can be provided bypelletand/orneutral beam injection. The technicalconstraints (power to be injected, efficiency, accuracy
requirements) of the control systems will be an important
issue when profile control scenarios have to be developed.
In conclusion, while profile control is expected to play
an important part in optimising plasma performance intokamaks, optimisation of scenarios for reactor operation andthe definition of specifications of control devices will haveto be based on a more detailed knowledge of the impact of
profile effects (plasma current, density and temperature) thanat present available.

REFERENCES[1] INTOR GROUP, International Tokamak Reactor, Phase Two A, Part II (Rep. Int. Workshop
Vienna, 1984 and 1985), IAEA, Vienna (1986).
[2] ENGELMANN, F., BORRASS, K., Fusion Technology (Proc. 14th Symp. Avignon, 1986),
Pergamon Press, Oxford, to be published.
[3] VON GIERKE, G., et al., in Controlled Fusion and Plasma Heating (Proc. 13th Europ. Conf.
Schliersee, 1986), Part I, European Physical Society (1986) 33.
[4] INTOR GROUP, International Tokamak Reactor, Zero Phase (Rep. Int. Workshop Vienna,1979), IAEA, Vienna (1980).
[5] TSUNEMATSU, T., Japanese Contribution to the INTOR Workshop, Phase Two A, Part HI,
13th Session, March 1986.
[6] YAMAZAKI, K., et al., Contribution to the INTOR-related Specialists Meeting, Vienna,
January 1986, paper P.2.2.
[7] GRIMM, R.C., et al., Nucl. Fusion 25 (1985) 805.
[8] TODD, A.M.M., et al., Contribution to the INTOR-related Specialists Meeting, Vienna,
January 1986, paper P.2.5.
[9] SEKI, S., et al., Beta Enhancement of Tokamak Plasma with Nearly Circular Cross Section,
JAERI Memo 61-241, Japan Atomic Energy Research Institute, Tokyo, 1986.
[10] NIEDERMEYER, H., et al., in Plasma Physics and Controlled Nuclear Fusion Research 1986
(Proc. 11th Int. Conf. Kyoto, 1986), Vol. 1, IAEA, Vienna (1987) 125.
[11] NIEDERMEYER, H., et al., in Controlled Fusion and Plasma Physics (Proc. 12th Europ. Conf.
Budapest, 1985), Part I, European Physical Society (1985) 159.
[12] MILORA, S.L., et al., Plasma Phys. Controll. Fusion 28 (1986) 1435.
[13] KAUFMANN, M., Plasma Phys. Controll. Fusion 28 (1986) 1341.
[14] MAISEL, M.W., et al., Nucl. Fusion 25 (1985) 1113.

230
[15]
[16]
[17]
[18]
[19]
[20]
[21]
[22]
[23]
[24]
[25]
[26]
[27]

ENGELMANNWESSON, J.A., et al., in Controlled Fusion and Plasma Physics (Proc. 12th Europ. Conf.
Budapest, 1985), Part I, European Physical Society (1985) 147.
LIPSCHULTZ, B., Plasma-Surface Interactions in Controlled Fusion Devices (Proc. 7th Int.
Conf. Princeton, NJ, 1986), to be published in J. Nucl. Mater.
McCRACKEN, M.G., ibid.
DIPPEL, K.H., ibid.
HAWRYLUK, R.J., et al., in Plasma Physics and Controlled Nuclear Fusion Research 1986
(Proc. 11th Int. Conf. Kyoto, 1986), Vol. 1, IAEA, Vienna (1987) 51.
JACQUINOT, J., et al., Plasma Phys. Contrail. Fusion 28 (1986) 1.
COPPI, B., Comments Plasma Phys. Controll. Fusion S (1980) 261.
LAURENT, L., et al., Plasma Phys. Controll. Fusion 28 (1986) 85.
REBUT, P.H., BRUSATI, M., Plasma Phys. Controll. Fusion 28 (1986) 113.
KADOMTSEV, B.B., Plasma Phys. Controll. Fusion 28 (1986) 125.
ESIPTCHUK, Yu.V., RAZUMOVA, K.A., Plasma Phys. Controll. Fusion 28 (1986) 1253.
GOLDSTON, R.J., et al., in Controlled Fusion and Plasma Heating (Proc. 13th Europ. Conf.
Schliersee, 1986), Part I, European Physical Society (1986) 41.
FREDERICKSON, E., et al., ibid., p. 148.

IAEA-CN-47/G-I-5

INTOR: HEATING AND CURRENT DRIVE
N. FUJISAWA 1
INTOR Group2,
International Atomic Energy Agency,
ViennaAbstract
INTOR: HEATING AND CURRENT DRIVE.
Heating and current drive are critical issues for the INTOR Workshop during Phase Two A.
The paper summarizes design studies for heating and current drive, reviews the experimental andtheoretical status, describes various kinds of modelling and their application to INTOR plasmas,
and deals with design and R&D needs.

1. INTRODUCTION
The fast wave in the ion cyclotron range of frequency (IC) was selected as
the reference approach to heating up to ignition. The choice is based on thepotential technological engineering design and on the economic advantages ofRF systems with respect to a neutral beam system and on the fact that IC wave
heating is the most highly developed RF method and has shown satisfactory
heating efficiency. Neutral beam heating is the first backup option. Amongthe various heating regimes in the IC frequency range, heating at the deuteronsecond harmonic had been chosen as the most suitable regime, although otherheating methods, such as fundamental minority heating, have been widely used
and appeared to be promising. The total power requirement for heating to
ignition is 60 MW (including 10 MW for redundancy) at 85 MHz to be distributed among four modular launcher assemblies which are based on a loop antenna.
An electron cyclotron (EC) wave injection system was chosen as a necessarycomponent of INTOR for plasma formation and preheating through the radiation
barrier at the initial stage of the discharge, taking into account the limitation on
the maximum inductive loop voltage to be provided, as well as for forming acurrent profile suitable for subsequent plasma heating with the IC wave. In fact,
the formation of the proper plasma current profile might prove essential inrealizing high beta plasmas in INTOR. Furthermore, there is a possibility ofsuppressing the tearing modes and, hence, the disruptions, by local EC wave
heating. The parameters of EC startup assist and profile control are a frequencyof 140 GHz, a pulse length of 3 s, and a total power of 10 MW.
1Permanent address: Naka Fusion Energy Research Establishment, Japan Atomic
Energy Research Institute, Naka-machi, Naka-gun, Ibaraki-ken, Japan.
2
The members of the INTOR Group are identified in Paper No. IAEA-CN-47/G-I-1.

231

232

FUJISAWAThe use of lower hybrid (LH) waves in INTOR was considered at theoutset for various functions (heating to ignition, current drive, startup assist),
but was not included in the reference concept. The main reason for this wasthat LH wave studies were not sufficiently advanced for extrapolation to reactor
conditions. It was, however, recognized that the use of LH waves for current
drive was the most promising choice. Although the use of LH waves for current
drive during the burn pulse was estimated to require too high a power for INTOR
parameters, application for current rampup and sustaining the current duringtransformer recharging was reasonable. The potential attractiveness of LH
waves, in particular for current drive applications, is considered sufficiently highthat special emphasis has been given to it during Phase Two A.
2.

PLASMA HEATING TO IGNITION

2.1. Experimental status
Considerable progress has been made since the start of Phase Two A inclarifying and verifying specific aspects of IC heating physics. Specifically,
detailed information on coupling, absorption, and thermalization of the wave
energy,has been gathered from a variety of experiments. These experimental
results have been used to verify existing theoretical models, to a certain extent,
and they have further raised the confidence level for the use of IC heating in INTOR.
The Faraday shielded loop has remained the most widely used method of
coupling to the fast wave in present day experiments. Typical experimental
values of the coupling efficiency range from 80 to 95%, depending on the specificcoupling geometry. The device size and plasma density in these experiments are
sufficiently large that no isolated eigenmodes can be observed in most, though
not all, devices. Moreover, the wave damping rate is strong enough to preclude
the formation of toroidal eigenmodes.
Results of high power experiments have been reported which tend toqualitatively confirm the mode conversion/minority heating model in TFR,
JFT-2 and T-10. Notably, PLT has achieved remarkable heating results to date,in a D-3He plasma with 3He as a 5-10% minority. Direct second harmonic
heating of the majority species has also been achieved, with efficiencies
comparable to those for the minority heating regime.
Associated with these heating results is a degradation of the energy
confinement, which appears to be highly device dependent. For the highest
power results on PLT, the confinement time decreases moderately from Ohmicvalues as the applied auxiliary power is increased. Another aspect of the effectof IC heating on confinement is observed in the electron heating results. The
application of power causes an increase in the sawtooth amplitude and period as
observed in JET and PLT. While this heating can be roughly accounted for bycollisions with ions, direct heating of the electrons by the waves may also takeplace.

IAEA-CN-47/G-I-5

233

All IC heating experiments to date have indicated an increase in the plasma
impurity level and consequent radiation losses. Impurities were thought tooriginate from the walls and limiter but also, to some degree, from the antennas.
To reduce the impurity influx level, an appropriate choice of the Faraday shieldmaterial was made in many devices. After installing carbon shields and limiters
in TFR, the intensity of metallic impurity radiation was reduced ten times, andperipheral low-Z radiation increased by a factor of two. There are several likely
causes of this impurity generation; further work is, however, necessary to
clarify the causes of impurity generation and their possible suppression.
2.2. ModellingOn the basis of the growing body of experimental information, considerable
progress in the development of the theoretical model has also been made during
Phase Two A. In particular, the discovery that wave absorption is sufficiently
strong that the wave power is nearly completely absorbed on a single passthrough the resonance zone has led to widespread use of the single pass model
to predict heating results in large tokamaks. Refinements of this theory havebeen made, including the effects of a finite sized (3D) antenna geometry via ray
tracing or global wave solution techniques. Once the waves have been absorbed,
thermalization of the wave power has been modelled by one- and two-dimensional
Fokker-Planck codes including the effects of RF induced velocity space diffusion.
The power absorption profile having been determined in this way, the results areused in conjunction with one-dimensional transport codes to assess the IC heating
results and the effects of heating on plasma confinement.
Modelling has been carried out for the case of 3He minority heating, which
is facilitated by strong absorption at the cyclotron layer. In the limit of small
concentration, ray tracing or even single pass absorption models have been
applied to produce the IC power deposition profile. Such a model, coupledwith the Fokker-Planck and radial transport codes, has, with some success, beenused to model results in PLT. Both the spatial localization of the IC power andthe equipartition of the energy among the plasma species have been found to bein agreement with the experimental results. In the second harmonic heating
regime, simulations based on one-dimensional velocity space models haveindicated that electron heating should become significant and comparable tothe ion heating as the temperature is raised above a critical value. This feature
is consistent with the experimental tendency on JFT-2M, although somediscrepancy exists on the details of wave absorption.
The overall agreement with the existing experimental results is good, as far
as equipartition of the RF power among the various plasma species is concerned,
although many details of the experimental observations remain unmodelled, suchas power deposition profiles, plasma-wall interaction and the generation of
impurities.

234

FUJ1SAWATABLE I. BENCHMARK SINGLE PASS ABSORPTION CALCULATIONSFOR INTOR(Parameters: RQ = 5.3 m, circular equivalent cross-section with a = 1.52 m,
50:50 D-T mixture, no impurities, parabolic profiles of plasma temperatureand density with finite pedestal, f = 85 MHz. kj = 10m" 1 , antenna outboard,symmetrical about midplane, angular range: within ±40°)Power fraction
absorbed insingle pass
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EC
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Case 3:T(r = 0) = 20keV,T(r = a) = 0.1 keV;
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EC
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Case 4:T(r = 0) = 20keV, T(r = a) = 0.1 keV;2O
3n e (r = 0) = 2Xl0 rrr ,n e (r = a) == O.5XlO"m- 3

EC
USSR
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* Fine structure in ki.

IAEA-CN-47/G-I-5

235

Modelling applications to INTOR encompass coupling, propagation anddamping of the fast wave. These results are then incorporated into a FokkerPlanck code, coupled to a one-dimensional radial transport code to predict theoverall heating expected in INTOR.
A comparison of single pass calculations has been made by using models
formulated by each of the participants; the results are shown in Table I. Overall
agreement was achieved on the relative equipartition of RF power among thevarious plasma species. Further studies indicated that second harmonic heatingof deuterium can be dominated either by direct electron heating at sufficiently
high k|| values or by minority proton absorption if a small concentration ofhydrogen is present as an impurity. For direct deuterium heating, it was observedthat the proton concentration must be kept below a specific value. The optimum
k|| for heating during the pre-ignition phase is found to be 2 to 5 rrT 1 , which canbe accomplished by the 2 X 2 array planned for INTOR.
Global field models have been developed, including the effects of mode
conversion, finite sized (3-D) antennas and toroidicity. Global field solutions
for the D-T-H case in INTOR indicate adequate radial focusing of the wave
power accomplished-by localized absorption near the proton fundamental
cyclotron layer, as is shown in Fig. 1. Power deposition tends to be peaked
but heating is smoothed by radial transport. Ignition is expected to be reachedwithin the planned heating time of 6 s. Because of the relatively high density
and large plasma volume, ion tail formation is expected to be minimal, provided
excessive wave focusing is avoided. In this model, no explicit degradation ofconfinement with IC power was assumed.
2.3. Launcher design
The INTOR bulk heating system using IC waves is required to provide a
power of 6 MW (including 10 MW for redundancy) for 10 s at a frequency of
85 MHz, using a minimum number of access ports. The wave mode injected
into the plasma is the fast magnetosonic wave: the slow mode must be suppressedby an electrostatic shield (Faraday screen) to avoid heating of the plasma edge.
The requirements for the wave spectrum to be launched have not yet been fully
clarified.
There are four candidate launcher concepts for this heating system:(1) a ridge loaded waveguide, (2) a U-slot, (3) a resonant cavity, (4) a loop
antenna. Four general areas were considered in evaluating the suitability ofthese concepts for INTOR: three interface issues (plasma, device and transmission),and the overall impact on the design. The results of the evaluation are summarizedin Table II.
In general, design studies indicate that a loop antenna and a ridged waveguidearray are technically promising, but a great development effort is required. The
concept of a ridged waveguide array requires the demonstration of good coupling236

FUJISAWAIra E uPoyncing vector

IAEA-CN-47/G-I-5

237

to the plasma for large RF power. For the loop antenna concept, the selectionof materials and the precise shape of the conductors for the Faraday shield aswell as the optimum position of the ceramic vacuum window remain to beresolved. A more detailed optimization of the conceptual designs as well as abetter database to come from the R&D studies are necessary before a reference
choice of an IC launcher for INTOR can be made.
2.4. Additional R&D needs
While the basic physics underlying IC heating is considered to be generally
well understood in the ongoing programmes, many details of the heating process
as well as the effect of heating on the confinement remain to be clarified. In
addition, major technological improvements are required to permit the use ofhigher power levels under the expected edge conditions in INTOR.
Specifically, in the following physics areas additional research is needed:(a) scaling of confinement with IC heating, including effects on ion transport,
fast-ion losses, stability, impurity control, and edge phenomena, (b) scaling ofwave absorption, including the effects of electron absorption, alpha particle
absorption, k\\ optimization of the launched spectrum, and competition between
resonant ion species.
Additional technological development is required in the following areas:(a) the development of reactor compatible launchers, including Faraday shields
with attention to the problems of heat removal, radiation damage, power
handling capability, voltage breakdown, maintainability, and component
longevity, (b) the development of generator and transmission line components
for long pulse and high power levels.

3.

STARTUP ASSIST AND PROFILE CONTROL

3.1. Startup assist
In earlier INTOR phases, an RF system for startup assist appeared to berequired, because of the limitation of the maximum loop voltage that can beprovided by a superconducting transformer solenoid. Recent experiments on
JET have shown that if a maximum inductive loop voltage of 35 V is provided(as foreseen in the reference startup scenario for INTOR), there would be noneed for an RF startup system. Nonetheless, since a further reduction in theloop voltage would allow the simplification of the design of the transformer
and of the discharge chamber, EC startup assist remains an important point ofinterest.
Experiments on plasma formation with EC waves were done in various
toroidal systems, both tokamaks and stellarators. In the tokamak experiments,

238

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IAEA-CN-47/G-I-5

239

it was demonstrated that, by using low microwave power densities (typically,
20 to 100 kW-rn"3), the filling gas could be ionized and heated up to a bulk
temperature of about 10 eV. The densities of the plasma thus produced spana wide range, typically from 10 18 to 2 X 10 1 9 m" 3 . Further increase of thepower generally leads only to a very small improvement in plasma quality,
because of the lack of confinement in a currentless tokamak. In the experiments
done on small and medium sized devices, a two to four times reduction of theloop voltage required for startup was observed in the presence of EC waves.
This reduces the number of runaway electrons leading to a decrease in the hard
X-ray emission as well as in the impurity radiation during RF assisted startup.
In recent experiments on the WT-2 and JIPP T-IIU tokamaks, current rampup
was done with LH waves applied to a target plasma produced by EC waves,
without applying an inductive loop voltage.
To describe the startup phase of INTOR discharges, 0-D models are usedwhich differ in the details of treating losses during the ionization stage. Thesemodels typically underestimate energy losses and yield significantly higher
values of the electron temperature at the beginning of current rise than thoseachieved in experiments. Additional, ad hoc losses having no clear physics
justification are therefore added to cure this discrepancy. Such an approachprovides results which fit the experimental results, but extrapolation to INTOR
with such models remains unsatisfactory. Computer simulations of the initial
current rise show that for INTOR reference conditions the maximum loop
voltage does not exceed a value of about 20 V, provided an RF power of about
5 MW is applied during 0.2 s. The resistive voltage is only a few volts. To
determine, however, the minimum inductive loop voltage, as well as the EC
power required in INTOR for startup assist, further work is necessary inrefining the model and checking it against experiments.
3.2. Profile control
Recent studies of plasma heating by EC waves in medium sized tokamaks
are of significant interest for the INTOR project, in particular, in the light ofthe possibility of controlling the profile of plasma current and pressure andthereby affecting the MHD activity of the plasma column. The parameters of theplasma in present devices are similar to those expected during startup in INTOR,
so the experimental results in these devices allow extrapolation to INTOR. In
addition, the most advanced EC wave technology available at present is used inthese experiments.
Two tokamaks operating with gyrotrons of higher frequency (Doublet III
and T-10) have recently demonstrated efficient plasma heating with input power
levels of about 1 MW. Both experiments show high heating efficiency (the power
absorption in T-10 is 70 to 75% of the launched power) and relatively insignificant
deterioration of the energy confinement with respect to the Ohmic regime.

240

FUJISAWA

Experiments on Doublet III and T-10 also confirm the possibility of controlling
the MHD activity by means of EC heating. An effect of the position of the wave
resonance on the sawtooth oscillations was observed on both. In T-10, heating
just outside the m = 2 resonant magnetic surface led to a sharp decrease in theamplitude of MHD oscillations. Thus, the experiments indicate a possibility ofsuppressing the m = 2 mode, which is responsible, according to current thinking,for the development of disruptions by local EC wave heating.
Combination of geometrical optics with linear wave damping along the ray
allows the calculation of the power deposition profile across the column. Calculations of this kind for specific systems are now in common use. Variations of
the toroidal magnetic field, plasma density, temperature, and current can betaken into account. The comparison of results of linear absorption theory withexperiments carried out at low (less than Ohmic) EC power levels shows
reasonable agreement although, so far, no thorough analysis of non-thermal
effects has been carried out. The T-10 experiments on m = 2 mode stabilization
are in qualitative agreement with the theoretical expectations, but a quantitative
comparison is still lacking. Transport code simulation of a local EC heating
effect on sawtooth oscillations reasonably fits the Doublet III experiment.
The model included ray tracing using the cold plasma dispersion relation,
relativistic local power absorption, and 1-D energy transport; it failed, however,to describe effective heating of plasma with a density above the critical value
for high power launch into Doublet III, while there was acceptable agreement
for low input power.
3.3. Launcher design
The electron cyclotron startup assist system for INTOR is required toprovide a power of 10 MW for 3 s at a frequency of 14 GHz in the ordinary
wave mode. The main components of this system are the gyrotrons, thetransmission line and the coupling system.
A few different concepts of the launching system for INTOR were
proposed. In one concept, quasi-optical power transmission over a distanceof about 100 m by oversized waveguides of large cross-section (500 X 1000 mm 2 )with metallic mirrors at the bends, and a coupling system with tunable mirrors
have been considered. This concept offers the possibility of placing the
gyrotrons outside the torus hall, of providing more flexibility in beam deflectionin the plasma, and of potentially smaller power losses in the transmission line.
Double ceramic gas cooled windows are proposed to serve as vacuum andtritium barriers. However, transmission of EC waves by a quasi-optical systemdoes not yet have a database.
The design studies performed show the basic feasibility of an EC wave
launcher meeting the INTOR requirements. A more detailed analysis of assemblyand maintenance, of wave physics issues, and of reliability in conditions of large

IAEA-CN-47/G-I-5

241

neutron and heat loads will have to provide the basis for the choice of an
optimum launching system. The most critical technical problem is the development of a reliable window for the transmission line, apart from the necessity ofproviding RF power generators of high efficiency for the frequency range inquestion.
3.4. Additional R&D needs
The successful completion of the ongoing programmes on using EC waves
in tokamaks appears to be able to provide a level of physical understanding
sufficient for INTOR. Determining the scaling, with the machine size, of thepower necessary for plasma profile control to stabilize MHD modes requires
specific emphasis. On the contrary, none of the EC experiments, either existingor currently under construction, addresses the problems of long component
lifetime and operation under conditions of thermonuclear burn. Therefore, in
addition to the existing programmes, the following development work isnecessary: (a) studies of materials properties under these conditions, foremost
the search for a suitable window material, (b) design and experimental testsof long power transmission lines (waveguide optimization, tests of quasi-opticalsystems), (c) development of a fault protection system.

4.

CURRENT DRIVE

4.1. Experimental status
A great effort was made to investigate LH current drive in steady andtransient states. The experimental (normalized) current drive efficiency, atconstant current and in the absence of an electric field, was found to be
19
3
T? CD = n(10 nT )R(m)I p (MA)/P(MW) » 1 in the best cases (ALCATOR-C,
ASDEX, JFT-2M, PETULA, PLT, T-7). In the presence of a small residual
electric field (loop voltage: 0.1 V) in the current direction, in PETULA an
enhanced current drive efficiency of 7}CD « 3 was observed. The duration of
non-inductive current drive is only limited by the capability of the powersources and has reached several seconds. The largest currents sustained by
LH waves are 400 kA in PLT and ASDEX. In PETULA, under appropriate
current drive conditions, the excitation of the m = 2 mode could be strongly
reduced, allowing for disruption free operation of discharges with a safety
factor, at the limiter, as low as q a «* 2. The suprathermal population extends
up to the largest parallel phase velocities of the waves excited (a few hundred keV),and a substantial population of this kind is always created under current drive
conditions, even if the wave spectrum injected does not extend down to close
to the electron thermal velocity, i.e. to high N|.

242

FUJISAWA

Current rampup by LH waves was demonstrated in various devices(ALCATOR-C, ASDEX, JIPP-T II U, PLT, WT-2). The conversion efficiency of
RF energy into poloidal magnetic field energy depends on the experimental
conditions, and values of up to 25% were reached. In the presence of an electricfield along the current, current rampup is considerably more efficient than whenthe electric field is absent or in the opposite direction, although the separationof inductive and non-inductive effects for this case is still under investigation.
In ASDEX, a first convincing demonstration of sustaining the plasma current
by LH waves during transformer recharging was given. A transformer flux of 0.2 V- s
was generated for over 1 s, corresponding to an average negative loop voltage
of 0.2 V, applied by ramping up the current in the transformer coils.
4.2. Modelling
As a theoretical basis for describing LH wave coupling, Brambilla's linear
theory provides the basic information on the coupling efficiency and the spectrumin N|| that is launched. Wave propagation can be predicted from the dispersion
relation, by using a WKB approach including ray tracing techniques.
The theoretical basis for describing LH current drive quantitatively is
Fisch's quasi-linear theory which in its original form is limited to steady state
conditions, vanishing DC electric field and strong wave excitation. For the(global) normalized current drive efficiency, taking into account that theaccessibility condition limits the wave spectrum in NJJ which can penetrate into
the plasma from below to Ny > 1.5, this theory yields, as a maximum value forclean plasmas (Z e f f « 1): T? CD =R (10 19 m- 3 )R(m)I p (MA)/P(MW) = 2 to 3.
Still higher current drive efficiencies might be possible for optimized shapesof the wave spectrum.
In the steady state current drive in the absence of a DC electric field, below
the critical density for switchover of the wave coupling to the ions, the current
drive efficiency well follows the above scaling, but in absolute terms the experimental results, as was mentioned, rather correspond to TJQO *» 1 than to the
maximum value of TJ^D = 2 to 3, as predicted by theory. Experimentally,a substantial suprathermal population is always found, which covers continuously
the velocity range from a few times the thermal velocity up to the velocities
resonant with the waves, even if the wave spectrum as calculated does notextend to high Ny. Various mechanisms are under consideration to explain
this phenomenon. Direct measurements of the wave spectrum in the plasma
would be necessary to clarify this problem. The results of the theoreticalanalysis of current rampup experiments and current sustainment duringtransformer recharging are consistent with the experimental findings, although
the models have been based on different simplifying assumptions. A moredetailed comparison of the models among themselves and with experiments
is necessary to develop a satisfactory understanding of the underlying physics.

IAEA-CN-47/G-I-5

243

The above mentioned models were applied to INTOR. Three different
models were used. All of them are 0-D in space. One is based on the quasilinear evaluation of current drive in the absence of an electric field and includes
a power balance and circuit equations. The two others contain electric field
effects. One of them again uses a (one-dimensional) quasi-linear approach forevaluating kinetic effects and includes a power balance, while in the other theplasma temperature is taken to be a free parameter and kinetic effects aretreated by a two-dimensional test particle model. Nevertheless, the obtained
results converge to a considerable extent.
The possibility of non-inductive current drive during plasma burn, to allow
for steady operation, although most desirable, remains uncertain, because thedatabase for current drive in high density plasmas, requiring comparatively
high wave frequencies, is still very limited. Extrapolating present experimental
and theoretical results to INTOR reference conditions would lead to very highpower requirements.
In contrast, non-inductive current rampup and current maintenance duringtransformer recharging at low plasma density and temperature have been confirmed
to be promising. Then, inductive current drive during burn over 1000 s, or more,
becomes a possibility with the present transformer capability of INTOR, allowing
for a duty cycle approaching 90%. The additional physics risks of this scenario
lie in the extension of the burn pulse length to the global classical skin time ofthe discharge without a means of controlling the current density profile being
available, in the necessity of low density operation for rather long times wherea high density/low temperature divertor plasma cannot form and the plasma istransparent to neutrals, and finally, in the problems related to the still incomplete
understanding of lower hybrid current drive in the presence of an electric field.
Therefore, it is recommended that INTOR be prepared to work with thisadvanced scenario if it proves to be viable, but INTOR should remain able toreach its objectives with purely inductive current drive and a burn pulse lengthof 200 s.
It is recommended to include a lower hybrid system with the following
specifications in the INTOR concept: power of 20 MW, frequency range of 0.5
to 2 GHz, pulse length of CW, N\\ of 5 to 3 (to be optimized; for lower frequency,a large N|| is admissible).
4.3. Launcher design
Different types of launchers for the lower hybrid waves have been consideredduring the INTOR study. All are based on the same principle, i.e. that of theconventional grill, a phase array of open waveguides stacked along their broadsides with apertures facing the plasma as used in present experiments. In the
launcher design, the requirements to be met are considered, such as specificationof the launched spectrum, RF matching of the launchers to the edge plasma,

244

FUJISAWARF power consideration, vacuum and tritium barrier, and mechanical stabilityof the configuration.
4.4. Additional R&D needs
There are no physics fields which are not covered by the ongoing andplanned programme. This includes startup assist and current initiation, inparticular, the development of a theoretical basis, the details of wave penetrationand coupling to the ions in the ion heating regime, and the impact of high power
LH launch on the edge plasma.
On the other hand, beyond the ongoing engineering R&D programmes,
it is necessary to make a special effort in two areas: (a) the development ofvacuum windows for the waveguides, complying with the constraints of areactor environment, the requirements of tritium containment, and the presenceof a large number of separate waveguide feeders; (b) the development of techniques of waveguide conditioning for waveguides working under reactor conditions
and, in particular, close to the thermonuclear plasma.

5.

CONCLUSIONS

On the basis of the studies carried out in the present phase of the INTOR
Workshop, IC heating at the deuteron second harmonic (85 MHz) is selected
as the reference method for heating to ignition (50 MW, plus 10 MW for
redundancy). The conclusion of the modelling effort is that central power
deposition is likely throughout the heating phase. While the Ohmic phase in apure D-T plasma is not expected to exhibit single pass absorption, the presenceof a small amount of hydrogen will, in fact, provide complete single passabsorption even in the Ohmic case; this will, however, cause hydrogen absorption
to dominate throughout the heating phase. In launcher studies, attention was
focused on the use of a Faraday shield with its associated requirements forheat removal and hardening to radiation and particle bombardment. At thepresent stage in the investigation, experimental data exist only for the loop
antenna although the ridged waveguide appears to have potential technicaladvantages.
An EC startup assist system remains a desirable, if not necessary, componentof the INTOR design. In addition, the same system might be useful for providing
current profile control during the Ohmic phase of the discharge, aiding in theprevention of disruptions and the achievement of high beta values. The present
status of experimental and theoretical investigations suggests that using electron
cyclotron waves is the most proper choice for this purpose. Conceptual design
studies of EC wave launchers show that such systems meeting the INTOR

IAEA-CN-47/G-I-5

245

requirements could be built, provided some critical components are developedin a timely manner.
Using lower hybrid waves in INTOR is not a reference option, but remains
an attractive option. As far as non-inductive current drive applications areconcerned, lower hybrid waves are the most highly developed method, andtheir use on a tokamak reactor does not appear to encounter major technicalobstacles, although both in physics and in technology more work is needed toassess all constraints.

ACKNOWLEDGEMENTS
This paper is based on the work of scientists and engineers in the EC, Japan,
the USA and the USSR who have contributed to the INTOR Workshop.

REFERENCE[1 ]INTOR GROUP, International Tokamak Reactor, Phase Two A, Part II (Rep. Int.
Workshop Vienna, 1984 and 1985), IAEA, Vienna (1986).

DISCUSSION
M. PORKOLAB: I am inclined to disagree with you that the problem
with lower hybrid current drive is that it is limited to low densities. From
Alcator C, which I described in my paper, and from our previous results, it isclear that current can be driven even at densitiesHe ~ 1 X 1020 m~3 if thefrequency is high enough. Our studies indicate that the limiting factor for LH
current drive will be electron temperature. Accessible lower hybrid waves will
be absorbed in reactors at Te ~ 10 keV, which will restrict the waves towards
the outer edge regions.
N. FUJISAWA: Thank you for the comment. We really need more data
before we can finally establish that point.

Session H
TECHNOLOGYAND REACTOR CONCEPTS
Chairmen of Sessions
Session H-I

K. TOMABECHI

Japan

Session H-II

R.W. CONN

USA

Session H-III (Posters)

IAEA-CN-47/H-I-1

CONCEPT AND PARAMETERS OF NET
NET TEAM(Presented by F. Engelmann)
Max-Planck-Institut fur Plasmaphysik,
Garching, Federal Republic of Germany
Abstract
CONCEPT AND PARAMETERS OF NET.
The objectives, main features and reference parameters of NET are presented. The
physics assumptions on which the device is based are described. Possible alternative options
that could enhance performance and illustrate the range of working conditions under consideration are also touched upon.

1. OBJECTIVES OF NET
The Next European Torus (NET) is conceived as anexperimental fusion reactor to follow JET after completion ofits mission in the European Fusion Programme. It is aimed atdemonstrating the feasibility of fusion energy production in aplant based on the tokamak principle which integrates all
essential components and technologies of a fusion reactor
[i].
This
impliesthat
NET has both
physical andtechnological objectives.
The physical objectives are to produce a plasma withparameters and performance relevant to a fusion reactor. This
includesgenerating an ignited plasma in a controlled way, andoperation with extended (up to 1000 s) and reproducible
burn pulses.
In general terms, the technological objectives consist of
an integrated test of reactor components and systems as wellas engineering tests and development of in-vessel components(plasma-facing components such as the first wall and power
exhaust devices, as well as blanket modules). Specifically,
they can be summarized as
selection and qualification of design concepts which meet
the basic performance requirements of a demonstration
reactor,testing of materials and components in an integrated way
in a fusion environment,
testing of tritium and power extraction under reactorrelevant conditions,demonstration of the maintainability of a fusion reactor,

249

250

NET TEAM
demonstration of plant reliability at levels relevant toa demonstration reactor, anddemonstration of the safe and environmentally acceptable
operation of a fusion reactor-like plant.

A total neutron fluence of about 0.8 MWa/m2 and aneutron flux of about 1 MW/m2 onto the first wall are judged
to be necessary to achieve these objectives. The physicalobjectives will be addressed in the first phase of NET
operation, during which a neutron fluence of the order of 0.1
MWa/m2will be produced. Significant testing of thecomponents of the basic machine will also take place during
this phase. In particular, the performance of the plasmafacing components will be assessed. Moreover, first tests ofblanket modules will be performed. The later phases of NET
operation will be concerned with an extended performance
testing of the NET components and medium term endurance
testing of blanket sections.
2. CONCEPT EVOLUTION AND TIME SCHEDULEWork on NET was started in 1983. After a definition
phase, during which various concepts and parameter choices
were investigated, pre-design work is now under way for two
variants of a reference concept. In addition, some effort isbeing dedicated to the analysis of alternative options(parameter choices and operation scenarios) with the aim offurther enhancing the performance characteristics of thedevice. Development of the reference concept will keep abreastof progress in tokamak physics and technology. It is
anticipated that in the preparation of NET there will be two
important check points: in 1989 before a detailed design canbe started, and in 1993 before construction of the device is
proposed. Such a schedule would lead to commissioning of thedevice at the end of the next decade.
3. REFERENCE CONCEPT AND PARAMETERS
Before selecting the reference parameters of NET
extensive optimization studies were performed. In theseinvestigations scoping studies done with the system code
SUPERCOIL [2,3] which took the main physical and technicalconstraints into account were combined with quite detailed
design studies for a number of different parameter sets.
A system integration scheme that allows assembly anddisassembly of in-vessel components in the oblique direction
from the top of the device was chosen (see Figs 1 and 2)
since this allows the overall size of the apparatus to beminimized. Both the toroidal and the poloidal magnetic field
Coils are superconducting. The superconducting material of the251

IAEA-CN-47/H-I-1

10R0IDAL FIELD COILSINNER POLOIOAL COILSBLANKEI
PLASMAFIRST HALLOUTER POLO1DAL COILS

D1VERTOR PLATE5_
PLASMA EXHAU5T

FIG. 1. Reference device NET SN (cutaway model).

TOROIDAL FIELD COILSBLANKETj

INNER POLOIDAL C0IL5

FIRST WALL;ii

SHIELDI

CRYOSTATPLASMA
OUTER POLOIDAL COILSi

PLASMA EXHAUST

1;. i... •VACUUM PUMP

DIVERTOR PLATESACTIVE CONTROL COILSi!
i

FIG. 2. Reference device NET DN (cross-section).

252

NET TEAM

16 toroidal field coils is l^Sn, while the poloidal fieldcoils could be made of Nb Ti or f^Sn.
The reference choice for the plasma shape as well as forcontrol of plasma-wall interaction and for power and particle
exhaust is an open poloidal divertor configuration (see Fig.
2). Large elongation and triangularity of the plasma column
are adopted. The variants being considered relate to theplasma and divertor configurations: the NET DN reference
concept
has an up-down
symmetric
plasmacolumn andconsequently a double-null divertor, whereas NET SN has asingle-null divertor, the plasma column being asymmetric.
The main working assumptions adopted for tokamak plasma
performance are:(i)the maximum plasma beta that can be reached is
3(ii)

max i%)= 3>5 I

(MA)/a("l) B(T)5the maximum electron density nobtainable is

6

n

(iii)

e,max

(m *) = 1.6X1(T U B(T)/R(m);the minimum for the safety factor
qj = 5 K 2 a(m) 2 B(T)/R(m) I(MA)that can be admitted isql,min " 2 - 1 ;(iv)

operation at the optimum temperature
T = 10 keV
is possible (the ion and electron temperatures
being about equal);(v)

the confinement capability of the device issuch that
P /Pn
= 2.9
a lossfor a power loss PiOss from the plasma extrapolated with the ASDEX H-mode scaling in theform proposed by Gruber [n], i.e.
TE(S)= 0.1 I(MA) R(m);IAEA-CN-47/H-I-1
(vi)

253

the toroidal field ripple at the outboard
plasma edge must not exceed6

max= ± 1.5 %(maximum compared toaverage field).

In the preceding relations the symbols,used have the same
meaning as in Table I; the quantity K = <p dl/2na is a form
factor characterizing the shape of the poloidal cross-sectionof the outermost closed magnetic surface.
The reference operation scenario relies on inductive
current drive, the burn pulse having a duration of at least
200 s and possibly up to 1000 s. The times needed for start-up
and shut-down are 25 s each. A dwell time of 20 s between two
consecutive discharges is taken. In accordance with this
scenario the number of pulses needed to reach a neutron
fluence of 0.8 MWa/m 2 is set at 1 0 s .
The divertor configuration is formed before heating to
ignition starts [ 5 ] . It is anticipated that, during the burnpulse, the divertor operates in the high-recycling regime.
This requires the plasma density at the edge of the discharge
to be rather high ; typically 0 . 5 M 0 2 ° / m 3 , if a large fractionof the power is to be exhausted into the divertor. Under these
conditions, the most demanding requirements for the first walland the divertor plates result from the occurrence ofdisruptions although the disruption specification adopted is a"mild" one: the duration of energy quench is taken to be 2 ms
and that of current decay 20 s, the frequency of occurrence ofdisruptions under routine operation conditions being as low as
10~ 3 . Another critical issue is peaked loads on the firstwall due to localized losses of fusion a-particles, which
could be quite high for the toroidal field ripple limit
specified above [ 6 ] .
The main configuration and plasma parameters of the NET
DN and NET SN reference devices are given in Table I.
H. ALTERNATIVE OPTIONS
With
thereference
optionsasstarting
point,
alternatives and additional variants are being investigated toanalyze possibilities of enhanced performance as well as theflexibility inherent in the reference devices with respect toconfiguration and size of the plasma and the operation
scenario. Specifically, it is being considered whether it ispossible to accommodate a larger plasma with higher plasma
current and allowing various plasma configurations (double andsingle-null D-shaped configurations as well as mildly indented
plasma columns) in the NET DN reference device reducing the254

NET TEAMTABLE I. MAIN CONFIGURATION AND PLASMA PARAMETERS
OF THE NET REFERENCE OPTIONS NET DN AND NET SN
NET DNNET SN

5.18

5.79

Plasma major radius

R (m)Plasma minor radius

a (m)

1.35

1.53

Plasma elongationb/a2.183>

1.6i/1.83a)

Plasma triangularityY

0.65a)

0.29/0.71a)

Aspect ratio

A=R/a

3.8

3.8

Plasma volume

ra3)

390

110

Total radial machine size

V2 R (m)
t

21.10

23.50

Total machine height

Ht(m)

11.00

12.80

Magnetic field on axis

B(T)

5.0

5.8

maximum on coils

Bm(T)

10.1

11.6

Toroidal field ripple ^

& (%)

+ 1.2

± 1.5

Plasma current

I (MA)

1

Safety factor

10.8

9.9

2.1

2.1
10

Average burn temperature

T (keV)

10

Average DT densityfi^dO^/m3)

1 .20

1.13

Average electron densityc)Murakami parameterHe (10V) 1.56

1 .17

M(10l Vm2.T) 16

15
3.5

Beta scaling factorg

3.5

Total beta

B

5.6

3.9

DT beta

1.2

2.9

Poloidal beta

1.12

1.39

600
fus ( M W )120
P (MW)
a-particle powera3
P(MW/m)Fusion power density
1.53
r
fus
2
Average neutron wall loading Q n (MW/m )625

Fusion power

Confinement capability

tot ( ? )P

C=P /P..
a lossConfinement time during burn T E (S)Burn time (minimum)

,„<•)

2.9

d)

125
1.12

d

L>

2

2

200

200

a) based on location of null point; for NET SN lower half.
b) at outboard plasma edge; maximum compared with average
field
,
1Q
c) defined as M = H e d O 1 9 m" 3 )R(m)/B(T).
d) based on ASDEX H-mode scaling [ i ] .

IAEA-CN-47/H-M

255

TABLE II. CHARACTERISTICS OF NET DN WITH ENHANCED
PLASMA SIZE

Plasma major radius

R(m)

5.41

Plasma minor radius

a(m)

1.68

Plasma elongation

b/a

2.17

Plasma triangularity

Y

0.61

Aspect ratio

A = R/aV (m )

600

Magnetic field on axis

B(T)

4.8

Toroidal field ripple

&(%)

±3.4

Plasma current

I (MA)

14.8
2.2

Safety factor

Average DT density

T(keV)
H D T d0

ryiO

10

20

20

Average electron densityc)Murakami parameter

M(10Beta scaling factorg

Total beta

e tot (*>DT beta

B DT (*>

19

3

/m )
3

An )
2

/m -T)

1.26
1 .64
18.5

3.5
6.4
4.8
1.2

Poloidal beta
Fusion power

Pfus(MW)a-particle power

P (MW)

1000
200
3

1.7

Average neutron wall loading Q (MW/m )1.4

C=P a /P nloss

4.4

Fusion power densityp"us(MW/m )2

Confinement capability

a)

3.2

3

Plasma volume

Average burn temperaturea)d)a) based on location of null point.
b) at outboard plasma edge; maximum compared with average
field.
c) defined as M = n e (10 19 m"3) R(m)/B(T).
d) based on ASDEX H-mode scaling [ 4 ] .

256

NET TEAM

thickness of the blanket/shield structure [7]. This is ofinterest in particular for the first phase of operation
dedicated to physics investigations. The characteristics ofsuch a case are given in Table II. The possibility of using
ferromagnetic inserts to reduce the toroidal field ripple atthe plasma edge is being studied [8], to avoid enhanced
losses of highly energetic ions. Note also that the density
limit tends to become more critical in this case. The use ofnoninductive current drive to extend the burn pulse durationis also under consideration. This includes an assessment of
the potential and the implications of applying non-inductive
current
drive
for current
ramp-up and/ortransformer
recharging at low plasma density to enhance the inductive
voltseconds available during burn [7].

REFERENCES[1] THE NET TEAM, NET Status Report, Commission of theEuropean Communities, NET Report 51 (1985),[2] BORRASS, K., Scoping Studies for NET, Commission of
the European Communities, NET Report 47 (1985); to be
published in Nucl. Engineering and Design/Fusion.
[3] BORRASS, K., SOLL, M., Nucl. Engineering and
Design/Fusion 4 (1986) 21.
[4] GRUBER, 0., Plasma Physics (Proc. Int. Conf. Lausanne,1984); invited papers, Vol. I, CEC Brussels and CRPP
Lausanne (1984) 67.
[5] MITCHELL, N., SALPIETRO, E., Plasma Engineering for the
Next European Torus; to be published in Nucl. Engineeringand Design/Fusion.
6] TANI, K., et al., Nucl. Fus. 23 (1982) 659.
l\ MITCHELL, N., SALPIETRO, E., NET Report to be published8] MITCHELL, N., PASOTTI, G., RICCI, M.V., Fusion Technology(Proc. Symp. Avignon, 1986), to be published.

DISCUSSION

B. COPPI: There are several parameters of merit that those of us who havebeen involved in the design of ignition experiments consider as an indication of
the confinement quality of the plasma column produced by a given machine.
One of these is I p B p or 'confining strength', where I p is the plasma current
and B p the mean poloidal field. From your list of parameters, NET would have
IpBp < 10 MAT, which is well below the values prescribed for both the
Ignitor and the CIT designs in order to have n 0 r E > 2.5 X 1014 s • cm " 3 . For

IAEA-CN-47/H-I-1

257

example, I P B P can be as high as 45 MA-T in the Ignitor design, for qy «=< 2.8.
My question is what new plasma regimes can this machine produce that JET
cannot explore?
F. ENGELMANN: On the basis of present knowledge, a definitive
quantitative prediction of the confinement capability of reactor grade devices
is not yet possible. However, NET has a larger plasma current I p and a larger
magnetic field B than JET. Applying NET physics assumptions, the confinement
capability scales as IpBR/a, which for the NET reference options is larger by afactor of 15 than for JET when operated with a separatrix bounded plasma.
While this opens up reasonable prospects for reaching ignition in NET, it is infact not impossible that more precise knowledge in the future of the actual
operational limits — in particular, the plasma density and possibly the plasma
temperature - and of plasma confinement might require a modification of theparameters of NET. Such an adjustment is possible within the framework of the
NET time schedule, as presently envisaged.
B.G. LOGAN: Can you use lower hybrid current drive in the full steadystate NET option, or will you need other current drive methods?
F. ENGELMANN: I would definitely not rule out the possibility of using
lower hybrid current drive. In fact, in present day experiments the observed
current profiles appear to be centrally peaked, in contrast to the results ofmodel calculations. However, other current drive methods may well be needed
to provide the desired current profile.

IAEA-CN-47/H-I-2

A COMPACT IGNITION EXPERIMENT*J. SCHMIDT, G. BATEMAN, C. BUSHNELL,
J. CITROLO, P. COLESTOCK, D. IGNAT,
R. IZZO, S. JARDIN, L.P. KU,
G. KUO-PETRAVIC, M. PETRAVIC, C.K. PHILLIPS,
N. POMPHREY, D.E. POST, R.B. WHITE, K.M. YOUNG
Princeton Plasma Physics Laboratory,
Princeton University,
Princeton, New Jersey
D. BLACKFIELD, T. BROWN, C. FLANAGAN,
Y.-K.M. PENG, T. SHANNON, D. STRICKLER
Fusion Engineering Design Center,
Oak Ridge, Tennessee
L. BROMBERG, D. COHN, M. GREENWALD,
B. LIPSCHULTZ, D.B. MONTGOMERY, R. PARKER,
D. SIGMAR, R. THOME, S. WOLFE
Massachusetts Institute of Technology,
Cambridge, Massachusetts
W. HOULBERG, J. SHEFFIELD, D. SWAIN, N. UCKANOak Ridge National Laboratory,
Oak Ridge, Tennessee
M. HUGHES, C. PAULSON, M. PHILLIPS, A.M.M. TODD
Grummann Aerospace Corporation,
Princeton, New Jersey
G. KERBEL
National Magnetic Fusion Energy Computer Center,
Livermore, California
D.G. SWANSON
Auburn University,
Auburn, Alabama
F. PUHNGA Technologies Inc.,San Diego, California
United States of America* Work supported by the United States Department of Energy,
under Contract No. DE-AC02-76/CHO-3073.

259

260

SCHMIDT et al.

Abstract
A COMPACT IGNITION EXPERIMENT.
A national design team, with representation from essentially all of the major US fusion laboratories, has developed the conceptual design for a compact tokamak device to build on the good resultsfrom existing tokamak experiments. This device is called the Compact Ignition Tokamak (CIT), andhas the mission of 'realizing, studying, and optimizing ignited plasma operation'. The features of thisare compact size, liquid nitrogen precooled copper coils producing about 10 T toroidal field and about
10 MA plasma current at plasma elongations of two, divertor and limiter operation, and ICRF heating.

Relatively small, high field tokamaks as represented by the
Alcator devices have been very successful in obtaining excellent
plasma parameters with relatively low cost. For some years, it
has been advocated that a high density, high field approach would
be an effective means for studying the physics of an ignited
plasma^' >2'. During the past two years, an extensive study of thephysics and engineering issues of this class of devices has been
conducted in order to evaluate their feasibility. Based on thiseffort, we have concluded that a tokamak, the Compact Ignition
Tokamak. (CIT), with the following parameters, would have theperformance necessary to study the physics of burning plasmast
Units

Limiter

Divertor

Major Radiusm
Minor Plasma Radiusm
Plasma Elongation
MA
Plasma Current
MW/m2
Peak Divertor Plate Heat Flux
T
Toroidal Fields
Toroidal Field Flat-Tops
Plasma Burn Time
Plasma Current Ramp-up Times
Neutron Wall Loading at 300 HW
Fusion Power
MW/m2
Toroidal Field Energy Requirement
GJ
Poloidal Field Energy Requirement
GJ
Combined Peak Power for TF 6 PF Coils MVA
ICRF Initial InstallationMW
ICRF Future Capability
MW
Number of Full Field Pulses
Number of 70% Field Pulses

1 .324
0 .4272 .0
10

1 .339
0.412
2.1
9
5.0
10.27

10.39
4.2
3.6
5.8

7.5
2.8
1 .7
650
10
20
3000
50000

The primary physics requirements for CIT are that it achieve
ignition with a burn time of about ten times the energy
confinement time for at least three thousand full powered pulses,
and be operable with both a limiter and divertor. Achievement ofsmall radius is facilitated by the requirements for pulse lengthand number of full-power pulses, which have been relaxed relative

261

IAEA-CN-47/H-I-2
Hydraulic System

Close-In
Shieldln-Vessel Remote
Maintenance System

Preload Structure

ThermalInsulation

ICRH
SystemFIG. 1. Isometric of CIT device in its compact cylindrical test cell. Included are the 'igloo' shield andthe remote handling mechanism for the hardware internal to the vacuum vessel.

to previously studied fusion engineering test facilities.
The
relatively
short TF
flat-top(12 times estimatedenergyconfinement time) can be accommodated by a magnet design in which
temperatures rises adiabatically during the pulse. The minimum
number of full-power burn pulses has been set to 3.000, but theexperiment is capable of operating 50,000 pulses at 70% of thefull field and current, limited in a practical sense by theoperating span of the machine, which is less than ten years. In
order to achieve the minimum major radius, it has been necessary
to utilize liquid nitrogen precooling of the magnet in a manner
similar to the Alcator-C and the Frascati FT (high-field compact)
tokamaks now in operation.
The need to recool between pulses,
coupled with the short TF flat-top pulse, results in two
operational changes relative to the majority of present-day
tokamaks; namely, a restriction of one to two full parameter
pulsea per hour and the requirement to ramp up the plasma current
concurrently with the toroidal field.
Figure 1 shows the CIT
device inside its cylindrical test cell.
The CIT mechanical configuration is based on the following
structural approaches, chosen from a number of alternatives
studied during the first phase of the conceptual design; (1) useof an external frame and hydraulic press to apply sufficient
preload to the TF coil to support the vertical separating force on
the TF magnet and thereby minimize the stress in the inner leg of
the TF coil; (2) use of wedging in the inner TF legs to support
the inward load on the TF coils; (3) us« of a gap between the TF
coil inner legs and the central ohmic heating (OH) coil, to assure

262

SCHMIDT et al.

structural independence of the two systems; (4) use of coppersteel laminated conductors in the TF and OH coil in order to
optimize the conductivity/strength ratio of the conductors; (5)use of an average stress criterion where the average TF stress isallowed to approach 0.85 times of yield strength of the composite
material, in accordance with the guidelines for allowable stressdeveloped for externally supported structures having to performthrough a limited number of cycles; (6) use of partial coil caseswith significant horizontal access between coils to provide foroverall support against out-of-plane loads; and (7) the poloidal
field (PF) coil system i3 external to the TF.
A physics analysis based on existing experimental data has
been used to develop the primary physics parameters for the device
listed in the table.
Particularly important elements of theexperimental physics basis used in the design are the TroyonGruber1- ' beta
scaling and the Murakami' * density limit
scaling. A low edge q is specified (2.8) at a level consistent
with tokaraak experiments to maximize plasma current. The plasma
current is further increased by a relatively large elongation,
limited to c=b/a of about 2 by considerations of verticalstability and operational experience on tokamaks.
A poloidal
divertor and a pellet injection system have been made part of the
CIT design to provide a margin for ignition, and to provide
impurity and particle control.
The confinement for CIT has been estimated using simple zerodimensional and sophisticated one and one-half dimensionaltransport calculations with a wide variety of empirical andtheoretical transport scaling formulas ' .
One of the major
uncertainties of these scalings is the dependence of the energy
confinement time on the heating power.
To assure maximum
conservatism in the machine design requirements, it is assumed
that the full alpha heating power is used when the scaling
includes a negative dependence on heating power. These estimatesindicate that ignition is easily achieved using scalings for the
H-mode (Figure 2 ) . H-mode scaling should be applicable since theCIT design contains a poloidal divertor
to allow H-mode
operation. The ignition margin ranges from acceptable to marginalfor the commonly accepted L-mode scalings.
Peaking the profile can increase the ignition margin for agiven global energy confinement time, or allow ignition for alower energy confinement time compared to relatively flat
profiles. A significant problem with peaking the profiles is thatthe q = 1 surface is located at r ~ .6-.7 a, due to the low q at
the plasma edge (~ 2.8). Thus, sawtooth oscillations would be
expected to flatten the central portion of the density andtemperature profiles.
Time dependent transport calculations
indicate that a combination of off-axis ICRF heating, a carefully
tailored current ramp, and pellet fueling can produce a peaked
density profile sufficient to allow ignition with even very
pessimistic L-mode scalings.

263

IAEA-CN-47/H-I-2
10

-l

1

1

r

1

1

1

1

1—

CIT H-mode Power Contours
2 05(l^otl-X )

0

4

8

12
<T>(keV)

FIG. 2. Contours of constant auxiliary heating power required for steady state operation plotted on agrid of average plasma density and temperature. The two zero power curves are the Ohmic heating
equilibrium, and the ignition contour. The operating region is likely to be bounded by a maximum densityand beta shown here and described in the text.

The auxiliary heating system Is designed to provide the extra
heating above the ohmic heating to reach the plasma temperatures
required for ignition.
The heating systems are to be staged.
Initially, 10 MW of ICRF is to be provided.
Should that proveinsufficient for ignition, another 10 MW of either ICRF or BCRH
will be added.
Fast wave ICRF has been chosen for the initialsystem because it has demonstrated efficient ion heating at highpowers and can be implemented in a high density tokamak pla3ma
using rf sources currently available at reasonable cost.
The frequency range has been chosen to be 80-110 MHz. The
nominalfrequency is 90 MHz for full field operation withminority "*He and second harmonic tritium heating which is thereference heating method. Lower field operation will be possible
with minority *H and second harmonic deuterium heating.
Thesingle-pass absorptivity of the two primary modes has been
calculated from a one-dimensional mode conversion model1
to beadequate for efficient plasma heating. The calculations indicatethat He absorption dominates second harmonic tritium when He ispresent, but even when He is absent the residual second harmonictritium absorption is adequately strong over a broad range ofparallel wave numbers.

264

SCHMIDT et al.

Wave propagation studies show that the power will beadequately focussed, and Fokker-Planck studies show that there
will be relatively little or no high energy "tail" formation due
to the high plasma density. The antenna design studies show thatthe double resonant loop antennas recessed in cavities in thevacuum vessel should be capable of adequate coupling to the plasma
at the power densities required for heating the plasma.
Detailed analyses show that operation with an elongation in
the range between 2.0 and 2.2 will allow stable operation of a 9
MA diverted plasma. The vertical instability, which has a vacuum
vessel stabilized growth rate of about 15 msec, is controlled by
two radial field coils located inside the TF. The plasma will
evolve from a circular shape at 500 kA to a diverted shape at fullcurrent.
Analyses using optimized pressure and current profiles with
the PEST code indicate for a triangularity of .5 and elongation of
2, the theoretical beta limit for free boundary, n - 1, externalkink modes in high-n, internal ballooning modes is above 10%.
This is well above the Troyon scaling value (~ 6%) and provides
some margin for practical operation of CIT.
Analyses indicatethat the peaked pressure profiles produced by strong central alphaparticle heating between the sawtooth oscillations are unstable toballooning modes when a flat current profile is produced by thestrong sawtooth oscillations.
However, flattening the pressure
profile inside the sawtooth mixing radius can produce a stable
plasma with high beta's.
The high fusion power of CIT in a compact design poses severe
impurity control problems due to the high peak heat fluxes and thelarge total energy to be handled. The problem is exacerbated by
the short radial decay length of .5-.6 cm expected for CIT
conditions.
The divertor plates and limiters are inclined
relative to the flux surfaces to minimize the heat flux, andfabricated using carbon tiles which are not actively cooled.
Studies have indicated that the nominal peak heat loads are of theorder of 5 MW/m .
The double null poloidal divertor was chosen to provide
access to "H-mode" confinement and to minimize sputtering of thedivertor plates or limiters by operating in the "high-recycling"
regime in which a cool, dense plasma is produced by intense,
localized recycling and radiative losses near the divertor
plates.
Detailed calculations indicate that the CIT divertor
design can localize the recycling to maximize the probability ofobtaining good confinement with the H-mode.

IAEA-CN-47/H-I-2

265

REFERENCES
[1]
[2]

[3]
[4]
(5]
[6]
[7]

COPPI, B., Comments Plasma Phys. Controll. Fusion 3 2 (1977) 47.
COHN, D.R., PARKER, R.R., JASSBY, D.L., Nucl. Fusion 16 (1976) 31;
COHN, D.R., WILLIAMS, J.E.C., BROMBERG, L., KREISCHER, K.,
MONTGOMERY, D.B., PARKER, R.L., in Fusion Reactor Design Concepts (Proc. 2nd IAEA
Meeting Fusion Reactor Design, Madison, 1977) 113.
TROYON, F., GRUBER, R., Phys. Lett., A 110 (1985) 29.
MURAKAMI, M., et al., Nucl. Fusion 16 (1976) 347.
UCKAN, N., SHEFFIELD, J., Simple Procedure for Establishing Ignition Conditions inTokamaks, ORNL-TM-9722, Oak Ridge National Lab., TN (1985).
HOULBERG, W., in Fusion Engineering (Proc. 1 lth Symp. Austin, TX, 1985), Vol. 1, IEEE,
New York (1985) 50.
COLESTOCK, P.L., KASHUBA, R., Nucl. Fusion 23 (1983) 763.

DISCUSSION
A. GIBSON: What outputs from the present experimental programmes do you
regard as essential prerequisites for funding approval of CIT?
J. SCHMIDT: Although we consider the present tokamak experimental
programmes as providing important information for developing the physics requirements for CIT, these results are not prerequisites for project funding. A physics
research and development plan has been prepared as a guide for the information thatwould be of most value to the CIT project.

IAEA-CN-47/H-I-3

TOKAMAK REACTOR OPERATION SCENARIO
BASED ON PLASMA HEATING
AND CURRENT DRIVE
BY NEGATIVE ION BASEDNEUTRAL BEAM INJECTOR
S. YAMAMOTO, K. OKANO1, S. NISHIO, M. SUGIHARA,
R. SAITO2, T. UEDE3, T. KOBAYASHI, N. FUJISAWA,
T. TONE, Y. OHARA, S. TANAKA, H. HORIIKE,
Y. OKUMURA, T. SHIBATA, S. MATSUDA, K. TANI,
T. HIRAYAMA, M. AZUMINaka Fusion Energy Research Establishment,
Japan Atomic Energy Research Institute,Naka-machi, Naka-gun, Ibaraki-ken, Japan

Abstract
TOKAMAK REACTOR OPERATION SCENARIO BASED ON PLASMA HEATING AND
CURRENT DRIVE BY NEGATIVE ION BASED NEUTRAL BEAM INJECTOR.
The paper describes tokamak reactor operation scenarios based on plasma heating andcurrent drive by negative ion based neutral beam injectors (NBI). Applying the NBI systemtechnology to a next generation tokamak (e.g. FER in Japan) represents a reasonable step
originating from the present database. The overall heating and current drive efficiency ofthe system is very high. The beam driven current profile is easily controlled by the system.
The impact of the NBI system on reactor structural configuration and maintenance isfavourable. In next generation tokamaks, quasi-steady state operation, which reduces cyclic
thermal and mechanical loads, is beneficial as compared to conventional pulsed operation.
In FER, it is also possible to demonstrate steady state operation with an energy multiplication factor Q of about 5. Moreover, steady state operation in power reactors with Q valuesof more than 20 may be possible if the expected future progress in the system materializes.

1.

INTRODUCTION

The positive ion based neutral beam injection technology is well established [ 1 ]and plays an important role in high power plasma heating of JT-60 [2]. Recent
experiments on beam driven current in large tokamaks [3,4] increase thepotentiality in applying the NBI system to next generation tokamaks. The
development of a high current negative ion source is essential for next generation
1

Toshiba Corp., Japan.
Mitsubishi Electric Corp., Japan.
3
Hitachi Ltd., Japan.

2

267

268

YAMAMOTO et al.

Plasma centre
Drift duct
Flexible joint
Neutron shutter
Gate valve
Calorimeter
Ion dump
Cryo-pump
Neutron shield

10
11
12
13
14
15
16
17
18

Magnetic shield
Neutralizer
Beam monitor
Retractable target
Accelerator
D" source
Cryo-pump
Beam steering mechanism
Beam profile controller( b )FIG. 1. (a) Top view of FER and 500 keVnegative ion based neutral beam injection system;(b) side view of NBI system.

neutral beam injectors, in order to achieve an acceptable efficiency at beamenergies in excess of 200 keV. A new concept of a negative ion based NBI systemwith a beam energy E^ of 500 keV and an input power Pb of 20 MW is proposed forheating and current drive in FER [5]. Applying the NBI system technology to
FER represents a reasonable step proceeding from the present database [5]. The
present paper describes operation scenarios based on the NBI system for next
generation tokamaks and power reactors.
2.

A 500 keV/20 MW NEGATIVE ION BASED NEUTRAL BEAMINJECTION SYSTEMAt JAERI, the development of H ion sources started in 1983. So far,an H" ion beam of 1.26 A/21 keV/0.2 s has been extracted at a current densityof 9.5 mA'cm" 2 through a. MX 26 cm2 grid system [6]. In parallel to thedevelopment of the ion source, a negative ion based NBI system is proposed.

IAEA-CN-47/H-I-3

269

The D" ions are produced by a volume production method and extracted by
DC acceleration. During the beam injection phase, the beam energy is variablein the 200 keV to 500 keV range. The top and side views of FER and the
NBI system are shown in Fig. 1. This injector is characterized by a long, slender
neutralizer, which is 0.3 m wide and 30 m long. The required beam divergenceis 0.3° or 5 mrad. Judging from recent experimental results, such a low beam
divergence can be realized by using a volume D" ion source. A divergence of
0.8-0.9° was obtained in the present experiment, in spite of the low beamenergy of 18 keV. A divergence of 0.18° has already been obtained in a helium
injector, with a beam energy of 200 keV and a beam current of 3.5 A. The
thickness of the neutron shield for the NBI system has been estimated; thiscorresponds to a dose value of less than 2.5 mrem-h" 1 , one day after shutdown
and following a two years' period of full power operation. The long, slender
neutralizer concept brings about a considerable reduction in the neutron damage of theion sources and in the required pumping speed. The total pumping speed is950m 3 -s" 1
for 20 MW injection, while that for the JT-60 NBI, for example, is 1400 m^s" 1
for 1.4 MW injection. Another feature is that the poloidal magnetic field fromthe tokamak is used to deflect the unneutralized ions and the positive ions
towards the ion dump near the injection port. Therefore, no bending magnet,
which would occupy a large amount of space, is used in this design. If these
design concepts are adopted, the NBI system becomes more reliable, has a much
longer lifetime and is easier to maintain. Although the overall efficiency of thesystem is 0.45, this value can be further improved by employing a plasma
neutralizer and/or an energy converter system.

3.

CALCULATION MODELS

To analyse beam driven current profiles and beam driven current efficiency,
the authors have developed an Orbit Following Monte Carlo (OFMC) code, whichcalculates beam deposition and subsequent collisional evolution of the fast ion
trajectories in an axisymmetric tokamak with elliptical cross-section [7]. The
OFMC code is combined with a transport code, an equilibrium code and aballooning stability code to investigate the effect of beam driven current and its
power input on the beta value. The authors have also developed a current drive
code with the analytic solution of the 2-D Fokker-Planck equation, which yields
the current profile in a non-circular (ellipse, Dee, etc.) geometry [8,9]. This codeis calibrated with the OFMC code and mainly used for parametric survey.

4.

CURRENT PROFILE CONTROL

Current profile control is important for beta enhancement [10], control ofsawtooth oscillations and suppression of major disruptions. In the NBI system,

270

YAMAMOTO et al.
ZIm]

Device porameter
E b =500keV, R l a n g =4.6m
B T =5.3T
Plasma parameters
R0 = 5.2m , a = I 12 m

-0.5
1

•

1

C

4 -Ic)
m
sy
*^
/s
im

i-A
\^

l\

3 \
2

T (+/+„) = '(0K1-+/V"
!

\

-"S

-1.5
h(Z)

/A

1
n
0

K" = 1.5 , Z eff = 2.2-1(d)0.2 0.4 0.6 0.8
it / it

1.0

lb[AJ/Pb[WJ 0.097

fsMO'2)

~0

0.10
~ 0.3

0.10
~ 2

FIG. 2. Current profile control: (a) magnetic surface; (b) beam power profile h(z); (cj beam
driven current profile J^Bl and poloidal flux function ty; (d) beam shinethrough fractions,fs, and global beam driven current efficiency, I\,[^V^b[ W]. Device and plasma parameters
are also shown.

almost no beam is neutralized at the neutralizer entrance, because of the low
system pressure. Hence, by changing the beam power profiles with horizontal
magnetic fields less than 0.03 T, produced by coils in the beam profile controller
shown in Fig. 1, the beam driven current profile is easy to control. Figure 2
shows typical beam driven current profiles Jf BI corresponding to three beam
power profiles along a vertical direction at R tan g, which is the minimum major
radius along the beam path. The global beam driven current efficiency,
Ib[A]/Pp[W], and the beam shinethrough fraction, fs, are also shown. Figure 2
only shows the capability of current profile control because the target plasma
parameters are fixed. The upper and lower beam profiles with respect to theequatorial plane can be controlled independently. The required total beam
driven current and its desirable profile, depending on the scenario, can beobtained by combination of the individual current profiles as obtained in a few
units of the NBI system. The drift duct is 0.44 m wide and 1.2 m high at thereactor entrance and 0.5 m wide and 2 m high at the first wall, respectively.
The drift duct does not seriously affect the reactor structure such as toroidal
field coils, shield structure, passive shell conductors, coolant channels andelectrical insulators for one-turn resistance.

271

IAEA-CN-47/H-I-3(a)|Burn|

Major Radius
Minor RadiusElongation
Field on Axis!
n. ( 1 0 % ) 1.16
Electron Temperature
T. ikeV) 10
Ion Temperature
T, (keV) 10
Ohmic Current
5.9
low (MA)
0.05
Loop Voltage
VL (V)
Beam Current
Ib (MA)
3.98
Toroidol Beto [thermal) ft ml*)
Toroidal Beta (beam)
0i b 1%)
Poloidal Beta (thermal) /3pm
1.76

Electron Density

Poloidal Beta (beam)ft.b

Effective Ionic Charge

Zrt I

Fusion Power

Pi (MW)
Pa (MW)
Pb (MW)f.

a-heatingPower

Beam Input Power
Shine-through fraction

Recharge)

5.2
1.12
1.5
53

Ro (m)°P (m)K
8 T (T)

0.1

73

05

8.0
-1.6

-Q6

1.5

10

70
nip

I
n, [xto'9rri3]Inu i o Lpt total

1.98 ' 2.0Q053)j3 P total

0.579J 063
1.5
297
60

13.7 (X e = 1 % j
He=5£40.9 (32)0.2

FIG. 3. Quasi-steady state operation in next generation tokamak. Beam energy, £b, is
500 keV: (a) plasma parameters during burn phase and recharge phase; (bj plasma density
n e dependence of electron temperature, Te, beam driven current, /(,, and fs. Beam input
power, Pb, is 40.9 MW; (cj Pb dependence ofTe, / b and fs. Plasma density ne = 1019 m'3.

5.

QUASI-STEADY STATE OPERATION REDUCING CYCLIC THERMALAND MECHANICAL LOADSIn FER, a quasi-steady state operation scenario was proposed, based on
plasma heating and current drive by the RF system [11, 12]; various problems
were discussed [13]. In this operation, the plasma current is non-inductively
ramped up by LHRF. In the ignition approach phase, LHRF or ICRF is used
for plasma heating. In the burning phase, the current is inductively maintainedby the OH coils, and the plasma profile control is performed by ECRF. Afterthe burning, the current is ramped down to the former value and then noninductively maintained by LHRF in a low density plasma. During this phase, thetransformer coils are recharged. The engineering advantages due to introducing
RF assisted quasi-steady state operation were discussed [14] and can be summarized as follows: (i) reduction in machine size without decreasing the burntime, and (ii) remarkable reduction in cyclic stress due to the overturning force on
the TF coil. The same engineering advantages are expected in the quasi-steadystate operation scenario based on the NBI system. In this operation scenario,
the NBI system only is used for heating, current drive and current profile control
during all phases except for the plasma initiation phase by ECRF, with about

272

YAMAMOTO et al.

3 MW power and a current rise phase up to 2 MA as assisted by OH coils.
Figure 3(a) shows typical plasma parameters during the burn and recharge phases.
A beam input power Pb of 40.9 MW with an Eb of 500 keV is injected. Xenon
impurity is used to improve the current drive efficiency and to decrease therecharge time. By fully using the prepared heating power for theignition approach, a relatively high density plasma can be used as a target
plasma. Thus, cold and dense divertor plasma can be obtained. The plasma
poloidal beta, 0 p , during the recharge phase is about one third of that of theburn phase. If the plasma current is reduced from 5.9 MA to 5.4 MA, the time
variation of the overturning force on the TF coils can be reduced significantly [14].
The heat load on the divertor plate during the recharge phase is about halfthat of the burn phase, which leads to a reduction in the cyclic heat load on thedivertor plate as compared to pulsed operation. The shinethrough fraction, fs,
is 0.2. To reduce the heat load due to shinethrough beam particles, a parametric survey was performed. Figure 3(b) shows the n e dependence of beam
driven current It,, electron temperature Te and fs. The injection power Pb
is fixed at 40.9 MW. Figure 3(c) shows the P b dependence of I b , Teand fs. The plasma density is fixed at 1 X 1019 m~3. Figures 3(b) and (c)suggest that 2 X 1019 m~3 is a preferable value of the target plasma density,from an engineering viewpoint, even if a higher injection power is necessary.
6.

STEADY STATE OPERATION

6.1. Steady state operation for next generation tokamaks
We now consider the steady state operation scenario for FER, using theheating power prepared for the ignition approach. Figure 4(a) shows anequi-contour map for minimum heating power, Pb [MW]. The plasma temperature and density profiles are assumed to be T(r) = T(0)[l + (r/a) 2 ] 0 - 6
and n(r) = n(0)[ 1 + (r/a) 2 ] 0 - 5 , respectively. The beam energy E b is 500 keV.
Figure 4(b) shows the relation between the Q-values and the other plasma
parameters, corresponding to the contour denoted by the dash-dottedline in Fig. 4(a). The Q-value includes the fusion output due to beam-target
interaction. The dotted line in Fig. 4(b) is the contour corresponding to the
Troyon beta scaling, expressed as 0 cr = CTl[MA]/a[m]Bj[T], where I, a andBT are plasma current, plasma radius and toroidal field strength, respectively.
In this case, Or, the coefficient of the Troyon beta scaling, is assumed to be 5.
This value is to be expected when the current profile is controlled [10]. If thetotal beta value, /?T, is restricted by this Troyon scaling, the attainable Q-value
in FER is about 5. The pressure due to suprathermal alpha particles and ion
beam particles amounts to 40 to 50% of the total pressure. Figure 4(c) shows
an example of self-consistent plasma profiles, when the beta value is enhanced
by current profile control.

273

IAEA-CN-47/H-I-3
11

(a!

«

„ 9

20.

\

«w
so eo
/
B 10 12 14 16 18 20
T, [keV]
\

•="3

6

(C)

•o-

30

30

I

•io.

S50
5 40
|-?30
^20

10

4

//IT

2

S

a>
£=

0
*E 2

3

3

2

cr

>1

•r °

3
lc?

12 14

16
T

18 20 22

<

3

*

2

i".

'

"

0

4.0

[keV]

5.0

6.0

R Cm]

FIG. 4. Steady state operation in next generation tokamak. Beam energy, Eb, is 500feeV:(a) equi-contour map for minimum heating power P[MW] in density and temperature plane,calculated from point model of power balance with INTOR-Alcator scaling law for electronenergy confinement. Xenon impurity fraction of 0.04% and helium impurity fraction of 5%in the ion density are assumed; (b) maximum attainable Q-value. Parametric survey is
performed along contour denoted by dash-and-dotted line in (a). Critical beta value, )3cr, isassumed to follow Troyon beta scaling law, expressed as /3cr = CjI[MA]/a[m]BT [T](scaling coefficient Cj 5). (3T.P total toroidal beta value, including beta values due to thermal
populations, /?th, to suprathermal alpha particles, j3a, and to ion beam particles, ft,;(c) example of self-consistent plasma profiles, p-p.1 total pressure, pt^: thermal pressure,q: safety factor, JQ: toroidal plasma current component.

6.2. Steady state operation of power reactors
Let us, next, consider the requirements for steady state operation in powerreactors (R o = 6.75 m, a = 2 m, K = 1.6) [15]. This machine scale is pertinent
for a net electrical output of about 1000 MW(e), within reasonable beta limits.
The beam energy, E b , is assumed to be 1 MeV, which seems to be reasonable
after the 500 keV injection system will have been completed for the next
generation tokamak. Where should the operation point be fixed? The beam
driven current efficiency is good in high temperature and low density plasmas,
for any given beta. High plasma temperature (about 35 keV) ensures maximum
Q-values, but the fusion output is rather small, within certain beta limits. On
the other hand, low temperature (about 10 keV) ensures maximum fusion

YAMAMOTO et al.

274

Device parameters(a)BT = 5 . 2 T ( B m a , =ond B r = 6 . 9 T ( B m o , = l 6 T )Plosmoporamelers

R0 = 6.75m , a = 2 m , K = 1.6 , Z e n
ne=l.lxl020m"3,

ne(0)[l-(r/a)2]a"Te[r)=

Te(O)[l-(r/a)2)ar

(b)
On

1

1

1

1

1

\

J3 9
flft
0. II*

~o 3
\

0

I
0.8
0.6
0.4

"T "I*
7 8 9 10 11 12 13 14 15 (Bmax = l2T)
6

7

8
CT

9 10

T=25keVne(r)=P<
Pb
aT CMW)(MW) [%]

00.3 .0 3 0 0 0 173 6.8 17.4
1.0 .0 4 0 6 0 192 6.7 21.1
1.5 .0 4 7 6 5 160 6.6 29.7
0.3 c).5 3095 160 6.7 19.3
Te(r)0.3 .0 3000 173 6.8 17.4
dep.
0.3 .5 2899 187 6.9 12.3
0.5 ().5 3345 163 6.6 20.5
nelrlJetr
1.0 .0 4 0 6 0 192 6.7 21.1
dep.
1.5 .5 4821 230 6.7 2i.a
ne(r)
dep.

(Bm
- 12MA

FIG. 5. Steady state operation in power reactor (Ro = 6.75 m, a = 2 m, K = 1.6). Beamenergy, Eb, is 1 MeV. Plasma density and plasma temperature profiles are assumed to ben(r) = n(0) [1 + fr/aJ2]an and T(r) = T(0) [1 + fr/a)2]aT. (a) CT dependence ofQ-values
and other parameters. Plasma temperature and plasma density are fixed to obtain constantfusion output. Maximum field strength on TF coil, 5 m a x , is 12 and 16 T. Pressure due tosuprathermal alpha and ion beam particles amounts to 30 to 40% of total pressure;(bj influence of density and/or temperature profiles on Q-values. Total beta value is fixedat almost identical value.

output. In this case, however, most of the fusion output is spent for the NBI
system. In the present situation, the net electrical output seems to be an
appropriate indicator in evaluating the performance of a power plant, because of
the complexity and/or the uncertainty in evaluating the busbar cost [8]. An
electron temperature, T e , of about 20 keV ensures maximum net electrical
output. Figure 5(a) shows the Op dependence of the Q-value and other parameters.
The plasma density and temperature profiles are assumed to be n(r) = n(0)[l+(r/a) 2 ] an
and T(r) = T(0)[l + (r/a) 2 fT ) respectively. In Fig. 5 (a), a n = 0.3 and a T = 1
are adopted. The influence of the density and/or temperature profiles on the
Q-values is shown in Fig. 5(b). A broad density profile gives a relatively lower
Q-value than a peaked density profile, when the total beta value is fixed at almost
the same value. In Fig. 5(a), the beam driven current can be decreased in accordancewith the enhancement of the critical beta value, under conditions of constant

IAEA-CN-47/H-I-3

275

plasma temperature and density, i.e. constant fusion output. This results in a
reduction of the beam input power, P^ and, hence, in an increase of the Q-value.
It is expected that Cj for tokamak plasmas with conventional cross-sections is
improved more than twice as compared to the normal case [10], since the current
profile can be controlled by the NBI system. In addition, a maximum fieldstrength on the TF coil, B m a x , of 16T may be possible with the expected future
progress [ 16]. As a result, a Q-value of more than 20 will be realized, and a net
electrical output of more than 700 MW(e) is expected, with an overall efficiencyof 0.5. In Fig. 5(a), toroidal and poloidal beta values are also shown. The
pressure due to suprathermal alpha particles and ion beam particles amounts to
30 to 40% of the total pressure, which may improve MHD stability and raise
the critical beta [17]. In addition, if the bootstrap current can be considered
to be a realistic aid in the current drive of a steady state tokamak [4] therequired power for current drive will be reduced.
7.

CONCLUDING REMARKSThe operation scenario, based on plasma heating and current drive bynegative ion based neutral beam injectors, has been proposed as one of theoperation scenarios for FER. Applying the NBI system technology with a beamenergy of 500 keV to FER represents a reasonable step originating from thepresent database. The system is used for plasma heating, current drive andcurrent profile control. The quasi-steady state operation by the NBI system
realizes nearly steady state condition for the reactor structural components.
Steady state operation can also be demonstrated by using an injection powerof 40 MW, prepared for ignition approach. Q-values of about 5 can be expectedin this steady state operation if the Troyon beta scaling coefficient is increased
from 3.5 to 5. According to the numerical results, this beta enhancement canbe realized when the beam driven current profile is controlled. The control of
the current profile is easily performed by the system. The NBI system favourably
affects the reactor structural configuration and maintenance. The steady state
operation scenario in a power reactor is also considered. The beam energy
for the NBI system in the power reactor is assumed to be 1 MeV. This value
seems to be reasonable, after the 500 keV injection system will be completed
for FER. The critical beta value for tokamak plasmas with conventional crosssections is expected to be improved by more than twice, as compared to thenormal case, when beam driven current profile is controlled by the NBI system.
In the steady state operation of a power reactor with fixed fusion output, thecurrent required for equilibrium and stability can be reduced when the criticalbeta value is increased. This results in a reduction in the required beam powerand, hence, in an enhancement of the Q-value. In addition, if the maximum
field strength on the TF coil is increased from 12 to 16 T, a Q-value of morethan 20 is realized, within a reasonable beta value.

276

YAMAMOTO et al.
ACKNOWLEDGEMENTS

The authors are indebted to Dr. Y. Shimomura for many stimulating
discussions. Thanks are also due to Drs Y. Sawada, H. Shirakata, S. Tamura,T. Iijima, M. Tanaka, M. Yoshikawa, K. Tomabechi, Y. Iso, and S. Mori forsupport and encouragement.

REFERENCES[1] OKUMURA, Y., et al., in Plasma Physics and Controlled Nuclear Fusion Research 1984
(Proc. 10th Int. Conf. London, 1984), Vol. 3, IAEA, Vienna (1985) 329.
[2] YOSHIKAWA, M, et al., these Proceedings, Vol. 1, p. 11.
[3] REBUT, P.H., et al., these Proceedings, Vol. 1, p. 31.
[4] HAWRYLUK, R.J., et al., these Proceedings, Vol. 1, p. 51.
[5] OKUMURA, Y., et al., in Fusion Engineering (Proc. 11 th Symp. Austin, Texas, 1985);
TONE, T., FER Design Team, in Fusion Reactor Design and Technology (IAEA Tech.
Comm. Meeting and Workshop Yalta, 1986); more detailed descriptions (in Japanese)are given in JAERI-M 86-064 (1986) by HORIIKE, H., et al.
[6] OKUMURA, Y., et al., Production and Neutralization of Negative Ions and Beams(Proc. 4th Int. Symp. Brookhaven, 1986).
[7] TANI, K., AZUMI, M., HIRAYAMA, T., et al., to be published.
[8] MIKKELSEN, D.R., Nucl. Technol./Fusion 4 (1983) 237.
[9] OKANO, K., et al., Tokamak Concept Innovations, IAEA-TECDOC-373, p.209.
[10] See, e.g. the following papers: SEKI, S., et al., to be published in Nucl. Fusion,
YAMAZAKI, K., et al., these Proceedings, Vol. 2.
[11] TONE, T., et al., in Plasma Physics and Controlled Nuclear Fusion Research 1984
(Proc. 10th Int. Conf. London, 1984), Vol. 3, IAEA, Vienna (1985) 287.
[12] SUGIHARA, M., et al., Nucl. Eng. Des./Fusion 1 (1984) 265.
[13] FUJISAWA, N., et al., Tokamak Start-up, Physical Science, Vol. 26, Plenum Press,
New York and London (1986).
[14] NISHIO, S., et al., Nucl. Eng. Des./Fusion 3 (1985) 59; more detailed descriptions are
given in JAERI-M 85-141 (1985)[ 15] TONE, T., et al., in Fusion Technology (Proc. 6th Top. Meeting Technology of Fusion
Energy, San Francisco, 1985), Vol. 8, No. 1, Part 2(A) (1985) 731.
[16] See, e.g. the following paper: SHIMAMOTO, S., et al., Tokamak Concept Innovations,
IAEA-TECDOC-373, IAEA, Vienna (1986) 243.
[17] TSANG, K.T., et al., Nucl. Fusion 21 (1981) 1227.

DISCUSSION
R.J. GOLDSTON: Perhaps you could describe for us the status of thenegative ion based neutral beam programme at JAERI?
S. YAMAMOTO: As I mentioned, up to now we have extracted an
H" ion beam of 1.26A with a current density of 9.5 mA-cm~2 at 21 keV,lasting 0.2 s. A test stand with a beam energy of 75 keV and beam current of

IAEA-CN-47/H-I-3

277

5A has been completed at JAERI. We plan to construct a test stand with abeam energy of 500 keV and beam current of 25A for the FER tokamak.
R.W. CONN: The beam line system, I must say, begins to make thetokamak look like an inertial confinement device. My questions are, first,
whether you have determined the cost of this NBI system, in yen per wattdelivered to the plasma, including the cost of the long neutralizer tube andthe shielding, and second, whether you can give the reasons why you prefernegative neutral beams to an ICRF system, which would be more compact andprobably less expensive?
S. YAMAMOTO: We have no precise cost data for the negative ion based
neutral beam system. The NBI scenario is presented as one of the operational
scenarios for FER. This NBI scenario is very attractive. Applying the NBI
system technology to FER represents a reasonable step forward from the present
database. However, to make the FER project more successful, work related toboth NBI and RF scenarios will be continued until the point where a decision
on FER construction can be made.
B. COPPI: The parameters of FER that you have given lead to valuesof IpBp, where I p is the plasma current and B p is the mean poloidal field, as I
mentioned in my comment on Dr. Engelmann's paper this morning (seepaper H-I-l in these Proceedings, this volume). This indicates a relatively low
confinement quality when compared to ignition experiments currently being
designed. What values of n^rg do you expect FER to produce?
S. YAMAMOTO: My presentation reports plasma parameters for the
beam driven steady state operation of FER. The optimum plasma parameters
for the steady state are different from those for a self-ignited plasma. In FER,a global energy confinement time is assumed to follow the INTOR-Alcator or
the Mirnov scaling law. However, the confinement properties for steady state
operation are less important than those for a self-ignited plasma.
B.G. LOGAN: On the basis of the TFTR results, you would obtain a much
improved Q if you considered bootstrap current amplification, even by a modest
factor, at /3p « 1.
S. YAMAMOTO: Thank you for your comment. I think it would be useful
to calculate the beam drive current plasma profile self-consistently, taking
account of the bootstrap current.

IAEA-CN-47/H-I-4

INTERNATIONAL LARGE COIL TASK:TESTING OF THE LARGEST
SUPERCONDUCTING TOROIDALMAGNET SYSTEM*
M.S. LUBELL, W.A. FIETZ, P.N. HAUBENREICH,
J.W. LUE, J.N. LUTON, S.S. SHEN
Oak Ridge National Laboratory
Oak Ridge, Tennessee,
United States of America
K. OKUNO
Japan Atomic Energy Research Institute,Tokyo, Japan
A.R. ULBRICHT
Kernforschungszentrum Karlsruhe GmbH,
Karlsruhe, Federal Republic of Germany
J.A. ZICHY
Schweizerisches Institut fur Nuklearforschung,
Villigen, Switzerland
Work performed at
Oak Ridge National Laboratory,
Oak Ridge, Tennessee,
United States of America

Abstract
INTERNATIONAL LARGE COIL TASK: TESTING OF THE LARGEST SUPERCONDUCTINGTOROIDAL MAGNET SYSTEM.
The Large Coil Task is an international collaboration of the United States, Euratom, Japan and
Switzerland to develop large superconducting magnets for fusion reactors. The first tests of all six coils
were begun in February 1986 when the 420 tonne test array was cooled to 4.2 K. Each of the heavily
instrumented coils was tested alone to full design current. Heat perturbations (recovery, simulating
nuclear heating, or current sharing temperature measurements) to investigate design limits were carriedout. All six coils have been tested in the full array to 100% design current, producing an 8.1 T maximum
field. All the coils have been energized simultaneously to 8 T midplane field, and they have performed
very well without quenching.

Research sponsored by the Office of Fusion Energy, United States Department of Energy,
under Contract No. DE-AC05-84OR21400, with Martin Marietta Energy Systems, Inc.

279

280
1.

LUBELL et al.
INTRODUCTION

The international Large Coil Task (LCT) is a collaborative effort of the United States, EURATOM, Japan, and Switzerland to develop large superconducting toroidal field (TF)
magnets for fusion reactors [1] . The U.S. provides the test
facility at Oak Ridge and three of the D-shaped 2.5 x 3.5 mbore coils; the other participants, one coil each. The LCT
project is unique in three aspects.
First, these are thelargest superconducting (SC) coils fabricated and tested todate for research and development of SC magnets. All otherlarge magnets have been fabricated for a well-defined use in aparticular facility. Second, six widely different designs aretested under similar conditions. Comparison of the operating
features between pool-boiling (PB) and forced-flow (FF) andamong the various performance features are inevitable and will
yield a wealth of information. Third, as the magnets themselves are the experiment, they are each heavily instrumented
with some 200-300 sensors and voltage taps and, consequently,
detailed information on thermal, electrical, and mechanicalperformance has been obtained; some for the first time on SC
magnets of any size.
The facility is also instrumented
extensively, and additionally useful information is obtainedon the operation of a large cryogenic system.
Shakedown
operation of the International Fusion Superconducting Magnet
Test Facility (IFSMTF) and preliminary tests of the first
three coils already delivered were accomplished in 1984 [2,3],
Late in 1985, the assembly of the full six-coil toroidal array
was completed (Fig. 1) and cooldown of the 420-tonne test
array began. The test program was initiated in February, andby June each of the six coils was successfully tested byitself to full design current and maximum field on the conductor of 6.4 T. Since then the full six-coil array has beenenergized and all six coils have been operated at full design
current and 8.1-T field. In the full array, the test coil isoperated to 100% current and 8-T field while it is sufficientfor the other five background background coils to be atcurrents in the range of 72 to 91% of their design current.
Experiments were performed on all of the coils with intentional heat perturbation to determine some aspect of stability(i.e., for either recovery tests, simulated nuclear heating,or current sharing).
It is gratifying to report that all
coils have been charged to 100% design current without quenching. This paper will summarize the significant findings to
date.

2.

TEST FACILITY

The IFSMTF comprises the toroidal array of test coils,
with supporting structure in an 11-m-dia. vacuum tank, andIAEA-CN-47/H-I-4

281

FIG. 1. IFSMTF vacuum tank with all six coils in place, October 1985.

ancillary electrical, cryogenic, vacuum, control, and dataacquisition systems. The test stand is depicted in Pig. 1.
Test coils are attached through upper and lower collars to thehexagonal bucking post and are clamped in torque rings attheir outer corners.
The "cold wall" consists of panels
cooled with liquid nitrogen, with surfaces covered by reflective blankets. During operation, the tank is maintained in
the range of 1 0 " s to 1 0 ~ 4 torr by two turbomolecular pumps anda diffusion pump.
The facility refrigerator, which is rated at 1.5 kW at
4.2 K, supplies liquid helium at 4.2 K to three coils and thetest stand structure while circulating up to 300 g/s of heliumat 1.5 MPa and 3.8 K through the FF coils.

282

LUBELL et al.

Each coil has its own power supply, capable of 12 V andcurrents of 16 or 25 kA.
The control system provides for
simultaneous, coordinated ramping up or down of the systemwith different current ratios for each coil.
The quench
detection system is designed to cancel large self- and mutualinductance and thereby detect very small resistive voltages.
If a significant normal zone voltage is detected, the automatic control system initiates a rapid discharge of all coils
through external dump resistors.
Experimental data are acquired through four front-end
microcomputers (LSI-11) connected to a central processing
computer (VAX 780), which in turn is connected to a cluster ofcomputers and disk drives for data storage and analysis.

3.

COOLDOWN

The cooldown took one month, and aside from facility
problems at the start, it proceeded smoothly at an overall
average rate of about 0.4 K/h. This rate includes a holding
period to switch over from refrigeration to liquefaction tobuild up a 10,000-L helium inventory and to change over thecooling below 100 K to the turboexpander instead of LN 2 -cooled
helium gas. The cooling rate was limited by the available
mass flow and the allowable AT across a coil [4]. The cooldown was monitored by temperature probes on the winding andcase and by calculating the average winding temperature from a
resistance measurement of a small (5-A) constant current
flowing through the coil.
No undue thermal stresses werecreated during cooldown, and no unusual thermal displacements
were noted.

4.

TEST COILS

Except for certain constraints on basic dimensions,weight, and performance, each participant had a wide latitudeof freedom in the design of their coil. The coils are about
half the size envisioned today for tokamak fusion reactors,
but conductors are full size, carrying currents from 10-18 kA
at the design field. The coils by EURATOM (KfK), Switzerland(SIN) and U.S.-Westinghouse (EU, CH, WH) are cooled by FF of
subcooled supercritical helium at 1.5 MPa. The coils by Japan(JAERI), U.S.-General Dynamics, and U.S.-General Electric/Oak
Ridge (JA, GD, GE) are cooled by PB helium at atmosphericpressure. The conductor for the WH coil is Nb 3 Sn; all theothers use NbTi. The GD coil is the only layer-wound coil,IAEA-CN-47/H-I-4

283

all others are pancake wound. Westinghouse used aluminum
plates for the structural support; all others used stainlesssteel case.

5.

INSTRUMENTATIONNo two coils contain all the same types or number of
sensors, but they all have voltage taps, thermometers, fieldprobes, compensation coils, strain gages, and resistanceheaters. Most have displacement transducers, pickup coils,
and acoustic emission sensors, and the FF coils have pressuretransducers. The WH coil has inductive heaters.
The Pt and carbon-glass resistance thermometers andthermocouples are used to monitor cooldown and temperaturerise on dumps. The thin-film carbon resistance thermometers
are used to measure small excursions of conductor temperaturein stability tests of the PB coils. Both Hall- and magneticresistance probes are used for field measurements. The pickup
coils will monitor the pulse fields to be superimposed lateron the coils. Compensation coils are used in the quenchdetection system (QDS) to buck out self- and mutual-inductive
voltages to provide resistive voltages which, if large enough,automatically trigger a dump protection system. Strain gages
on the conductor and case are used to measure both the tensileand compressive strains.
Resistive heaters are used forrecovery and simulated nuclear heating tests (GD, GE, JA) ,simulated nuclear heating (CH), and current-sharing temperature measurements (EU, WH) . Only the WH coil has inductive
heaters, which are used for recovery tests. The acoustic
emission sensors provide signals from both the winding packand structural deformation.
The moving coil displacementtransducers (MCDT) were used to measure overall change in thecoil bore dimensions during single-coil charging (solenoidal
forces), and the moving slug displacement transducers (MSDT)
are used to measure movement of the winding pack either to oraway from the structural support.

6.

TEST PROGRAM

The detailed test program was planned and agreed upon by
all the participants, including the responsibilities, organization, guidelines for arriving at joint decisions in difficult
situations, schedule, and testing steps. The tests performed
in 1986 were 100% design current tests of each coil by itself(B= 6.4 T ) ; low-current tests of partial and full array
foranecessary adjustment of the QDS and balancing of the power

284

LUBELL et al.

supplies for multicoil operation; and full-array tests to 100%
design current and 8-T field at the midplane (B=8.1-8.2 T ) . For both single-coil and full-array design point
tests, some form of heat perturbation tests (i.e., recovery,
nuclear heating, current-sharing temperature measurement) anddump tests were performed.
In addition, the FF coils had
design-point tests made with varying amounts of flow and as afunction of inlet temperatures which could be lowered from
4.2 K to 3.8 K. Persistent mode operation [i.e., putting ashort in parallel with the dump resistor to provide a slow
discharge (time constant in hours instead of seconds)] wasalso used selectively. In all cases the voltage transients on
initiating the persistent switch is of short enough duration
so as not to initiate any dump.
Subsequently, the power
supply could be matched to the coils to remove them frompersistent mode operation to obtain a controlled discharge.
In all cases, operation was smooth enough to avoid triggering
the sensitive QDS.

7.

DESIGN POINT TESTSIn discussing the various test results below only one
example will be given for each. Many detailed test results
were recently presented for each coil [5].
All six coils achieved their design current, albeit at a
maximum self-field of 6.4 T, without exhibiting any training.
Because of differences in design the 100% current level ranged
from 10,200 A to 17,760 A. The three PB coils and one FF (WH)are cryostable, and therefore the results are not wholly
unexpected; however, two of the FF coils are adiabaticallystabilized (CH and E U ) , but no difficulty such as training was
encountered with either of them. They are both very stablecoils, which is a reflection of the rugged conductor and rigid
winding. These are the largest noncryostabilized coils everbuilt and operated.
All six coils have been measured in full-array tests to100% design current at 8-T field on the midplane of thestraight leg of the "D". The maximum field in the corner
region was therefore about 8.2 T. These are the largest(2.5 x 3.5 m bore, 40-tonne weight, and stored energy of
100 MJ) superconducting magnets ever fabricated which produceda field of 8 T. Figure 2 is a graph of all coil currents andstored energy (E ) in two charges of the full array to 100%
design current in the GD coil. The five background coils only
needed to be charged to 77% design current in order to produce
8.1 T in the GD coil.

IAEA-CN-47/H-I-4

285- 8 / 2 0 / 8 6 19 < 0 5 ^ 0 0
GD (TEST COIL)^BACKGROUND COILS-TOTAL STORED ENERGY45
TIME (h)FIG. 2. Current and stored energy of all six coils plotted against time in the historic charging of the
GD coil to 100% design current and 8.1 T field.

8.

TOLERANCE TO PERTURBATIONS AND OPERATION LIMITSTo test the operating limits of the LCT coils and obtainsome measure of how stably they perform, their tolerance toheat perturbations was investigated during the design point
tests.
8.1

Recovery Tests

Recovery tests were performed on all three PB coils (GD,
GE, JA) using resistive heaters in intimate contact with theconductor over approximately a 5-m length (half turn).
To
investigate the stability of whole turns (10-m lengths) or
different grade of conductor and different regions of thewinding pack where bubble accumulation might be important,some number of heaters (11 in GE, 20 in GD, 24 in JA) werespread throughout the winding pack.
The experiment was
performed in the following manner. The coil under test (e.g.,
GD) was energized to some level and the current was held
constant. A heat pulse of 3.1 kW for 300 ms was used to drivea whole innermost turn fully normal. The magnitude of theenergy necessary to create the normalcy was found by trial. A
low value was initially used, and the heat pulse time wasincreased until the full normal state voltage across theLUBELL et al.

286
HEATED ZONE

5033-5034

5032-5033
GE

5031-5032

4040-4041

5030-5031

3034-3035

5029-5030

4036-4037

3033-3034

5028-5029

4037-4038

3031-3033

5027-5028

4038-4039

3030-3031

HEATER CURRENT(1.2J/cm3)
JACOIL:HEATER CURRENT(0.4J/cm3)
GDCOILTIME (250 ms/div)

HEATER CURRENT(0.5J/cm3)
GE COIL

FIG. 3. Measured voltage along the conductor of a heated zone in a recovery testat 100% design current.

heated zone was observed.
As the current was increased to
100%, the heat pulse energy was slightly decreased. Recovery
of normal zone to the SC state occurs in about 0.5 s. The
results of such stability tests for all three PB coils areshown in Fig. 3 for the single-coil test.
Cryostatic stability also was demonstrated for all three
PB coils in full-array tests at 100% current and 8-T midplane
field. To our knowledge, stability tests on such large SC
magnets (E = 100 MJ) are unique. The largest stability test
previously was performed on an 11-MJ coil.
The FF coils were not tested in this manner. The supply
for the inductive heaters of the WH coil was not ready at thetime of the single-coil test, and so recovery tests wereIAEA-CN-47/H-I-4

287

postponed until the full-array design-point test. The othertwo FF coils are not designed to be cryostatically stabilized.
It is worth noting that the performance of a FF coil depends
on the mass flow, and in the single-coil tests it was demonstrated that all the FF coils operated reliably with considerably less than the design flow value.
8.2

Simulated Nuclear Heating

All fusion reactors will produce significant nuclear
heating. The actual amount impinging on the SC magnets isstrongly dependent on machine design. Since this amount is
unknown for SC magnets, the recent design studies [6] for TFCX
and the EPR assumed nuclear heating rates of 10 mW/cm 3 for thenominal design and 50 mW/cm 3 for a high-performance (optimistic) design. For the FER project in Japan, a peak value of
1 mW/cm 3 has been assumed [7] . The experiments to be described in this paper are the first to explore this important
problem on SC magnets. Never before have large SC magnets
been constructed with heaters embedded in their windings.
While the LCT heater tests do not fully model nuclear heating,they do take a major step toward answering the question of how
much heat can be generated and removed from a SC magnet
without inducing a quench.
The simulated nuclear heating test (e.g., on the JA coil)
was performed in the following manner. After the coil wasenergized to some current level and held steady, a heat test
was made by simultaneously powering eight heaters for 60 s
(550-740 W ) . The heating time was chosen to get a steadystate environment yet not create too much pressure rise in thecoil and undue heat load to the refrigerator. This corresponds for the JA coil to local unit power density of 70 and
52 mW/cm 3 for the single-coil test and design-point test,
respectively.
The test was performed up to 100% design
current (10.2 kA) in both the single-coil test (B= 6.4 T)and full array (B= 8.1 T) . No normal zone was observed
nor thermal runaway induced. Similar results were found forthe other two PB coils in both single-coil and full-arraytests. It should be noted that in these tests, the distribution of heat was rather localized; therefore, these values
cannot be considered as the tolerance for actual nuclear
heating. Nevertheless, the confidence to design a SC magnet
to sustain some level of nuclear heating is considerablyincreased by these favorable results.
A very different type, but equally informative, simulated
nuclear heating test was performed on the FF (CH) coil. The
heat was generated by constant ohmic heating applied to 11

288

LUBELL et al.

15 —

£

300

4/16/86

18=00/22:30 RUNN0.34

5.6

10

UJ

a:
or
3

3.8
1.5
2.0
TIME (h)3.0

3.5

FIG. 4. Simulated nuclear heating for the forced flow CH coil while it is charged
up to 100% design current.

full-turn heaters while the coil was ramped up to its nominal
current and then slowly discharged.
This procedure wascarried out at two different inlet temperatures of the coolant, namely, at 3.8 and 4.1 K.
Before starting a run theinlet temperature of the helium and the mass flow were set andmaintained constant. The heat was applied to the coil where
the helium enters the winding (i.e., the temperature of theentire winding was increased). The heat input to the winding
was also measured by the heat loss of the winding. The total
heating power was approximately 259 W for both runs, corresponding to a mean power density of 3.6 mW/cm 3 in the heatedvolume. The temperature at the helium outlet and the heatloss of the winding together with the coil current are shownin Fig. 4 for a helium-inlet temperature of 4.1 K.
The 16 TF coils of the Next European Torus (NET) design
are expected to be cooled by supercritical helium in FF mode.
The expected average power density is 0.058 mW/cm 3 , whichcorresponds to a nuclear heating power of 500 W per TF coil.
From the volume of the CH coil winding (2.13 m 3 ) and from theapplied heating power, an average power density of 0.12 mW/cm 3
is calculated. This power density is twice as large as thevalue specified for the TF coils of the NET design. This fineresult should alleviate one of the principal concerns ofplasma physicists.

IAEA-CN-47/H-I-4
8.3

289

Current-Sharing Temperature

Current-sharing temperature(T ) measurements were
performed on two of the FF coils [EU flJbTi conductor) and WH(Nb3Sn conductor)]. Such measurements have never been performed on a large SC magnet. The T
is a linear function ofcurrent (I) in a superconductor. Ir is equal to the criticaltemperature, T (B) (which is a function of magnetic field),when the superconductor is not carrying current and it isequal to the bath temperature (T ) when the superconductor isat the critical current I (B,T) u . e . , the onset of voltage).
cT= T . + [1 - I/I (B,T)] [T (B) - T.J
(1)
cs
b
c
c
b
The experiments were performed in the following manner. The
coil was energized to a specified current (and hence field)
and held constant. Power was supplied to a resistive heaterand thereby a slug of helium many meters long was raised intemperature. Control was good and small increases in thehelium temperature were achieved. The T
is the temperature
for which the smallest voltage (i.e., some current flowing in
the copper matrix) is first observed. Joule heating occurs
above Tcs
For the EU coil the measured values for T
were used toobtain I (B,T). These were in fair agreement with measurements on c short-sample cables extrapolated to the temperatureand field of the full conductor. The defined cooling properties allow operation of the coil at every temperature below
the T . The necessary margin is determined by the increasingheat level of thermal or AC losses. In the single-coil test
the EU coil was operated stably at full current (6.4-T field)with a helium temperature of 5.2 K.
For the WH coil, T
measurements span the current range
from 16 to 63% of design current, corresponding to maximum
self-fields of roughly 1-4 T. At the 4-T field, the coil
operated stably with helium at 11 K. The measurements wereused to determine I (B,T). These values were extrapolated to
B = 8 T and T = 4\ 2 K to compare with expected conductor
current based on short-sample measurements. This comparison
reveals that the WH conductor, although sufficient for fullarray 8-T performance, has suffered serious degradation causedby mechanical strain during manufacture, shipping, and windingof the conductor.
8.4

Dump Tests

All the coils were intentionally dumped from some current
level and some from 100% design current to determine both the290

LUBELL et al.

effect on the cryogenic system and the amount of the initialstored energy which could be removed and dissipated in theroom-temperature dump resistor.
Careful measurements weremade on a dump of the EU coil from 100% current (E = 103 M J ) .
The losses in the EU coil were about 2% (mostly eldy currents
in the case), 2% was deposited in the other coils, and 96% in
the dump resistor. The pressure rise in the coils on a dumpof the whole array was lower than expected, and fast-acting
valves were able to isolate the refrigerator and prevent
damage to turboexpanders.
8.5

Mechanical Measurements

In all coils, significant measurements of strain in
conductor and structure were obtained under various loadconditions. In four coils, displacements of the winding packrelative to the coil case were also measured.
Significant,understandable differences between PB and FF coils wereobserved, reflecting inherent differences in force transmission paths from conductors to structure. These differences
also were evident in correlations of AE signals and compensated voltage measurements (resistive voltages due to winding
movement or flux jumps) for the various coils. No excessive
strain was encountered, but comparison of test data withdesign calculations identified areas in which modeling anddesigns for force transmission can be improved.
9.

CONCLUSIONS

It is gratifying to report that all six LCT coils havebeen energized in turn in the full array to 100% design
current and 8.1 T maximum field. Furthermore, the symmetric
torus test has been performed wherein all the coils simultaneously achieve 8-T midplane field (with the currents ranging
between 88% and 101% of design current). The total stored
energy in this test was 761 MJ. These are the largest coils
to ever produce an 8-T field. The simulated nuclear heating
tests indicate that these designs can likely meet the necessary requirements, even though vapor accumulation in the PB
coils degrades the heat transfer.
Additional studies areneeded, particularly on the mechanical properties, to see if
all the present designs can be extrapolated to reactor-size
magnets.
ACKNOWLEDGMENTS
This work has been performed in the framework of the
IEA-IMPLEMENTING AGREEMENT FOR A PROGRAMME OF RESEARCH AND

IAEA-CN-47/H-I-4

291

DEVELOPMENT ON SUPERCONDUCTING MAGNETS FOR FUSION POWER, ANNEX
I, "LARGE COIL TASK," between U.S./DOE, EURATOM, Japan, andSwitzerland.
Thus, the reported results could only beachieved by the common effort of all participants involved.
The dedication and skill of the operators of the test facility
at ORNL, working around the clock, are gratefully appreciated.

REFERENCES
[1]
[2]
[3]
[4]
[5]
[6]
[7]

HAUBENREICH, P.N., LUBELL, M.S., CORNISH, D.N., BEARD, D.S., Nucl. Fusion 22(1982) 1209.
LUTON, J.N., et al., IEEE Trans. Magn. 21 2 (1985) 238.
LUE, J.W., et al., Fusion Technol. 8, 1, Part 2A (1985) 807.
HAUBENREICH, P.N., et al., in Fusion Technology (Proc. 14th Symp. Avignon, 1986),
Pergamon Press, Oxford (1987).
Proceedings of Applied Superconductivity Conference, Baltimore, MD, 1986 (to be published byIEEE Trans. Magn.).
TFCX Pre-conceptual Design Review, Princeton Plasma Physics Laboratory, 6-8 June 1984.
IIDA, F. (JAERI), personal communication.

DISCUSSION
R.W. CONN: It is good to see such comprehensive results after ten years ofprogramme building and development. What are your future plans?
M.S. LUBELL: The next stage in this programme will be tests with pulsed
fields with all the coils, and it is hoped that it will start very shortly. This will befollowed by extended tests with some of the coils, using higher fields and higher temperatures; it is particularly important to test their performance with out-of-planeloads.

IAEA-CN-47/H-I-5

RADIOLOGICAL IMPLICATIONSOF LOW ACTIVATION STEELSAS STRUCTURAL MATERIALSIN FUSION REACTORS
R. HANCOX, B. MITCHELL1, G.J. BUTTERWORTH
Euratom-UKAEA Fusion Association,
Culham Laboratory,Abingdon, Oxfordshire,
United Kingdom
Abstract
RADIOLOGICAL IMPLICATIONS OF LOW ACTIVATION STEELS AS STRUCTURAL
MATERIALS IN FUSION REACTORS.
New austenitic and martensitic steels are being developed as possible alternative structural
materials in a fusion reactor, with elemental modifications designed to reduce their levels of activity
after exposure to high energy neutrons. The radiological advantages of using such materials are
discussed. No advantage is found for severe accident situations; there may be some benefit in reducing
occupational exposures, and the main advantages appear in long term waste disposal and the possibilityof recycling structural materials.

Introduction
Nuclear fusion research is undertaken because it offers theprospect of a new energy source based on abundant fuel resources
and with safety and environmental advantages. The ideal nuclear
reaction would be one in which all reaction products are charged
and remained in the plasma, minimizing radiation hazards andallowing a wide choice of constructional materials for thereactor. The first generation of fusion reactors will, however,almost certainly operate with deuterium-tritium fuel to reduce
plasma confinement requirements and capital costs, with theresult that 14 MeV neutrons will activate the structural
materials. It is therefore of interest to consider alternative
structural materials that minimize the hazards associated withthis induced radioactivity, and to assess the radiologicalimplications of possible 'low-activation' materials.
In thiscontext, 'low-activation' is used to denote any material having a
residual radioactivity below that of conventional stainlesssteels.

1

Associated Nuclear Services, Epsom, KT17 1HB, UK.

293

294

HANCOX et al.

The development of low activation materials for fusion reactors
has followed three routes - elemental substitution in existing
and well-tested materials such as steels, the development of newor unproven alloys, and the use of isotopically tailored
elements. Only the first of these offers the possibility of thelarge scale, economic
production of material in time for ademonstration fusion reactor, although the other routes may
eventually offer materials with greater advantages. Development
effort has therefore concentrated on alternative austenitic andmartensitic steels with modifications designed to eliminate
elements that give rise to significant long-lived activity after
irradiation with 14 MeV neutrons [l].
Materials and assessment indices
Four steels were considered in the assessment which represent a
range of activation behaviour and all of which possess
potentially suitable engineering characteristics. These were:1.

AISI type 316 austenitic stainless steel (0.03%C, 17%Cr,
12%Ni, 2.5%Mo).
This material has been chosen forseveral reactor designs, including INTOR and NET, on thebasis of the broad data base obtained from its use infission reactors.

2.

The martensitic steel FV448 (0.15%C, ll%Cr, 0.2S%V, 0.3%Nb,
0.6%Mo, 0.75%Ni). This alloy is similar to the DIN 1.4914
steel which is considered as an alternative for NET and has
been proposed on the basis of reduced swelling underirradiation and reduced thermal fatigue.

3.

A low-activation austenitic composition OPTSTAB, in which
the usual nickel component is replaced by manganese (12%)
and nitrogen (0.3%), and molybdenum replaced by tungsten(2%) [2].
4.

A low-activation martensitic steel LA7, based on FV448
steel, with molybdenum replaced by tungsten (3%) [3].

For comparison, calculations were also made for the vanadiumalloy V-15%Cr-5%Tl and silicon carbide. There is, however, lessinformation on possible impurities in these materials and agreater uncertainty as to their mechanical and corrosion
properties in the fusion reactor environment.
The specificactivities of these materials as a function of time were
calculated and the effect of impurities assessed as far as ispossible on current information. From the calculated specificactivities, several hazard indices were derived as a basis forassessment of the radiological implications of the selectedmaterials. These indices were:IAEA-CN-47/H-I-5

295

1.

Specific radioactivity.
This parameter is the primarymeans of expressing the activation of materials, though it
takes no account of the widely varying nuclear, chemical or
biological properties of the different radionuclides.

2.

Contact Y dose rate.
For convenience, the contact
dose rate at the surface of an infinite half-space of
uniformly contaminated material was calculated.

3.

Biological hazard index for ingestion. This was expressed
as a multiple of the ICRP annual limit of intake forradiation workers relative to one kilogram of the activatedmaterial.

4.

Biological hazard index for inhalation.
This wascalculated in a similar way to the index for ingestion.

Although the last two indices provide a simple way of comparing
the potential hazards of different mixtures of nuclides, they
fail to take account of mechanisms for the release and transportof radionuclides to the point where they are ingested orinhaled, or of differences in material properties that affectthe probability of their release during accidents.
Examples of the variation of two of these indices with timeafter shutdown of the reactor are illustrated in figure 1. The
values correspond to the relatively hard neutron spectrum at thefirst wall position in a reactor with liquid lithium breeder,
following 2.5a irradiation at 5 MW-nT2 neutron power flux. On
the basis of these indices the radiological implications ofalternative materials have been assessed in several situations.
Maintenance of the coolant system
Contact Y dose rates around the coolant circuit of a watercooled fusion reactor may be comparable with those found inpressurized water fission reactors, where they are a major
contributor to occupational radiation doses.
The maincontributions in the PWR arise from deposition of 5 8 Co, 51*Mn and60
Co, with 58 Co arising mainly from nickel released by corrosion
from the Inconel steam generator tubes and then irradiatedduring its passage through the high neutron flux region.
The above considerations indicate that it will be necessary to
avoid the use of high activation materials throughout thecircuit in order to reduce y dose rates around the coolant
circuit in water-cooled fusion systems.
The use of the
low-activation martensitic steel would only lead to alimited improvement since it would provide similar sources of5l*Mn, while the low-activation austenitic steel could give

HANCOX et al.

296
~

I

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10'

10"

10'

10"

10'

10
Id 1w 1m 1y 10y toy 10/ 10y
10At TOU . t10. t . t10. M
6

8

2 3 A
W lw 1m 1y lOy lOy lOy lOy

10
106
108
Time after shutdown (s)

1010

1012

FIG. 1. Contact gamma ray dose rate (Sv-h'1) and specific ingestion hazard index (ICRP annual
limit of intake per kilogram) for representative materials after irradiation for 2.5 years at S MW- m ~z.

higher doses on account of its higher manganese content.
Theeffect of using low-activation materials is difficult toestimate, however, since corrosion and transport properties will
be at least as important as the nuclear properties.
Nevertheless, relative to fission reactors, fusion systems havethe advantage that the use of corrosion inhibitors is not
limited by their nuclear properties.
In helium-cooled systems, corrosion by impurities and erosion by
particulates provide mechanisms for the transport of materials

IAEA-CN-47/H-I-5

297

into the high flux regions.
However, neutron sputtering isbelieved to be the most important means of introducing wallmaterials into the coolant and, therefore, coolant contamination
should be characteristic of first wall/blanket piping materials.
Calculations for conventional steel walls [4] indicated that the
nuclides 5**Mn and 5 6 Mn would give rise to about 60% of total
contact Y dose rates. Hence the use of low-activation steels
only offers the prospect of slightly reduced y dose rates from
coolant system components.
The use of vanadium alloys mightoffer substantial reductions in y dose rates during maintenanceof the cooling circuits.
Reactor maintenance and component repair
To minimize reactor down-time, the maintenance strategy willprobably involve the replacement of components or modules,
followed by their repair in workshops after an appropriate
cooling time. Only remote maintenance will be used within theshielding and it is highly probable that similar techniques will
be used throughout the reactor hall, though the option oflimited human access to regions outside the shielding may remain
as a possibility.
Low-activation steels show no significant advantage for shorttimes after shutdown, when dose rates are dominated by 5 4 Mn.
The dose rate for the vanadium alloy falls quite rapidly overthe first week or so, associated with the decay ofSc, but
this fall is not maintained as **^Sc and ^ C r dominate for longer
times and a factor of 10 to 100 below the steels appears to be
the best that can be achieved with this particular alloy. Dose
rates for silicon carbide fall very rapidly and are likely to be
dominated by impurities after a few days.
For workshop repairs to be performed routinely by manual
operations, it is necessary for y dose rates to fall to values
typically below 25 nSv-h"1. Pure low-activation steels would haveto be stored for about 100 a before dose rates diminished to thisvalue.
In addition, part per million concentrations of
impurities such as niobium, silver, europium, terbium, holmium,
iridium and bismuth give rise to dose rates comparable with or
greater than this value [5]. Hence it is likely that, althoughradiation shielding requirements may be somewhat reduced,
low-activation steels will not offer an alternative maintenancestrategy.
Release of activation products in severe accidents
The nature of mechanisms that could cause significant releasesof activation products is a subject of current debate, with

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HANCOX et al.

lithium fires and radioactive decay heating as the most likely
causes of extensive mobilization of radionuclides.
On theassumption that exposure will largely be caused by drinking
contaminated water and by eating contaminated crops and animals,
rather than by direct inhalation of radionuclides, the relative
merits of the materials may most easily be compared using theingestion hazard index. Those radionuclides that are active fortimes up to a few years may be expected to be significant. Lowactivation steels exhibit no significant advantage during thisperiod because the nuclides ^ M n and ^ F e which contribute mostto the ingestion hazard index arise from the activation of iron.
The alloy V-15Cr-5Ti has a hazard index that is at least afactor 10 lower than those of the steels and the hazard index ofsilicon pure carbide is about six orders of magnitude lower.
Radioactive waste disposal
The ingestion hazard index is considered to be the mostappropriate quantity for the comparison of different materials
for the purposes of radioactive waste disposal, since it isassumed that transport in contaminated groundwater, followed byingestion in plants or animals represents the most likely routefor human exposure.
For waste disposal it is appropriate tocompare the materials for times greater than 100 years after
removal from the reactor. After this time, all 6 0 Co will havedecayed. The low-activation materials nevertheless demonstrate
a substantial advantage over the conventional steels because of
the much lower levels of the very long-lived activation products
arising from irradiation of nickel (63Ni and 5 9 Ni),. molybdenum( 9 3 M o and 9 9 T c ) and niobium ( 9 3 m N b and 91| Nb) . The most
persistent radionuclide resulting from irradiation of the
low-activation steels is ll*C, arising from the nitrogen content.
Although the long-term ingestion hazard index from this llfC issmaller, by a factor 100, than the indices of the conventional
steels, it may be noted that the specific activity of 1I+C issome 10 to 100 times greater than that permitted in 'low-level1
waste under current UK regulations.
Recycling of structural materials
Resource limitations or restrictions on disposal may necessitate
the recycling of structural materials. Recent studies indicatethat the operations involved in the recycling of steel scrap arefeasible at surface y dose rates up to about 10mSv-h~* [6],compared with the current notional 'hands-on' dose rate of
25nSv-h~*. Dose rates of conventional steels, however, do not
fall below 10 mSv-h
for at least several hundred years andtherefore recycling is not a reasonable possibility with thesematerials.

IAEA-CN-47/H-I-5

299

Surface dose rates of pure low-activation steels decline to lessthan 10 mSv-h" 1 after a few decades, 5**Mn and 60 Co being thedominant radionuclides during this period.
Thus recycling
should be possible with low-activation steels. It should be
noted, however, that dose rates during this period are dependent
on the neutron spectrum in the blanket and calculations fordifferent blanket designs have given variations by a factor 2 inthe delay required.
The existence of nickel and cobalt
impurities causes increased dose rates, but it is the presence of
impurities such as silver, terbium and niobium at the part per
million level that gives rise to long-lived y emitters andcauses the most marked departures from the characteristics ofthe pure material. For a mixture of impurities based on current
estimates, surface dose rates for low-activation steels fall to
10mSv-h~1at times only slightly greater than those corresponding
to the pure material, but they decline very slowly thereafter.
The continued recycling of materials through the reactor willlead to some differences in these contributions compared withthose from the first irradiation.
The nuclide 91+Nb may be
expected to build up linearly with the number of irradiation
cycles, and °"co, although decaying almost completely between
irradiations, will undergo accelerated build-up during
irradiation as its precursor, stable ^ C o , is produced from58
Fe.
Since ^°Co is produced via two successive (n,Y)
reactions, its rate of production is sensitive to the neutron
spectrum.
Conclusions
The use of low-activation steels in a fusion reactor is expected
to offer radiological advantages mainly in relation to recycling
and long-term waste disposal of structural materials rather thanto short-term maintenance or repair.
Whereas the post-service dose rates of conventional steels
remain for indefinite periods above the level at which recycling
is considered feasible, the low-activation compositions would
allow reclamation after storage for about 70 a. The storage timerequired, as well as the dose rate level reached, is controlled
by certain trace elements and it is important to achieve thelowest feasible concentrations of these impurities in order togain full benefit from the use of the low-activation materials.

REFERENCES[1] JARVIS, O.N., Selection of Low-Activation Elements for Inclusion in Structural Materials for
Fusion Reactors, Rep. AERE-R-10496, UKAEA, Atomic Energy Research Establishment,
Harwell (1982).
[2] BOTT, A.H., et al., The Development of Austenitic Stainless Steels as Low-Activation Structural
Materials for Fusion Reactors, Rep. CLM-R-255, UKAEA, Culham Lab., Abingdon (1985).

300

HANCOX et al.

[3] DULIEU, D., et al., Development of Low-Activation Martensitic Stainless Steels, Preprint
CLM-P-771, UKAEA, Culham Lab., Abingdon (1986).
[4] BICKFORD, W.E., Transport and Deposition of Activation Products in a Helium-Cooled Fusion
Power Plant, Rep. PNL-3487, Battelle Pacific Northwest Labs, Richland, WA (1980).
[5] GIANCARLI, L., Impurity activation and surface y-dose levels in first wall structural materials,
J. Nucl. Mater. 139 (1986) 1.
[6] GOMER, C.R., et al., "Radioactive structural waste management in fusion power generation",
Proc. ENC '86, Geneva, June 1986.

DISCUSSION
Y. SEKI: The results of the calculations that you described prompt me to ask,
in view of the large uncertainties in the activation cross-sections, whether you haveany plans to conduct irradiation tests using 14 MeV neutrons?
R. HANCOX: We plan more detailed cross-section calculations and sensitivity
studies, together with an analysis of the extent to which impurity levels can be
reduced in commercial materials. However, irradiation tests will eventually be necessary to confirm the calculations.
R.J. GOLDSTON: You indicated that there were three approaches to reducing
activation in a fusion reactor: elemental selection, new alloys and isotopic selection.
You chose to adopt the first of these, because the studies could be implemented on
the shortest time-scale. However, you only found long time-scale advantages. Could
you say whether you think advantages could be gained from the other approaches?
R. HANCOX: We expect greater advantages from the other approaches, but atgreater cost. These advantages will generally relate to the medium and long term, butwill always be limited by the difficulty of reducing impurity levels in the large scale
production of materials.

IAEA-CN-47/H-H-1

A MUON CATALYSEDFUSION-FISSION REACTOR
S. ELIEZER
Plasma Physics Department,
Soreq Nuclear Research Centre,
Yavne, Israel
T. TAJIMA, M.N. ROSENBLUTH
Institute for Fusion Studies,
The University of Texas at Austin,
Austin, Texas, United States of America
Abstract
A MUON CATALYSED FUSION-FISSION REACTOR.
Absorption of one muon in matter and muon induced fusion of deuterium and tritium can induce
about one hundred nuclear D-T fusion reactions in a liquid density medium during the muon lifetime(2.2 x 10~ 6 s). This end result of nucler fusion follows a chain of atomic and molecular processes.
The 'energy cost' for the creation of a muon is estimated to be about 5 GeV. Therefore, if one muon
catalyses about one hundred D-T fusions, the energy output is less than 2 GeV per muon, so that noenergy gain seems to be possible from 'pure' fusion nuclear reactions. However, it seems possible togain energy by combining the catalysed fusion with fission blankets. Two new ideas are presented whichshould improve the muon fusion reactor concept. The first idea is to combine the target, the converterof pions to muons and the synthesizer, making them one system. This is accomplished by injecting a
tritium or deuterium beam of 1 GeV per nucleon into the D-T fuel contained in a magnetic mirror. The
confined pions slow down and decay to muons; these are confined in the fuel, causing little muon loss.
The second idea is to collect the beam passing through the target and to recirculate it, while directing
the strongly interacting portion of the beam to an electronuclear blanket. Thus it is possible to reduce
the necessary quantity of tritium by about two orders of magnitude and to improve the energy gain
compared with previously suggested concepts. The present concepts are based mostly on readily available technologies and on known physical processes and data.

Muon induced fusion of deuterium (D) and tritium (T) is a remarkablephenomenon since the negative muon can induce about one hundred nuclear D-T
fusion reactions in a liquid density medium during its lifetime (2.2 x 10 ~6 s). Thisend result of nuclear fusion follows a chain of atomic and molecular processes [1].
The most crucial step in the physics of the muon catalysis cycle is the resonant
formation of the D-T muon molecule; this increases the probability of the end result
by at least two orders of magnitude (relative to non-resonant D-T muon formation).
This has been theoretically suggested and recently experimentally confirmed to acertain extent. This finding has aroused considerable excitement in the fusion
301

302

ELIEZER et al.

TABLE I. VALUES OF YIELD AND ENERGY REQUIRED TO
CREATE ONE NEGATIVE PION
Energy3

Yield3y*

T o - energy per nucleon

Eo = y ^ T o

(a)

(b)

(a)

D (1 GeV) -- D

0.22

0.17

4.5

5.9

T (1 GeV) - D

0.27

0.21

3.7

4.8

T (1 GeV) - T

0.50

0.25

2

4

T (1 GeV) - (70 % D + 30% T)

0.31

0.22

3.2

4.6

(b)

Column (a) gives results obtained by taking multiple scattering into account; in column (b), single
collisions are considered.

community and it motivates us to explore concepts of energy producing nuclear
reactors utilizing muon catalysed fusion.
The 'energy cost' for the creation of a muon is one of the most important practical parameters in an analysis of the relevance of muon catalysed fusion for energy
production. Muons are produced during the decay of pions, which can be created by
nucleon-nucleon collisions. The energy threshold for pion generation is about
500 MeV of projectile kinetic energy and the process of negative muon generation
seems to be most effective for nucleonic projectiles with a kinetic energy of 1 GeV
per nucleon. An optimistic estimate is that 5 GeV of energy is required to produce
one negative muon (see Table I). The yield yT is the number of pions created perunit energy of projectiles. Thus, if one muon catalyses about one hundred D-T
fusions, the energy output is less than 2 GeV per muon (one D-T fusion gives
17.6 MeV), so that no energy gain seems to be possible from 'pure' fusion nuclear
reactions.
Although it appears not possible at present to achieve an energy gain by purefusion, it is possible to gain energy by combining catalysed fusion with fission
blankets [3]. We present a reactor concept based on three ideas:
(a)

(b)

A high energy beam of tritium or deuterium is injected into the deuteriumtritium (D-T) fuel and, after passage through the fuel, part of the beam iscollected for reuse while the portion of the beam that underwent interaction is
directed into an electronuclear blanket (see Fig. 1).
The pions created in the target are surrounded by the deuterium and tritium fueland are magnetically confined until they slow down and decay into muons whichcatalyse the fusion in situ.

303

IAEA-CN-47/H-n-l

blanketbuncher (cooler)

FIG. 1. The mirror reactor concept. The gas deuterium-tritium fuel at the liquid hydrogen density inthe 400 K temperature range is contained in the mirror. The target (tritium) is held in the middle ofthe mirror. The tritium or deuterium beam is injected along the mirror axis; an undeflected portion iscollected for recirculation and a deflected portion is absorbed by electronuclear blankets. Pions are
confined by the mirror field. Fusion neutrons are absorbed by the surrounding fission blankets.
The heated fuel is cooled by circulating flow of the fuel.

(c)

The fusion created neutrons are absorbed by blankets to breed fissile matter forenergy production.

This concept differs from Petrov's concept [3] in (a) and (b). Instead of keeping
the target, converter and synthesizer separate, we combine these three functions. The
D-T fuel is the target and the produced pions are trapped in the fuel where they slow
down before they decay into muons. The muons are created in the fuel and trapped
there, catalysing the D-T fusion through the atomic and molecular processes until
they in turn decay. In this way we solve one of the most difficult problems of muon
catalysed fusion, namely the efficient trapping of muons in a relatively small physicalvolume. The present concept solves this problem by creating mesons in the fuel andby confining them magnetically. This can be called a synergistic approach.
Figure 1 is a sketch of a version of our reactor concept. The fuel is driven bya tritium (or deuterium) beam which is retrieved in part after passing through the target. The significantly scattered portion of the beam feeds into the electronuclear
blanket. The fusion created neutrons are captured in the fissile blanket surrounding
the fusion fuel. Figure 1 shows a magnetic mirror configuration. The mirror is filledwith a pressurized gas mixture of deuterium and tritium. The gas is circulated
through the mirror and passes a cooling section between traversals. The mirror is
enclosed by 238U or 232Th blankets with admixed lithium for the breeding of Pu,
232
U and tritium. Magnets provide a magnetic field configuration with a mirror ratio
Rm. The field at the mirror throats is typically 10 T. Our study indicates that for
1 GW of electric power we need about 2 kg of tritium, contained in a 20 cm radius

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ELIEZER et al.
Muon cotalyzotion chamber
Converter

TorgetIV)(lord)<l-r)PA(n)j (n)JrPAO! tBlanket ( 2 M U Of " * T h ond * L i )" • u « n - »ission^" 9 Pu; M 2 T h . n - fission(Pu)i
X B - Atomic energy
plantsai

KPA

Electric generator

7th p out

Aux. power

FIG. 2. Scheme of energy production in a hybrid muon catalysed fusion-fission
electronuclear reactor.

mirror (with mirror ratio 1.5) at half-liquid hydrogen density (about 1500 atm
pressure at the temperature of the fuel, T = 500 K).
The energy and particle flow for the muon catalysed fusion-fission reactor isshown schematically in Fig. 2. A particle accelerator of tritium (or deuterium)
delivers a power PA (about 1 GeV per nucleon) with an efficiency ijA into a /x~
generating catalysis chamber which contains the high density D-T fusion fuel. In thischamber the muon catalysis fusion takes place, releasing e ^ = 17.6 MeV per
fusion. The beam power PA is used for /* ~ production as well as for electronuclear(EN) direct beam-blanket interaction; r is the fraction of PA used for the fusion
processes, and the ir~ to n~ efficiency conversion y^ is estimated to be about 0.5
for <f> = 0.5 liquid hydrogen density. For the tritium-tritium interaction, r = 20%.
Then, (1-r) PA is used for the electronuclear portion where the direct beam-blanket
interaction is used for spallation breeding of the fissile fuel (e.g. 239Pu or 233U fora blanket of 238U or 232Th, respectively). The gain value for fusion in our schematic
diagram, Fig. 2, is defined by
Q =

(1)

IAEA-CN-47/H-n-l

305

where X^ is the number of fusion reactions catalysed by one muon. y "', y,,"1 and XM
are the most important input (physical) parameters for the fusion-fission processcatalysed by the negative muons.
The total gain factor is defined in Fig. 2 and is given by„ _ y^Vx ZJofen. +" 1 v" 1y

EN denotes the energy gain from 'direct' beam-blanket interaction and /tic denotes
the energy gain from muon catalytic processes, with input energy on target(l-t)yrly^lT0
for the EN process, and ry^'y^To for the p.c process. The EN
contribution is KEN = ZeT0f€fls/T0 and the muon catalysed gain is given by thesecond term. efis = 0.2 GeV is the uranium fission energy (for a 238U blanket),
Ze = 20 fissions per GeV in the 'direct' beam-blanket interaction (i.e. the EN
process), TOf is the beam kinetic energy before collision with the blanket and To is
the initial kinetic energy of the beam before it hits the target (for pion — muon
creation). 6f = 1 is the number of fissions in the blanket caused by one 14 MeV
neutron from the fusion process. The above reactor produces a considerable power
on its own. Furthermore, the fissile material produced in the blanket (e.g. 239Pu)
can be used to run between three and six satellite fission plants of equal output. XR
in Fig. 2 denotes the number of such satellite fission plants. Of course, the value of
XR depends on the design and breeding ratio of those plants and is thus somewhat
arbitrary. However, assuming a certain value permits an intercomparison between
this scheme and other schemes — pure EN, fusion hybrid, etc. We estimate that the
number of fissile atoms bred per fission in the blanket is about the same for EN andfusion neutrons (taking account of tritium breeding). The total electrical gain Ktot isgiven by

Kt0, = ^^= Ha (1 + XR) K(3)

where TJU, = 0.35 is the thermal power to electric power conversion efficiency. The
fraction a of the total electric power that has to be recirculated in order to operate
the accelerator and the auxiliary facilities is given by (see Fig. 2)
1

K (1 + X )

=

1
K

306

ELIEZER et al.

TABLE n. COMPARISON OF THE PRESENT REACTOR CONCEPTWITH OTHER DRIVEN REACTOR CONCEPTS2
Estimated
efficiency
ij (driver)

SchemeAccelerator

0.6

licFF

Accelerator

0.6

Laser

Drivera(energygain)"

T
(eV)

V
(m3)

M,
(g)

13.2

0.13

0.1

1

10J

EN

8.4

0.2

—

1

0

0.05

ICF
ICFF

= 430
Q = 75

0.13

104

1-10

10 " 2

Accelerator

0.6

ICF
ICFF

Q. = 37

0.13

104

1-10

10

Accelerator, RF, etc.

0.6

MF
MFF

SL = 37

0.13

10"

102

10" 2

Q = 6.5_2

Inertial confinement fusion (ICF) and fusion-fission (ICFF) with laser and ion beams as drivers, theelectronuclear (EN) breeder and the magnetic fusion (MF) and fusion-fission (MFF) reactors. The
order of magnitude for the fuel temperature (T), the reactor volume including the blanket (V) andthe tritium mass (M,) in the reactor at a given time are listed. M, does not include the inventory.
The thermal efficiency included in K,ot is assumed to be ifo, = 0.35. The number of atomic nuclear
plants is taken as XR = 5, The value of Q is taken in such a way as to yield a = 0.13 in theequation Q = l/OfoifA01) for the pure fusion scheme and Q = l/[»Jth'?A (1 + X R ) « ] f° r fusion-fission
schemes.

Substituting into Eqs (3) and (4) the following parameters for a fuel containing
30% tritium and 70% deuterium at 0.5 liquid hydrogen density:
Ze = 20 GeV" 1 , efis = 0.2 GeV, 5f = 1, To = 1 GeV, XR = 5, rjA = 0.6,
V0l = 0.35, y,, = 0.5, yT = 0.31, Tof = 0.8 GeV, X, = 100,
we obtain:
KEN = 3.2, K^ = 3.1, K = 6.3, K^, = 13.2, a = 0.13.
Note that although the fie input energy is smaller by a factor of four than the
EN input energy, the output gain from /tc is about the same as the EN gain. An
important result of this calculation is the estimate of the parameter a, which can serve
as a figure of merit in a 'driven' nuclear reactor. As usual, the assumed output
electricity for the system, PE = (1 + XR) x 109 W, with a - 0.13 and XR = 5,
requires an accelerator power of PA — 8.5 X 108 W, which for 1 GeV per nucleon
projectile implies an average current of 0.85 A.
A comparison of the present reactor concept with other driven reactor concepts
is presented in Table n. Since the reactor discussed in this paper cannot achieve
ignition, a comparison with self-ignited reactors is difficult. The drivers considered

IAEA-CN-47/H-n-l

307

in the table are: lasers and particle beam accelerators for inertial confinement fusion(ICF), accelerators for electronuclear (EN) breeders and for muon catalysed fusionfission (jtcFF) reactors, and radiofrequency (RF) wave generators and/or accelerators for magnetic fusion (MF) reactors. Also shown are the ICFF and MCFF reactors, which are fusion-fission hybrid schemes for the respective processes. The
efficiencies of the drivers are taken optimistically. Numbers of merit for reactors may
be taken as the total energy gain Ktot (and the fraction of driving energy, a). It isdifficult to predict the science and technology necessary for the different schemes.
We thus indicate in Table II the equivalent energy gain Q necessary to achievea = 0.13, which is the value we have estimated for the present scheme. The orderof magnitude for the fuel temperature (T in eV), the reactor volume including theblankets (V in m 3 ), and the necessary tritium mass in the reactor at a given time
(Mt in g) are also given in Table II.
The reactor concept discussed in this paper merits further serious investigations
on the basis of our preliminary study. It is worth pointing out that the present scheme
is based on readily available technologies and data available from experiments.

REFERENCES[1]

GERSHTEIN, S.S., PONOMAREV, L.I., in Muon Physics, Vol. 3 (HUGHES, V.W.,
WU, C.S., Eds), Academic Press, New York (1975) 141.
[2] JONES, S.E., et al., Phys. Rev. Lett. 51 (1983) 1757.
[3] PETROV, Yu.V., Nature (London) 285 (1980) 466.

DISCUSSION
T. TAKEDA: If I understand you correctly, the most important aspect of yourstudy is the improvement of the muon utilization efficiency by combining a numberof components of the reactor. However, the vessel window seems to cause a considerable projectile energy loss. What is the ratio of the muons produced in the vessel
window to the muons produced in the fuel?
S. ELIEZER: The muon losses in the window and in the wall of the vessel seemto be less than 5 %.
T. TAKEDA: Do you think it is possible to further increase muon production
through multiple production channels with an increase in projectile energy, forexample to 2 GeV?
S. ELIEZER: By increasing the projectile energy to more than 2 GeV per
nucleon, new and undesirable particles are created, such as the neutral pion TT°; thisis wasted energy as far as our scheme is concerned. However, the optimum valueof the projectile energy — which is between 1 GeV and 2 GeV — was not calculated
by us.

308

ELIEZER et al.

M. GOLDSWORTHY: I would like to ask you three questions. First, could you
describe the nature and size of the accelerator mechanism in your scheme? Second,
are we still talking about conventional particle accelerators, and are there any
alternatives? And, finally, is it possible to combine the accelerator and the fusion
reactor vessel?
S. ELIEZER: Regarding your first point, we did not consider in detail thenature and size of the accelerator. Second, there are certainly other alternatives —
possibly more compact accelerators with very high efficiency. Third, I believe that
Professor Herrera from Mexico has suggested a scheme for combining theaccelerator and the fusion reactor vessel, but I have no idea of the details. In any case,this idea looks very attractive.
H. HORA: To prevent muon sticking to helium, could one perhaps use
monochromatic radiation, i.e. an X-ray laser, for photo-ionizing the muon? How
would the gain increase without this sticking?
S. ELIEZER: So far we have no X-ray lasers with an energy of only a few
kiloelectronvolts. However, even if we used synchrotron radiation or if we had suchan X-ray laser, then in order to solve the 'sticking' problem we would need very highfluxes, which induce plasma D-T instead of gas D-T reactions. In a plasma medium,
all the advantages of molecular physics are lost, in particular the resonant formationof the D-T muon mesomolecule. However, for high density plasmas (about
1022 cm" 3 ) no calculations of D-T muon formation have been performed so far.

IAEA-CN-47/H-n-2

EFFECT OF ANOMALOUS TRANSPORT ON
PARAMETERS OF A COMPACT TORUS BASEDQUASI-STATIONARY REACTOR
V.A. BURTSEV, V.M. KOZHEVIN, A.N. MAKHANKOV
D.V. Efremov Scientific Research Instituteof Electrophysical Apparatus,
Leningrad, Union of Soviet Socialist Republics
Abstract
EFFECT OF ANOMALOUS TRANSPORT ON PARAMETERS OF A COMPACT TORUS
BASED QUASI-STATIONARY REACTOR.
Compact tori are attractive because of their high beta value. But high beta results insharp gradients of particle density and magnetic field, causing the development of anomaloustransport processes. The paper presents a study of the effect of these processes on the parameters of a quasi-stationary compact toroid reactor based on a field reversed configuration (FRC),
D-T, DD-Cat and D3He being used as fuel. Special attention is paid to the investigation of thepulsing mode of reactor operation, which makes it possible to realize plasma self-heating in aconfinement chamber and a quasi-stationary fusion burn.

1.

INTRODUCTION

Compact toroid (CT) fusion reactors based on FRC are attractive becauseof their relatively simple magnetic field geometry and their high beta value.
Previous theoretical analysis of CT reactor aspects has indicated the possibilityof creating a reactor with relatively small dimensions and unit power [1-3]. It
was not, however, taken into account that high beta resulted in high gradientsof density and magnetic field, causing anomalous transport processes and areduction in plasma lifetime. So, this study examines the effect of anomaloustransport on the geometrical and power parameters of a CT quasi-stationary
reactor. Three types of fusion fuels are analysed: D-T, DD-Cat, and D|He.

2.

MAJOR PROCESSES DETERMINING THE CT LIFETIME

To study the possibilities of a CT fusion reactor, it is, first of all, necessary
to know the dependence of lifetime of particles ( T ^ ) , magnetic flux (r^) andenergy (T%) on the plasma parameters. The results of experimental andtheoretical studies of the CT particle lifetime (r^) are in good agreement [4].
309

310

BURTSEVetal.

Theoretical estimates indicate the dominant role of anomalous transport, due
to lower hybrid drift instability (LHD) [4, 5]. An analytical expression for theparticle lifetime, determined by LHD, is derived by a thin flux layer model [6]and may be expressed in terms of the plasma confinement time in a linear theta
pinch (T||):
8<J3> /

TN = X|I^I V 7Vu

0)Here, x s = r s /r w is the relative separatrix radius, and u is a dimensionlessparameter defined by'8 m i V'8

S3'4

J m~J

(l+Te/T;)" 8 ""(2)

where S = R/pj 0 is the ratio of the magnetic axis radius to the ion gyroradiuson the separatrix, and k = fi/R is the elongation of the toroid.
Among the transport processes as determined by Coulomb collisions, ion
thermal conduction appears to be the most dangerous one. The energy confinement time, determined by this process, equals r^ « S2 TJ, where TJJ is the ionion collision time. At the same time for large values of S, which are typical of
the reactor parameters, it is easy to obtain the scaling, r^ a S 4 from (1) and(2) and there is a certain danger of Tgj being equal to T^- However, the caseof r N s» T Ei may only arise at S «* 1000. The value of this parameter required
for reactor purposes is found to be S * 300, so that we assume ^N^TEIIt should be noted, at present, there is no good agreement between
experimental and theoretical results of the magnetic flux lifetime investigation.
The observed lifetime appears to be considerably shorter than the theoretical value, indicating that anomalous resistance prevails near the magnet axis.
We assume an ion acoustic instability to develop in this region since Ohmic
heating due to high current density may result in realizing the condition
T e > Tj. Estimates show that the second condition for the development of theion acoustic instability, i.e. v D > cs (v D is the drift velocity, cs the ion acoustic
velocity), can be realized. We assume that v D «* cs throughout the process,since, with growing v D /c s , the resistivity and, accordingly, the rate of diffusionof the magnetic flux increase (9B/9r| r = R being decreased), and, for v D /c s < 1,
the ion acoustic instability is damped, resulting in an increase of 3B/9r| r = R.
In the assumption made above, the magnetic flux lifetime may be derived from

at

311

IAEA-CN-47/H-H-2

9B
3r

47rne

(4)
mi

As a result, we obtain/dln^V1= I
I ( 70

at /

(5)
mPc

2

0

2

where \ =1J\, u = r /R - 1.
Note that Eq. (5) is true for the case of the rate of magnetic flux diffusionnear the magnetic axis being less than in the rest of the plasma bulk where theanomalous resistance is determined by the LHD. Otherwise, the flux confinement time is to be comparable with the particle confinement time (r$ «* r N ) .
In general,(6)
A comparison of Eq. (6) with the results obtained on the TRX-1 device [7]demonstrates that the former are in good agreement with the experimental data
(Fig. 1). For reactor parameters, it follows from Eqs (1), (2), (5), and (6) thatr $ > TN.
Thus, on the basis of the analysis, we assume in what follows that the
lifetimes of particles, magnetic flux and energy are equal and determined by
Eq. (1).

200

o

a

150
100
A 9 0.20 < P i 0 < 0.25
O 0.26 < p i 0 < 0.30
A 0.31 <p i o >;0.35

50
0
0.3

0.4

0.5
0.6
x s =r s /r w

0.7

FIG. 1. Comparison of theoretical and experimental values.

312

BURTSEV et al.

3.

COMPARISON OF FRC WITH TOKAMAK

Using the scaling relation proposed by Merezhkin and Mukhovatov [8],let us now compare the energy confinement time in a tokamak with the particle
confinement time in CT:
r E = 3.5 X 1CT21 ( a / R ^ R X q - V(7)Here, R and a are the major and minor radii, respectively, and q is the stability
margin.
Using the dependence of q on <3max,
0 m a x = a/Rq2(8)

and taking |3 m a x = 5%, we obtain for Eq. (7) the following form:
rE = 2.6 X KT
For R/a = 4, we obtain:_i

7 E = 1.45 X 10" 22 N T e 2(9)Comparing Eqs (1) and (9) we can show that the energy lifetime in atokamak ( T E ) is only two to three times greater than the particle lifetime in
CT (TN ), when the reactor volumes are equal and m «* 3 X 1020 m~3 • s. At
the same time, the ratio of the magnetic energy of the tokamak to that of FRC
equalsWCT

nCT

i.e. the presence of high beta in CT results in a considerable saving in magnetic
field energy in the confinement chamber, and, hence, the magnetic system
and its power supply are substantially simplified.
4.

ANALYSIS OF FRC QUASI-STATIONARY REACTOR PARAMETERS

From the technological point of view, it is essential to study the possibilityof creating an FRC quasi-stationary fusion reactor. To realize a quasi-stationaryburn, it is necessary to recover mass, energy and magnetic flux of the confined
toroid. To realize this process, the authors suggest using the pulsing mode of

IAEA-CN-47/H-H-2

313

fusion burn, realized by injection of compact tori into the confinement chamber
at v> r j j and their merging with the confined torus. As a result, the losses ofmass and main toroid magnetic flux are compensated for. The energy of thetoroid is recovered via the charged fusion products.
Let us assess the conditions in which quasi-stationary self-sustaining fusion
burn is realized, for three fuel types (D-T, DD-Cat, D 3 He).
From energy equilibrium, the following condition is obtained:

where P N is the energy flux due to particle diffusion through the separatrix, P R
are the radiation losses, and P c is the power of energy release in the plasma
by scattering of fast charged particles. From Eq. (1) and the pressure balance
we obtain(11)(12)
where A is the area of the exposed first wall surface, and /3S is the magnetic
field on the separatrix. The power of fusion energy release in the plasma, tothe accuracy of an unessential factor determined by the plasma composition,
may be written as P c = P R ^(T), where </>(T) is a function of the temperature
determined by fuel selection. Combining Eqs (10), (11) and (12) andeliminating parameters, we obtain a criterion for a quasi-stationary selfsustaining fusion burn, anomalous transport process being taken into account:(13)
where q c = PQ/A, and f(T) is a function of the temperature determined byfuel selection. We shall now consider particular types of fuels.
4.1. D-T
In this case, q^ = q n /4, and the neutron load on the first wall exceeds
the bremsstrahlung load (q n > q R ). The function f(T) has the minimum at
T as 8 keV, and Eq. (13) takes the form
(14)

314

BURTSEVetal.

TABLE I. MAIN D-T REACTOR PARAMETERSr w (m)V(m 3 )ne (nT 3 )B (T)T(keV)

2.5

350

3X10 2 0

1.6

8.0

5.0

In this paper, qn is measured in MW-m 2 , and rw is measured in m.
Table I gives the values of the main D-T reactor parameters as derived
from relation (14).
A comparison of these results with estimates made earlier indicates that
inclusion of the LHD influence results in an increase of the reactor dimensions
and the reactor power [1.-3]. At the same time, it is worth noting that themagnetic field in the confinement chamber is not high, compared to that in atokamak, resulting in a simplification of the magnetic system, for example,
the possibility of creating a CT reactor with normally conducting coils.
4.2. DD-Cat
The main advantage of this reaction is known to consist in excluding thenecessity of tritium breeding and storage. The reduction of the reaction crosssection results in a substantial increase in the plasma temperature. In this case,the plasma composition is chosen from a balance of 3He breeding and consumption. This condition determines the relation between the parameters q c , q n ,q R ,i.e.
220q n<3n = JQC> QR = ~^~05)

Optimizing the function f(T) by involving relations (15), we obtain an
optimum plasma temperature of T « 50 keV. Equation (13) may be rewritten
in the form
rwqn/3x!~5(16)

which results in a considerable increase in the system parameters, compared tothose of the D-T reactor (see Table II).
As in the case of D-T fuel, taking LHD into account considerably increases
reactor volume and power. In this case, the reactor dimensions are relatively
reasonable only for large loads on the first wall.

IAEA-CN-47/H-n-2

315

TABLE II. MAIN DD-Cat REACTOR PARAMETERSr w (m)V(m3)ne (nT3)

B (T)

T(keV)

qn(MW-nT2)

3.5

800

6X 1020

5.0

50

5.0

TABLE III. MAIN D3He REACTOR PARAMETERS
^(m)

V(m3)

n e (nT 3 )B (T)T(keV)

qR(MW-nr2)

2.5

350

5X 1020

5.0

40

5.0

4.3. D3He
If. an optimum composition of D and 3He is used as a fuel, the reactor
has smaller dimensions than a DD-Cat reactor. The main difficulty in creating
this reactor consists in the lack of 3He resources on the earth. This is why thisreactor is usually considered to be a staellite to a reactor producing 3He [9].
But, as estimations carried out at the University of Wisconsin, USA, have shown,
there is a large source of relatively easily extractable 3He on the lunar surface.
The energy multiplication, i.e. the ratio of the energy released in a D3He reactor
from combustion of 1 kg of 3He to the energy necessary for 3He mining anddelivery to the earth, amounts to 250, making the idea of using the moon asa source quite acceptable for our goals.
In evaluating the reactor parameters from Eq. (13), it should be borne inmind that a comparatively small portion of fusion energy is carried away byfast neutrons, and the bremsstrahlung load on the first wall becomes the main
parameter determining the reactor dimensions. In this case, the minimizationof the f(T) function results in T o p t « 40 keV, and Eq. (13) has the form

The main parameters of the D3He reactor are given in Table III.
Thus, with optimum D and 3He composition, the reactor volume and themagnetic field are comparable to the corresponding parameters of a tokamak
reactor, tritium breeding being excluded and the requirements for the blanket
being reduced, because of the decrease in fast neutron flux.

316
5.

BURTSEVetal.
REALIZATION OF FUSION BURN PULSING MODE

To study in detail the process of CT quasi-stationary or, more precisely,
pulsing burn, a number of numerical investigations'has been carried out within
the framework of a model which takes into account plasma compression andexpansion dynamics, the process of heating and cooling of its major components,
energy exchange between the components, particle and energy losses due toanomalous diffusion, and thermal conduction. The calculations indicate that
it is possible to realize a D-T fuel burn regime when the oscillations of theneutron load on the wall and the toroid energy do not exceed 15-20%, evenin the case of 'cold' toroid injection (injected plasma temperature equal to'zero'), although the plasma mass reduces by a factor of two during this period(Fig. 2). While using D3He fuel, the bremsstrahlung load becomes the main
contribution, and the amplitude of the oscillations happens to be considerable(«*50%).
The possibility of realizing CT plasma self-heating in the confinement
chamber due to the energy of the fast charged particles is essential, especiallyin operating on DD-Cat and D3He fuel, as the heating of CT up to an initial
temperature of T = 40-50 keV in the formation chamber constitutes a very
difficult problem. From the calculations carried out within the framework of
the above model we learn that these temperatures are attainable for toroid
injection into the confinement chamber with moderate temperature (To ** 4 keV).
In this case, by specially choosing the amount of trapped magnetic flux of theinjected toroids, it is possible to realize plasma self-heating for constant radiationload on the wall. This possibility is very important from the technologicalpoint of view. Figure 3 shows the self-heating process of D-T fuel. Plasma isheated from T«» 4 keV to T ^ 10 keV for six to seven periods for the quasistationary regime to be reached.
The plasma heating in the confinement chamber up to T«* 40-50 keV,which is necessary for the reactor to operate on the alternative fuel, may berealized by initiating D-T plasma burn. Then, with growing temperature, thetritium is substituted by 3He by corresponding programming of the isotope
composition of the injected plasma (Fig. 4). It is necessary to note that the selfheating process may be realized at constant magnitude of the bremsstrahlungload on the first wall, by a programmed change of the relative separatrix radius
radius (x s ). Taking into account that q R « B 4 x | T " 3 / 2 at constant open field
flux we obtain

This relationship establishes the connection between T and x s , where x s isdetermined by choosing the value of the closed field flux of the injected tori.

317

IAEA-CN-47/H-D-2

1

w

^

N
N

0

W_

\ r

• r

0.5 -

0

2
t(s)

F/G. 2. Variation of radiation load on the first wall (qj, energy (wj and particle inventory
in a quasi-stationary burn regime.

10

2

5

0.5

10

FIG. 3.

Variation oftoroid temperature (T) and energy in the regime ofD-Tfuelself-heating.

10,21

10 20

10 19
10

20

30
T (keV)

50

FIG. 4. D, Tand 3He containment necessary for realization of plasma self-heating intoroid plasma.

318

BURTSEVetal.

It is important to note that, energy and particle lifetime of CT beingnearly equal, it is possible to realize a quasi-stationary burn without plasma ash
accumulation. For example, an estimate of the a-particle fraction in the D-T
plasma obtained from Eq. (10) yields

101

N

100

i.e. in the case of r N «=» rg, a-particle content equals 2%, eliminating the problem
of ash removal from the confinement chamber.
Thus, realization of the toroid burn pulsing mode allows the oscillation
amplitude of the load on the first wall to be reduced substantially and therequirements for the toroid injector, due to the plasma self-heating in theconfinement chamber, to be relaxed.

6.

CONCLUSIONSThe calculations carried out indicate that anomalous transport in FRC
due to high beta is no obstacle on the way to the creation of a fusion reactor
although it consitutes an important factor that affects the pertinent prospects.
Further theoretical studies as well as large scale experiments are needed tosolve this problem.
Taking into account anomalous transport, according to present day
knowledge, we have obtained the results that a CT reactor has nearly the same
dimensions as a tokamak reactor at comparable load, but the corresponding
magnetic energy is by an order of magnitude lower.
The quasi-stationary mode of fusion reactor operation may be realizedby pulsing injection and merging of compact tori for particle and magnetic
flux losses to be compensated. The programmed transition from D-T to
D3He fuel makes it possible to develop a reasonably compact and radiologicallysafe fusion reactor (without tritium cycle, with dominant X-ray load on the firstwall). The magnetic fields required for 40-50 keV D3He plasma confinementin FRC are in the range of values characteristic of a D-T fuel tokamak («* 5 T).

REFERENCES[1] MILEY, G.H., Nucl. Instrum. Methods 207 (1983) 111.
[2] HAGENSON, R.L., KRAKOWSKI, R.A., in Fusion Reactor Design and Technology,
Vol. 1, IAEA, Vienna (1983) 377.
[3] MILEY, G.H., Physics and Technology of Compact Toroids (Proc. 5th Symp.),NISNW(1982)69.

IAEA-CN-47/H-H-2

319[4] TUSZEWSKI, M., LINFORD, R.K., Phys. Fluids 25 (1982) 765.
[5] DAVIDSON, R.C., KRALL, N.A., Nucl. Fusion 17 (1977) 1313.
[6] HOFFMAN, A.L., MILROY, R.D., Phys. Fluids 26 (1983) 3170.
[7] SLOUGH, J.T.,etal., Nucl. Fusion 24 (1984) 1537.
[8] KADOMTSEV, B.B., POGUTSE, O.P., SHAFRANOV, D.V., Fiz. Plazmy 7(1981)4.
[9] MILEY, G.H., et al., Lab. Rep. 645-1, University of Illinois, Fusion Studies Laboratory(1981).
[10] WITTENBERG, L.T., SANTARIUS, I.F., KULCHINSKI, G.L., Lunar source of 3He
for commercial fusion power, Fusion Technol. 10 No. 2.

IAEA-CN-47/H-H-3

INITIAL OPERATION OF THE HIGH FIELD
SUPERCONDUCTING TOKAMAK TRIAM-1M
S. ITOH, Y. NAKAMURA, N. HIRAKI, K. NAKAMURA,
A. NAGAO, S. MORIYAMA, S. KAWASAKI, E. JOTAKI
Research Institute for Applied Mechanics,
Kyushu University,
Kasuga, Fukuoka, Japan

Abstract
INITIAL OPERATION OF THE HIGH FIELD SUPERCONDUCTING TOKAMAK TRIAM-1M.
The high field superconducting tokamak TRIAM-1M with Nb3Sn conductors
was the first of its kind in the world and started to operate in June 1986,
after the completion of cooling down and charging tests with the superconducting
toroidal field coils. The superconducting coils were charged to their full operating current,
which produced a peak field of 11 T at the coil windings, and a high average current densityof 5740 A-cm"2 was achieved at the peak field. A high density plasma (n~e = 7 X 1013 cm" 3 )with I p = 150 kA, a = 12 cm, R = 80 cm, and B = 4 T was obtained by optimization of the
feedback control of the plasma positions. The effects of plasma current disruptions on the
superconducting coils were investigated, and the stability of the coils was confirmed intokamak operation. Stable and reliable performance of the superconducting tokamak system,
including the cryogenic system, is demonstrated.

1.

INTRODUCTION

Superconducting magnets are required for the next generation of tokamak
devices, and it is necessary to demonstrate a reliable and stable superconducting
tokamak system, including the cryogenic system [1,2]. We have completed
the high field tokamak TRIAM-1M with an Nb 3 Sn superconducting toroidal
field magnet, which can produce peak fields close to those required for the
projected requirements of the Integrated Test Facility Devices [3].
The machine was installed in the torus hall in March 1986, and the subsystems for TRIAM-1M were connected to the machine. Cooling down of thesuperconducting toroidal field coils started in June 1986, and the superconducting
coils were charged to the full operating current of 6202 A, producing a maximum
field of 11 T. This magnetic field strength constitutes a record among the actual
superconducting tokamak facilities. The first operation of TRIAM-1M wasattempted in parallel with major testing of the various components of thetokamak.
321

322

ITOH et al.

F/G. 1. Artist's impression of TRIAM-1M. Main tokamak components are identified.

The principal objectives of TRIAM-1M are to develop a superconducting
tokamak system and to study the confinement and heating of non-circularhigh field tokamak plasmas. The technical data on the superconducting coils
and the cryogenic system will be collected in the course of the operation.
Impurity dynamics and recycling problems in long pulse operation by lower
hybrid current drive will be investigated. A turbulent heating system with two
pulses is being prepared for additional heating of the tokamak plasma.
The first two months of experimental work were mainly devoted to thestable operation of the superconducting tokamak system, including the cryogenic
system, and to the generation and control of reproducible tokamak plasma
with Ohmic heating.
We report here on the performance of the superconducting tokamak, thefirst results on plasma properties in TRIAM-1M and on the effects of plasmas
current disruptions on the superconducting coils.

2.

THE TRIAM-1M DEVICE

The major components of the tokamak, apart from the diagnostics, areshown in Fig. 1, and the major parameters of TRIAM-1M are listed in Table I.

IAEA-CN-47/H-II-3

323

TABLE I. DEVICE PARAMETERS OF TRIAM-1M
Major radius, R(m)0.8

Limiter radius, a(m)0.12 X 0.18

Plasma elongation

1.8

Overall diameter (m)4Overal height (m)4.2

Toroidal field, B,(T)8

Plasma current, I p (MA)0.5

Discharge duration (s)0.5

2.1. Plasma chamber
The machine consists of a D-shaped plasma chamber, which is made up of
two rigid half torus sectors and two ceramic breaks with bellows. These ceramic
breaks enable us to apply a high loop voltage of 20 kV to tokamak plasmas forturbulent heating. Three conservative D-shaped SUS limiters are fitted inside
the plasma chamber and a pair of movable limiters (up and down) cooled bywater are set up. In addition, a water cooled stainless steel pipe is wound on
the plasma chamber for long pulse operation.
2.2. Poloidal field coil
The poloidal field coils for plasma position and shape control are located inside
the toroidal field coils in order to reduce the heating of the superconducting toroidal
field coils; they consist of a horizontal field coil and two types of vertical fieldcoils, which produce vertical magnetic fields with negative and positive decay
indices. In addition, these poloidal field coils are divided into two half-turncoils to reduce the voltages induced by turbulent heating pulses and are fixed
on the plasma chamber. The turbulent heating coil which consists of fourparallel windings is located just outside the plasma chamber.
2.3. Superconducting toroidal field coil
The toroidal field magnet for TRIAM-1M consists of 16 D-shaped superconducting coils, the parameters of which are shown in Table II. An Nb 3 Sn
conductor was chosen for the toroidal field coils because of a greater energy
margin for stability against thermal disturbances at higher fields. There are
two key features for cryostability of the superconductor: One is strong thermal

324

ITOH et al.
TABLE II. PARAMETERS OF TOROIDAL FIELD COILS
Number of coils

16

Major radius, (m)0.85

Inner diameter of bore (m X m)

0.44 X 0.67

Outer diameter (m X m)

1.035 X 1.25

Magnetic field (R = 0.8 m) (T)8

Peak field at the conductor (T)11Stored energy (MJ)

16

Conductor material

Nb3Sn
6202

Conductor current (A)Cooling

Pool boilingby 4.2 liquid He

coupling of the conductor to the helium through the rough surface which has
a heat flux of 0.82 W-crrT2, and the other one is the use of a high purity
aluminium stabilizer which has low magnetoresistance and serves as a current
carrier in the case of creation of a local normal zone. The conductor is wound
into double pancakes; six double pancakes with 2 mm thick spacers for thecooling channels are stacked as a coil. The superconducting coils are housedin stainless steel cases and cooled by pool boiling helium. Qualifying tests
involving charging tests of full operating current, recovery tests from a halfturn driven normal and measurements of AC losses due to pulsed fields havebeen carried out for the superconducting coil, and the stability of the toroidal
field coils in tokamak operation has been confirmed [4]. These superconducting
coils are assembled into the toroidal field magnet with two supporting bases,
which are located at the top and bottom of the coils and resist centring forces.
Further support is provided by shear panels which join the coils in pairs ofeight sub-assembly units at the outer portion of the coils. The coils andsupporting structure, both operating at 4.5 K, are housed in a single bell jar
shaped vacuum vessel of 4 m diameter. The cold mass of the toroidal fieldmagnet is about 3 X 104 kg. A thermal radiation shield, which is cooled withliquid nitrogen and wrapped with multilayer superinsulators, is interposed
between the toroidal field coils and the room temperature boundary of thevacuum region.
The sixteen toroidal field coils are electrically connected in series with adump resistor which is used to dissipate the stored energy in case of coilquenching. The maximum voltage during a dump is limited to 1 kV.

IAEA-CN-47/H-H-3

325

COMPRESSOR

FIG. 2. Refrigerator and superconducting magnet system.

3.

CRYOGENIC SYSTEM

A schematic flow diagram for the cryogenic system is shown in Fig. 2.
Cold helium is circulated from the refrigerator to the toroidal field magnet.
Helium gas which has cooled power leads is returned to the compressor suction.
The helium refrigerator consists of three major components: the compressor,
the purifier and the cold box. The capacity of the plant was determined by steadyheat rates of the toroidal field magnet and pulsed magnetic field losses in tokamak
operation. The refrigerator performance during acceptance tests demonstrated acapacity of 460 W at 4.5 K and liquefaction rates of up to 210 L-rT 1 wereobtained. Transfer lines of about 9 m from the refrigerator to the machine areused for both the cold helium supply and the cold return.
The cold helium is distributed to 16 toroidal field coils and to two coldbases through the valves, which control the helium flow rate so as not to cause
large temperature differences between the coils. Each helium vessel is linked
to a liquid helium reservoir tank which is located above the coils, and the liquid
level is controlled so as to be the vertical location of the power leads. The
temperature of the power leads is controlled by the helium flow rate throughthe valves which are placed in the downstream line.

ITOH et al.

326• Supply Gas Temp.
- Helium Vessel Bottom Temp. ( TEP T74 )• Helium Vessel Center Temp.t TEP T75)
o Helium Vessel Top Temp. ( TEP T76.TEA T77 )• Helium Flow Rale

350

35

Start ofPrecooling

5300 oooooooJo

30 ~

2 250

2 5 »—

UJ
DC

S

i

in

»2

UJ

200 Cooling of " R
Radiationo

20 2Starting of
Expander Turbine

Shield15 EL

150Transfer of

10100. • • • • • ! • : %

5

50

-24

25

50

75
TIME ( h )

100

125

150

0

FIG. 3. Cooling curves for superconducting toroidal field coil.

4.

COOLING DOWN AND CHARGING TESTS OF
SUPERCONDUCTING COIL

4.1. Cooling down of superconducting magnet system
Before cooldown from the ambient temperature, the complete helium
circuit is initially purged and dried to a dew point of-60°C or lower; in addition,
the vacuum tank in which the whole tokamak system is housed is evacuateddown to 10~s torr. Cooling curves for toroidal field coils are shown in Fig. 3.
After the radiation shields have been cooled by liquid nitrogen for 24 hours,
precooling is performed, first with helium of an initial temperature of about
80 K. This period lasts for about 75 hours. The expander turbine is thenswitched on, and the initial temperature is lowered. The total precooling time
is about 125 h. At the end of precooling, liouid helium is collected in the dewar
and then transferred to the coils. The magnet system is cooled to 4.2 K in sevendays without large temperature differences between coils or between the coils
and the cold bases. The cryogenic system performance is in accordance withthe calculated values. The measured steady heat rate to this magnet system isapproximately 180W.

327

IAEA-CN-47/H-II-3

6

8

MAGNETIC FIELD

10

12

14

(T)

FIG. 4. Load line for superconducting toroidal field magnet. Dashed and solid lines indicateload line for single coil and coils in full torus, respectively. Ic is critical current.

4.2. Charging tests of superconducting coils
After cooling down the toroidal field coils and filling up with liquid helium,
the coils were initially charged up to 3100 A. Plasma experiments were carriedout with the corresponding magnetic field, and then the coils were operated withno quench at currents up to 6202 A, which was the full operating current andproduced a maximum field of 11 T at the coil windings. The load line for thecoils at the toroidal field magnet is shown in Fig. 4, together with that for a single
coil. As a result, the highest average current density of 5740 A-cm" 2 in theworld was attained, including cooling channels and insulators.
While the coils were being charged, very small voltage pulses (~1 mV) wereobserved. It is assumed that the jumps were inductive pulses due to wire
movements under the magnetic loads. Mechanical stresses were measured by
the strain gauges which were fitted on the outside of the coil casings. The
maximum stress was about 35 kg-mm" 2 and less than the design value estimatedfrom finite element calculations.

5.

FIRST OPERATION OF TRIAM-1M

The experimental plan of TRIAM-1M is divided into two stages. The firststage of the experiment is limited by the capability of the power supply for thepoloidal field coils. The basic objective of this stage is to produce Joule heated

328

ITOH et al.

(a)
(b)

12

Bt= 4T6

P = 8xia5torr (Hi)

I °

'E

-4
120

\c

90

1
"*

60

30
0
10

?

0

I -io
-20
20
10
20

0
-10

20

40
Urns)

60

40
t(ms)

80

FIG. 5. (a) Typical waveforms of loop voltage, F L , plasma current, I horizontal position, AH,
and vertical position, A v . (bj Time evolution of line averaged density, central electron andion temperatures.

plasmas of about 100 kA in the high field superconducting tokamak and toestablish a necessary basis for the plasma control capability, turbulent heating
and RF current drive for the next stage.
5.1. First results in high field superconducting tokamak
The first operation was attempted in parallel with the integrated tests ofboth the toroidal and poloidal field systems. After ECR discharge cleaning(2.45 GHz, 5 kW) and pulse discharge cleaning had been carried out, we
obtained a hydrogen discharge as shown in Fig. 5 (a). A plasma peak currentof 115 kA was reached in a pulse of 70 ms duration. We have no shell, and329

IAEA-CN-47/H-II-3

P = 1.4x10* torr(H 2 )

10

20

30

40

TIME (ms)FIG. 6. Typical waveform of plasma current with disruption. Plasma current drops at rateof 25 kA-ms~l.
400

z
-

350

UJ
X

5

300

a

•

_j

Li.

1

•

Suspension of
the discharge

250

•

O
LJ
3:

•

200

.

f

•: )

20

40
60
TIME (min)80

FIG. 7. Time variation of liquid helium level in reservoir tank in suspension of discharges,
repeated every three minutes.

the time constant of the plasma chamber is about 1 ms. Accordingly, this formof operation was achieved with good horizontal and vertical position controls,
which were performed by a hybrid digital/analog system and analog circuits
to control the power supplies, respectively. Time evolutions of the line averaged
electron density, the electron temperature and the ion temperature are shownin Fig. 5(b). This discharge was done at low hydrogen pressure (8 X 1CT5 torr),

330

ITOH et al.

and there was no additional gas injection. However, we obtained a high density
plasma which had a peak density of 5 X 10 13 cm"3 and Zeff of about 1.7.
It is assumed that hydrogen gas filled into a number of long diagnostic ports
played the role of continuous gas injection. The time evolutions shown in
Figs 5 (a) and (b) were measured in the same discharge, except for the electron
temperature. The maximum values obtained in the different discharges are as
follows: Ip = 150 kA, n e = 7 X 10 13 cm" 3 , T e0 = 530 eV, Ti0 = 220 eV at
Bt = 4 T.
5.2. Stability of superconducting coil in tokamak operation
When we designed the superconducting toroidal field coil for TRIAM-1M,
we analysed the stability of the coil against magnetic field variations in tokamak
operation with a plasma current of 500 kA and confirmed its stability in case ofplasma current disruptions [4]. During this experimental stage, the plasma
current is limited to a maximum of 150 kA by the power supplies; as a result,the stability of the coil against the field variations should be guaranteed. In fact,
we have no experience in the creation of a normal zone during tokamak operation,
including plasma disruptions. In addition, the pulsed field loss due to plasma
disruptions is investigated. When the plasma disruptions with I p « 100 kA as
shown in Fig. 6 occur in series and the machine stops, the constant liquid helium
level increases by reduction of the heat rate to the superconducting magnet
system as is shown in Fig. 7. If the valves for liquid level control are left as they
are, we can estimate the pulsed field loss due to plasma disruptions from the rateof increase of the liquid level. The measured loss is about 7 kJ, which iscomparable to the calculated value. Effects of higher plasma current on the
superconducting coils will be investigated in the next stage, where the power-upof power supplies for poloidal field coils will be carried out.

6.

CONCLUSIONS

A superconducting toroidal magnet with Nb 3 Sn conductors has beendeveloped, and a peak field of 11 T has been achieved with a high average current
density of 5740 A-cm" 2 . A high density plasma has been produced by good
plasma position controls. The successful completion of the first operation
demonstrates that the major systems of the high field superconducting tokamak
TRIAM-1M have operated effectively, as a whole. In addition, the developmentof high field and high current density superconducting coils and the stable andreliable operation of the superconducting tokamak system will encourage designand construction of the next generation tokamaks.

IAEA-CN-47/H-II-3

331

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

AYMAR, R., et al., IEEE Trans. Magn. 17 5(1981)1911.
KOSTENKO, A.I., et al., IEEE Trans. Magn. 17 5 (1981)2226.
INTOR GROUP, International Tokamak Reactor, Zero Phase (Rep. Int. Workshop
Vienna, 1979), IAEA, Vienna (1980).
ITOH, S., et al., in Controlled Fusion and Plasma Physics (Proc. 11 th Europ. Conf.
Aachen, 1983), Vol. 7D, Part I, European Physical Society (1983).

DISCUSSION
B. COPPI: Your results are an important milestone in the area of high field
confinement. In fact, the TRIAM-1 toroidal magnet can be used to produce
currents of about 1.5 MA with record high current densities. In the long term,
the high field technology you have developed will allow us to begin exploring
the burning conditions for D-D and D- 3 He mixtures.

IAEA-CN-47/H-ID-2

MFTF-B ACCEPTANCE TESTS AND OPERATION*
F.H. COENSGEN, J.W. GERICH, P.M. HOLL, V.P. KARPENKO,
T.A. KOZMAN, K.H. KRAUSE, W.C. LABIAK, D.D. LANG,
G.H. LATHROP, W.B. LINDQUIST, R.L. NELSON,
R.L. PETERSON, D.L. PODESTA, J.R. REED, P.R. RUPERT,
D.W. SHIMER, P.D. SIEMENS, W.H. STERBENTZ, R.R. STONE,
J.S. SULLIVAN, A.L. THROOP, J.H. VanSANT, J.P. ZBASNIK
Lawrence Livermore National Laboratory,
University of California,
Livermore, California
B. HUFF, C.A. LEAVITTTRW Inc.,
Redondo Beach, California
United States of America
Abstract
MFTF-B ACCEPTANCE TESTS AND OPERATION.
Final acceptance tests of the Mirror Fusion Test Facility (MFTF-B) were completed in February
1986. These tests verified the performance of the following subsystems: the magnet system, the vacuum
system and vessel, cryogenic systems, 80 keV neutral beam sources and power supplies, microwave
power systems for plasma heating, the supervisory control and diagnostic system, and the local control
and instrumentation system. The entire magnet system was operated at full field continuously for
24 hours. The largest field alignment error under full load, determined using an electron beam technique, was 6 mm, well within the required precision of 15 mm. Absolute values of the magnetic field
were determined using nuclear magnetic resonance techniques and were found to be within 2% of thepredicted values. All magnetic protection systems and fault protection systems were tested at full load.
At 4.35 K, the cooling capacity of the liquid helium system exceeded the 11 kW rating and the nitrogen
reliquefier met its 500 kW rating.

1.

INTRODUCTION

The Mirror Fusion Test Facility (MFTF-B) construction
project was successfully completed February 28, 1986, withconclusion of the acceptance test [1], The magnetic field
configuration required for thermal-barrier tandem-mirror plasma
containment was achieved by simultaneous operation of thefollowing five major systems: the vacuum vessel, the external* Work performed under the auspices of the United States Department of Energy by the Lawrence
Livermore National Laboratory, under Contract No. W-7405-ENG-48.

333

334

COENSGEN et al.

IAEA-CN-47/H-HI-2

335

vacuum system, the cryogenic system, the magnet system, and theintegrated control system. Operation of two neutral-beam
injection systems and one electron cyclotron resonant heating(ECRH) system was demonstrated in a full magnetic fieldenvironment. Magnetic field alignment, magnetic field profile,and mechanical strains were determined at full field conditions.
Vacuum parameters, eryosystem performance, and magnet powersystem parameters were monitored continuously during the fivemonth test.
2.

THE MFTF-B VACUUM VESSEL

Positions of the plasma heating systems (ECRH gyrotrons,
ion cyclotron resonant heating (ICRH) antennas, and neutral
beams) with respect to the MFTF-B magnet system and vacuum
chamber are indicated in Fig. 1. The vacuum vessel system [2]shown in Fig. 1 provides an appropriate vacuum environment forthe plasma and provides thermal insulation for the magnet
system. Neutral beam sources, beam lines, and diagnostics aremounted externally on the vessel. Cryopanels, magnets, ECRH
antennas, ICRH antennas, beam dumps, plasma dumps, anddiagnostic sensors are mounted within the vacuum vessel. The
vessel has three main sections: two end sections, each 17.5 mlong and 10.6 m in diameter, and a center section 20.5 m longand 8 m in diameter. These are connected by two 1.2 m longtransitions to form a vessel that is 58 m in length and 1300 m
in volume and contains about 1.1 x 10 kg of type 304 stainlesssteel. The vessel supports the weight of the magnet system,
6
5
about 1.2 x 10 kg, and approximately 5.5 « 10 kg of otherequipment. In addition, it supports longitudinal magnetic
forces, forces associated with energizing the magnets,
atmospheric pressure loads, and seismic loads.
The vessel pressure is reduced from atmospheric pressure-2
to 1.0 * 10Torr by four pump t r a i n s , each having a f i r s t stage 385-L/s Roots blower, a second-stage 283~L/s Roots blower,
and a 142-L/s mechanical pump. Below 50 Torr the t o t a l pumping
speed of the roughing system i s 1000 L/s. Ten high-vacuum
cryopumps are used to further reduce the vessel pressure to1 x 10Torr. Four of the cryopumps are two-stage
cryocondensing/cryosorption pumps in which helium gas i s reduced-9t o a p a r t i a l pressure of 3 x 10
Torr by a sorption process ona panel t h a t i s continuously sprayed with argon gas. Measured
pumping speeds for He are 1.1 x 10 L/s per pump. Four turbomolecular pumps in a leak-check s t a t i o n provide an additional
16 x 10 L/s helium pumping capacity. At 10
Torr, LHe at
4.5 K i s introduced into the internal cryopanels to reduce the336

COENSGEN et al.

b a s e p r e s s u r e t o 1 x 10
Torr.
There a r e aboutc o o l e d c r y o p a n e l s t h a t , b a s e d on a c c e p t a n c e t e s tprovide a pumping speed of 67 * 103.

10mof LHemeasurements,

L/s for deuterium gas.

THE CRYOSYSTEM

The cryosystem provides liquid helium to thesuperconducting magnets, the cryopanels, the cryocondensing
vacuum pump, and the cryosorption vacuum pumps. These units
operate at 4.5 K and are shielded by liquid-nitrogen-cooled
baffles at 85 K. About 6.6 x 10 kilograms of materials, mostly
various metals, are cooled to these temperatures. Liquid helium
is supplied by two independent closed-loop systems (3 kW and
8 kW), which provided 11,000 watts of cooling at 4.35 K during
the acceptance tests. When operated as liquifiers, thesesystems produce 1700 L of liquid helium per hour. During
maintenance and repair periods the magnets were maintained at
LHe temperature by two large storage Dewars (25 and 60
thousand L) and 13 small individual magnet Dewars. Boiled-off
gas vents into a gas storage bag. During warmup of the system,
gaseous helium circulates through heaters (about 250 kW output)
instead of the liquid nitrogen heat exchangers that are usedduring system cooldown. The warm helium gas passes through thecold-box heat exchangers to bring the magnet system temperaturefrom 4.5 K to room temperature in about 48 hours. Emergency
shutdown of the magnets was tested during the acceptance test.
In such shutdowns, the supply Dewar is automatically isolated
and vent valves on the magnet return lines open to release cold
helium gas into the atmosphere. During operation, special
magnet-quench instrumentation automatically initiates the
emergency shutdown cycle when loss of superconductivity isdetected in a magnet.
4.

THE MAGNET SYSTEM

The MFTF-B magnet array and the on-axis magnetic field
profile used in the acceptance test are shown in Fig. 2. At
each end, two transition coils (T1 and T2) are needed between
the end cell formed by magnets M1 and M2 (i.e., the yin-yangpair) and the circular central-cell magnets to map theelliptical magnetic flux bundle leaving the end cell to circularcross-section in the central-cell region. The 24 coils in thelinear array provide the field environment necessary for tandemmirror plasma physics investigations. Eight rectangular trim
coils (TR coils) are positioned around the transition coils tocorrect possible field errors. Table I lists the importantmagnet system parameters. All coils (except A2i) areconstructed of filamented NbTi superconductor (up to 512

337

IAEA-CN-47/H-OI-2
B p = peak field location

(a)

East

TR

West

(b)

"25

-20

-15

-10

-5
0
5
Axial distance, 2 (m)

10

15

20

25

FIG. 2. (a) Magnet array for MFTF-B. (b) On-axis MFTF-B magnetic field profile, Feb. 1986.

filaments in the M1 and M2 windings) and are designed to beunconditionally stable. Coil A2i is a Nb.Sn magnet inserted
into coil A2o to produce the 12T magnetic field on axis. There
are 145,000 Nb,Sn filaments in the cross-section of the A2i
conductor. Coils are cooled to 4.5 K by natural convective flowof liquid helium into the bottom of each magnet through theconductor bundle to the top of a Dewar. The C-shaped coils M1,
M2, T1, and T2 are oriented at 45° to the gravitational force tofacilitate liquid helium flow.
Each of the 40 coils is contained in a vacuum-tightstructural case of nitrogen-strengthened stainless steel,
304 LN. Case thicknesses up to 127 mm limit stress below thepermissible 552 MPa (80,000 psi). Spreading forces of thetransition magnets are restrained by structural frames attachedto the outside of the large-radius lobes. The C-shaped magnets
M1 and M2 are interconnected to provide mutual restraint of thelobe-spreading forces. The combined magnet system and its
support weighs about 1.2 x 10

kg (1350 tons).

COENSGEN et al.

338

TABLE I. IMPORTANT SYSTEM PARAMETERS FOR
MFTF-B MAGNET SYSTEM

Coil

Major/
minor
or meanradius
(m)

M2
M1
T2
T1
A2o

2.5/0.75
2.5/0.75
1.20/1.375
0.875/0.70
0.928

Peak

Weight/ field in
conductor
coil
3
(T)
(x10 kg)(MJ)

Maximui
operaticurren
(A)

23

5.0

113
88

49.5
26.8
30.8

5.3

58

33

5.5

14

28

7.5

77

24

7.0

12.52

12

20

28.0
22.4

7.7

77

22

28.3
28.3
0.186

31
31

4409
3837
6278
5266
4648
1504
4238
2866
2866

35

900

0.113

35

900

182

5.8

182

A1

0.3883
0.928

S6

2.5

S5-S1
TR1-4

2.5
-

5.4

3.24
3.36
2.88

TR5-8

—

5.4

2.2

A21

StoredenergyWindingpackcurrent
density2(A/mm )

14.2

20

Calculation of radial transport of plasma in MFTF-B
indicated that the rate was extremely sensitive to the geodesic
curvature of the magnetic field in the transition between thehigh field coil A2 and the end-cell mirror coil M1. The
magnetic design minimized this curvature by providing opposing
curvatures in this region. However, further analysis [3] showedthat misalignment of critical magnets by a few millimeters
reduced the geodesic curvature compensation, leading topredicted unacceptably large radial transport. Trim coils areplaced in the transition regions to bring the alignment
requirements within practical limits. Individual magnets in thelinear array were placed as much as 2 cm from the desired final
position when installed at room temperature so that they would
be in correct positions when cooled to 4.5 K and energized. To
check the alignment at operating conditions, electron beam

IAEA-CN-47/H-III-2

339

measurements were made in vacuum with the entire magnet systemenergized. Displacement errors were found to be 3 mm or lessand rotational errors 0.22° or less, both well within the±15 mm and +0.6° correction capability of the trim coils.
Absolute values of the magnetic field, determined by using
nuclear magnetic resonance techniques, were found to be within
2% of predicted values. The measured values are indicated inFig. 2b.
5.

PLASMA HEATING SYSTEMS

Injection of intense beams of neutral particles isrequired to create the thermal-barrier plasma configuration in
MFTF-B. Figure 1 shows the locations of these beams. The beams
in MFTF-B are produced in sources having a four-grid accel-decel
extraction system with an 80 kV accelerating voltage. To
produce the thermal-barrier plasmas, 16 sources operate for
0.5 s and 6 operate for 30 s. In the acceptance tests, two
0.5 s neutral beams were mounted side by side as shown in Fig. 1,
on the right-hand end tank. These beams were operated for 0.5 ssimultaneously at 75 kV and 72 A in the presence of the fullmagnetic field. The magnetic shields were found to be adequate
to permit normal arc chamber performance and to prevent beamloss in the neutralizers. Also, the electromagnetic shielding
was adequate to prevent "crosstalk" between the beams wheneither of them interrupted. Measured pressure rise within theplasma region in the end cell due to the gas feed in the beamarc chambers was 6 x 10
Torr as calculated for beam operation
without cryopumping.
The MFTF-B facility includes 24 neutral-beam power
supplies, each capable of 88 ampere dc output at 80 kV. These
supplies draw power from a dedicated 250 M M substation. In
addition to the 80 kV accel supply, each beam is provided with a
5500 A, 20 V filament supply and a 4000 A, 70 V arc supply. In
the acceptance test, six of these systems were connected todummy loads and powered simultaneously from a common feeder todetermine whether or not the feeder terminal voltage drop waswithin acceptable limits. In the quadrant test, we found thatthe regulation on the 230-kV substation was adequate, all output
voltages and currents were as specified, and under operating
conditions that there were no interactions among the six sets.
Potential profiles in the thermal-barrier plasmas arecontrolled in part by locally heating electrons. Electron
heating in the thermal-barrier and potential maximum regions isprovided by eight gyrotrons, two operating at 28 GHz, two at
35 GHz, and four at 56 GHz. Each gyrotron is rated at 200 kW.
These eight systems are driven by one of the 80 kV neutral-beam
power supplies. The ECRH control system permits individual,
simultaneous, or staggered operation of the eight gyrotrons.

340

COENSGEN et al.

This flexibility of operation permits the spatial and temporalvariation of ECRH application needed for the wide range of
MFTF-B plasma physics experiments. One gyrotron at eachfrequency has been tested at full power. Also, a 28 GHz
gyrotron was tested in operating position in the presence of thefull field at both power and full duty cycle, to verify theadequacy of the magnetic shielding.
6.

THE INTEGRATED CONTROL AND DIAGNOSTICS SYSTEM

During the acceptance tests, all systems, with one
exception, were operated through the Integrated Control and
Diagnostics System (ICADS). The exception, the 3 kW helium
cryoplant, had been operated in a previous test through a
computer control system. ICADS includes a supervisory control
and diagnostic system (SCDS) made up of a network of nine
Perkin-Elmer 3200 series computers connected through a sharedmemory together with seven graphics-oriented operator consoles
and a network of three SUN Microsystems workstations for plasma
diagnostics operation. ICDS also provides for detailed andlocal control of hardware through approximately 70 Digital
Equipment Corp. LSI-II computers, numerous CAMAC modules, andseveral Gould Modicon controllers. SCDS provides the manmachine interface, data processing, and data storage facilities.
The local control computers provide a standard interface between
SCDS and the MFTF hardware being controlled. The hardware andsoftware of SCDS have been designed so that any piece ofhardware in MFTF can be operated from any console and so thatthe data from the entire system are available to any operator.
SCDS has been designed to tolerate the failure of any single
computer.
The personnel safety and interlock system is based on two
Gould Modicon 584 Programmable Controllers. It is functionallya stand-alone system that is monitored by the computer control
system so that safety status can be displayed on the consoles in
the control room. The safety system handles the functions of
personnel warning with audible and visual warning devices,radiation monitoring, oxygen deficiency detection, and fire
alarm monitoring. The system Integrates the personnel safety
features built into each MFTF-B system with global features suchas interlocks, access control, and warning systems for thevault.
ICADS was used throughout the acceptance test with very
few problems. None of the failures affected the availability ofthe overall system, and all failures were rapidly and easily
repaired. Central processor (CPU) utilization measurements
showed sufficient reserve capacity to handle error conditions or
CPU failure with operation in backup mode. Significant software
enhancements were made during the test in response to requests
by the operations staff, with negligible impact on operations.

IAEA-CN-47/H-IH-2

341

Throughout the tests ICADS met or exceeded the requirements
placed on it.
7.

CONCLUSIONS

Operation of the MFTF-B magnet set at full field has
demonstrated the feasibility of designing, constructing, andoperating large-scale complex magnet systems suitable foradvanced fusion devices. The successful MFTF-B acceptance test
clearly advanced not only superconducting magnet technology but
also the technologies of large-scale vacuum and cryogenic
systems and computer-controlled operation of large-scale
facilities.
For further details of MFTF systems see the Proceedings ofthe Eleventh Symposium of Fusion Engineering

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

KRAUSE, K.H., KOZMAN, T.A., MFTF-B Pace Acceptance Test and Project Final Cost
Report, Rep. UCID-20819, Lawrence Livermore National Laboratory, Livermore, CA (1986).
GERICH, J.W., The design, construction and testing of the vacuum vessel for the tandem
mirror fusion test facility, J. Vac. Sci. Technol., A 4 3 (May/June 1986).
HOOPER, E.B., Jr., Magnetic field errors in quadrupole tandem mirrors, Nucl. Technol.
Fusion 3 (Jan. 1983).
Fusion Power Engineering (Proc. 11th Symp. Austin, TX, 1985), IEEE, New York (1986),
Cat. No. CH2251-7.

IAEA-CN-47/H-IU-4ADVANCED ICF REACTOR TAKANAWA-I
WITH LOW RADIOACTIVITY
AND TRITIUM SELF-BREEDING IN THE PELLET
T. TAZIMA
Institute of Plasma Physics,
Nagoya University, Nagoya
S. IDO, T. YABE
Institute of Laser Engineering,Osaka University, Suita, Osaka
T. SHIBA, H. NAKASHIMA
Department of Energy Conversion Engineering,
Kyushu University, Fukuoka
M. NISHIKAWA
Department of Nuclear Engineering,
Kyushu University, Fukuoka
N. ASAMI
Advanced Nuclear Systems Engineering Department,
Mitsubishi Heavy Industries, Ltd, Minato-ku, Tokyo
Japan
Abstract
ADVANCED ICF REACTOR TAKANAWA-I WITH LOW RADIOACTIVITY AND TRITIUM
SELF-BREEDING IN THE PELLET.
To enhance the merits of inertial confinement fusion plants from the viewpoint of fueland material resources, energy cost, safety, and required technologies, the authors propose anadvanced ICF reactor concept, TAKANAWA-I. A D-T ignitor and a deuterium fuel pellet
expecting tritium self-breeding in the pellet and liquid lead curtains as energy and shock
absorbers for protecting a SiC shield and the reactor vessel are employed.

1.

INTRODUCTION

Inertial confinement fusion (ICF) reactors do not need magnetic coils forplasma confinement; they use a high vacuum of 10" 8 -10~ 9 torr. Therefore,a simple and compact reactor system is to be expected; no too serious material
problems are anticipated by employing liquid metal walls. It is, however, still
difficult to meet the requirements of social acceptance from the viewpoint offuel and material resource restriction, energy cost, and safety. For example,
343

344

TAZIMA et al.
Pb Inlet
VacuumPumpingPort

Cooling Gas
OutletLiquid•Pb

Curtain

Cooling Gas
Inlet

Pb Outlet

5(m)FIG. 1. TAKANA WA-I reactor concept.

to use liquid lithium for tritium breeding, as primary coolant and for first wall
protection, we shall have to face an enormous energy cost from the extractionof lithium from the seawater; remote maintenance will be difficult since thelithium flow shows good compatibility only with a stainless steel vessel which
has a long decay time for its high induced radioactivity. Therefore, we shouldfind a reactor concept without Li blanket, using materials whose induced
radioactivity decays rapidly. The advanced ICF reactor TAKANAWA-I [ 1 ](see Fig. 1 and Table I) is such a concept; it will be discussed in the following.

2.

D-T IGNITOR AND DEUTERIUM FUEL PELLET

We employ a D-T ignitor and deuterium fuel pellets [2, 3], expecting
self-sufficient tritium breeding in the pellet in order to eliminate the lithium
blanket; we use another material, for example lead, for the liquid metal wall.

345

IAEA-CN-47/H-DI-4

TABLE I. MAJOR PARAMETERS OF TAKANAWA-I
DT/D/ 3 He P e l l e t

DT/D Pellet
Total Thermal Output (MW)3597

Fusion Power From The Pellet (MW)

2658

3145
'ulOOONet Electric Power (MW)1

Repetition Rate (Hz)>v30

3eam Energy (MJ)3e l l e t Density (x l i q u i d density)

2500

I g n i t o r Final Radius (pm)100

D Fuel Final Radius (ym)

300
1.4

Internal Fuel Energy (MJ)

102

Pellet Gain

89

0 Burn-up (%)

22.1

24.6

teutron Energy (MW)

652

558

T Breeding Ratio

1.12

1.39

He Breeding Ratio

—

1.13

1.18/0.79/40.22
0.06/3.24
1.64
30.92
1.64
3.71
5.94

I g n i t o r AmountCJ

1.25/-/Fuel Amounti D

0.83/43. 39(ing/shot)
Exhaust Amount(mg/shot)

T Inventory(g)

Pb Coolant

3

l He
f

l

D
-. H

-/1.40
34.47
1.24

3

He

1.80

L4He

4.77

'Fuel Cycle

V80

182

Coolant

0.1

0.1

Structure

1212In-Site00

Off-Site

None

None

.Total

192.1

194.1

("Flow Rate (ton/s)

600•< I n l e t Temperature (° C)

505

Outlet Temperature ( °C)

550

IPumping Power (MW)

80

346

TAZIMA et al.
poRo
5(g/cm 2 )
10

15

1.5

JIgnitor/100pm
I D Fuel/300pm

•S

8.
2

'5000
1.0
S 3000

3
1000

1000
Density

2000
3000
p (x Liquid Density)

0.5

FIG. 2. Relations among fusion yield, tritium breeding ration (TBR), and PD.RD in case of
D-T/D pellet. Initial conditions: constant density = 9.6 X 102s cm'3, ion temperature =electron temperature = 5 keVfpR<1.0g-cm'2),
0.1
keV(pR>1.0g-cm'2j.

When a PrjRo value of 10-15 g-cm 2 (pj) and RJJ are the density and thefinal radius of the deuterium fuel at the end of compression phase) is attainedby effective compression, once the D-T ignitor is heated up to several keV, thea-particles will heat the ignitor itself, and the neutrons will heat the deuterium
major fuel surrounding the ignitor up to several hundreds of keV. The bulk of
the tritium produced through the D-D reaction will burn during explosion;
however, some amount of unburnt tritium is used as a fuel for the D-T ignitor.
The 1-D hydro code MEDUSA-IB including a time dependent neutrontransport routine is used to simulate such effects. In the calculation, charged
particles and recoil ions from neutrons are assumed to deposit their energy
locally and instantaneously. Bremsstrahlung radiation is also assumed to escapefreely from the pellet.
Figure 2 shows relationships between pR, fusion yield, and tritium breeding
ratio (TBR). The case of a compression of 2500 times satisfies the requirementsof 2500 to 3000 MJ fusion output, a TBR higher than unity, and a possible pellet
gain. The detailed burn characteristics of the above-mentioned case and the caseof 5% 3He concentration are listed in Table I [4]. No realistic driver system orpellet compression scenario has been found so far; this is among the most difficultand as yet unsolved problems in ICF research. In our case, on the assumption of
a 5% hydrodynamic efficiency, a high driver efficiency of 20% is required, becauseof the low pellet gain (e.g. ~100). An intense pulse power generator system is oneof the candidates for a driver system. Two possible scenarios for pellet compression
have been proposed: 'radiation conversion of light ion beams' [5] to reduce theIAEA-CN-47/H-HI-4

347

number of beam lines and to mitigate beam focusing, and 'indirect REB irradiation'[6] with direct power transport through a micro, self-magnetically insulatedtransmission line. However, these scenarios are in their infant stages and hence
left to further studies.

3.

REACTOR CAVITY

Liquid lead multilayer curtains at a mean velocity of 10 m-s" 1 are employed
as primary coolant and for first wall protection. The melting point of lead isrelatively low; lead shows a rapid decay of induced radioactivity and good compatibility with other materials. SiC is employed as a structural material for shieldand vessel, because of the rapid decay of induced radioactivity. At the present
state of the art, SiC is not an adequate structural material because it is brittle anddifficult to weld. As fine ceramic research is now rapidly progressing, we might
be justified in predicting optimistically that it will be usable in future. In theshield, liquid lead flows at a velocity of 0.5 m-s" 1 to remove the heat and toshield the 7-rays.
The results of the neutronics calculations are shown in Fig. 3, obtained byusing the energy spectrum of emerging neutrons from the pellet. A remarkable
result is that the incident neutron energy of 3.7 MeV/neutron (the total energy
is 652 MJ/shot) is multiplied in the lead curtains and in the lead flows in theshield by the (n,7) reaction up to 9 MeV/neutron (total: 1591 MJ/shot). The
lead curtain of 75 cm thickness is required to reduce the neutron wall loadingat the shield surface, in order to secure a long life, down to 0.5 MW-m"2 (i.e.
10 MW-a-m" 2 for 30 years' operation with 70% availability). This entails theneed for an intense lead mass flow which requires high pumping power. The
neutron dose rate of 1.6 X 10~2 mrem-h" 1 is sufficiently low at the outersurface of a heavy concrete layer of 1.5 m thickness used as a biological shield.
Figure 4 shows the total induced radioactivity per unit of thermal power.
The radioactivity level at shutdown is of the same order of magnitude as was
obtained in previous reactor studies [7, 8]; however, the level reduces rapidly,
by two orders of magnitude, during one day, and by six orders of magnitude
during one year [9]. When the entire amount of liquid lead is removed fromthe reactor cavity after shutdown, relatively easy reactor maintenance can be
expected within several days, because of the low radioactivity of SiC.

4.

FUEL CYCLE

Figure 5 is a schematic diagram of the fuel cycle. The flow rates in g- d"1
are shown for the case of injection of a D-T/D/ 3 He pellet. No tritium storageis provided in this fuel cycle. Tritium extraction from tritiated hydrocarbon or

348

TAZIMA et al.
Heavy Concrete
B4C Absorber
SiC Vessel/
VacuumShield(SiC 507.. Pb 507.)^
Vac urn—**rw«s5i.j!«v\ Mainlenance \VacuumPellet"0

Dose Rate(mrem/h)Neutron Wall Loading(MW/m2)r-Ray Wall Loading(MW/m2)

§0

^20

200

550 6 $ 62O"ij25

8J2Q

9_5_Q (cm)1.57xl016

7.96x10"1.77xi69

3.88xl'o7

2.27x'lO7

1.63X10"24.47x1024.65x10"'1.32x1O"8

5.32x10""

3.62X10"11

A.49x10"216.91x10°5.91x10"'2.68x10"*4.49x1O"5

2.59xlO"52.31x10"'*

FIG. 3. Dose rate, neutron and y-ray wall loading at mid-plane of reactor chamber.

10°
10"'

j 10-2
;

y

\

lio-*
? 10"5
-

;

\

10-6
10"7
m-B
10°

Ih W Impniy lOylOOy
102

I0 4

106

108 1CC°Time after Shutdown (s)FIG. 4. Induced radioactivity as a function of time after shutdown.

tritiated water could be required when the effects of chemical sputtering of
charged particles with the structural materials are taken into account [10].
Isotope exchange or an oxidation-reduction operation can be used to extract
the lithium. The 3 He related operations can be omitted from the fuel cycle
when the D-T/D pellet only is used. The tritium inventory of the reactor(in-site inventory) is one-third of LOTRIT [11] and 1/70 of HIBALL [7]. The
path of pellet injection proposed in this design eliminates in-site and off-site
tritium storages.

IAEA-CN-47/H-m-4

349

C(T,D.Hfc. (T,D,H)20
iTritiumExtraclorl
T.O.H

I

'He

3

He.'He[Fuel Purifier [Hlsolope SeparQtorl

HLJQ.D2 Reservoir r

FIG. 5. Fuel cycle (flow rate in g-d'1) in case ofD-T/D/3He pellet.

The use of lead as a coolant leads to reduction of the lithium inventory
in the coolant. The inventory in the structural material (SiC) is assumed toamount to the same value as those of the HIBALL and LOTRIT designs. The
tritium release from the fusion reactor must be minimized so as to avoid any
effect on the environment. Accordingly, a reactor building should have a system
for scavenging tritium not only as a steady operation but also in case of an
accidental release. It is required that the system be able to compensate for thedecrease in the capability of tritium scavenging due to humidity in the air of
the reactor building [12].

5.

CONCLUSIONS

The advanced ICF reactor concept TAKANAWA-I employs a D-T ignitor
and a deuterium fuel pellet to eliminate the lithium blanket and to use leadfor the liquid wall and SiC for shield and vessel whose induced radioactivities
decay rapidly (10~6 Ci-W(th)""1 after one year). Therefore, we may expect

350

TAZIMA et al.

a low tritium inventory (below 200 g), no restriction in the material resources,
easy remote maintenance, and reasonable energy cost (an energy ratio of
about 20). The system requires, however, high driver energy and pumping
power for the liquid lead. A realistic driver system and the pellet compression
scenario are being studied, at present.

REFERENCES
[1]

[2]

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

TAZIMA, T., IDO, S., MONSLER, M.J., in ICF Research by Light-Ion Beam (Proc. Int.
Meeting Nagaoka, 1982), Res. Rep. IPPJ-611, Inst. Plasma Physics, Nagoya Univ.
(1982) 130.
MILEY, G.H., CHOI, C.K., LEE, D., in High-Power Electron and Ion Beam Research
and Technology (Proc. 2nd Int. Top. Conf. Ithaka, NY, 1977), Vol. 1, Laboratory of
Plasma Studies, Cornell University, New York (1977) 243.
IDO, S., TAZIMA, T., Jpn. J. Appl. Phys. 22(1983) 1194.
SHIBA, T., MABESHIMA, K., NAKASHIMA, H., KANDA, Y., submitted to Nucl. Fusion.
YABE, T., Jpn. J. Appl. Phys. 23 (1984) L57.
TAZIMA, T., SATO, M., YONEZU, H., in High Power Particle Beams (Proc. 6th Int.
Conf. Kobe, 1986), Laser Society of Japan (1986).
BADGER, B., HIBALL, A Conceptual Heavy Ion Beam Driver Fusion Reactor Study,
Rep. UWFDM-450, Dept. of Nuclear Engineering, Wisconsin Univ., Madison (1982).
BADGER, B., NUMAK, A Tokamak Reactor Design Study, Rep. UWFDM-330,
Dept. of Nuclear Engineering, Wisconsin Univ., Madison (1979).
YAMACHIKA, H., NAKASHIMA, H., OHTA, M., IDO, S., J. Nucl. Sci. Technol.
20(1983) 701.
NISHIKAWA, M., et al., J. At. En. Soc. Jpn. 25 (1983) 739.
RAGHEB, M., MILEY, G.H., Fusion Technol. 8 (1985) 2061.
NISHIKAWA, M., et al., J. Nucl. Sci. Technol. 22 (1985) 922.

IAEA-CN-47/H-IH-5

STUDY OF COATINGSAS HIGH HEAT FLUX MATERIALS
M. OKADA, M. FUKUTOMI,
T. SHIKAMA, Y. YAMAUCHI
National Research Institute for Metals,
Tsukuba Laboratory,
Sakura-mura, Niihari-gun, Ibaraki-ken,
Japan

Abstraa
STUDY OF COATINGS AS HIGH HEAT FLUX MATERIALS.
Fundamental aspects of coatings for the application as a high heat flux material are studied.
The thermal stability of coatings under various heat load conditions, from an isothermal condition to a dynamic heat load test with a heat flux of more than a few tens of kW- cm""2, is demonstrated. The experimental results are analysed from the viewpoint of thermochemical andthermomechanical interactions between coated films and their substrates. Titanium carbide(TiC) deposited on molybdenum (Mo) and TiC deposited onto graphite are coatings thathave been studied extensively.

1.

INTRODUCTION

It is recognized unanimously that the problem of high heat flux represents
one of the critical technological issues in both near term and long term fusion
R&D programmes [1,2]. Among many scientific and technological research
fields relevant to high heat flux problems, the research into and the development ofhigh heat flux materials should be highlighted.
A variety of new materials are proposed as candidates for high heat flux
materials. Among them, coatings have some advantages over other candidate
materials. Some large tokamaks adopted the coatings as their first wall materials.
The coatings have accumulated records of performance, which have, however,revealed some shortcomings of the coatings as high heat flux materials. The
coatings have shown some failures under high heat flux conditions. As one of
the important failure mechanisms, thermochemical and thermomechanicalinteractions between coated films and their substrates are identified.
Here, we have studied fundamental aspects of coatings for the application
as a high heat flux material. Titanium carbide (TiC) deposited on molybdenum (Mo)(Mo), and TiC deposited on graphite are coatings that we have studied extensively.
351

352
2.

OKADA et al.
EXPERIMENTAL PROCEDURES

Titanium carbide films were deposited on substrates of molybdenum,
graphite and other materials such as tungsten, by the various vapour deposition
methods at 600 to 1400 K. Details of the coating procedures have been reported
elsewhere [3, 4]. Typical thicknesses of the deposited TiC films were about 10 /urn.
Typical dimensions of coated materials were (10-25) X (10-25) X 1 mm 3 .
The coated films were characterized by conventional procedures. Hereafter,
TiC will designate a stoichiometric titanium carbide, which is well crystallized
and has nearly the same chemical bonding state as bulk titanium carbide producedin thermally equilibrated conditions. When non-stoichiometric titanium carbide ismentioned, it will be denoted by TiC x .
Three types of heat load tests were applied in this study. In the first test,
the coatings were heated isothermally at temperatures between 1200 and 2300 K
for 15 to 300 min in a vacuum of 5 X 10"4Pa in a tungsten resistance furnace.
The second and third types of tests were undertaken by using an electron beam
apparatus. The electron beams used as heat source had 60 V in energy and
80 to 150 A- cm"2 in current in the second type test [5] and 3 kV in energy
and about 10 to 100 A-cm"2 in current, in the test of the third type. This low
energy electron beam is simulating near surface heat deposition. The last type
test had a heat flux of about 30 to 300 kW-cm"2, for 1 to 40 ms.
Also, a computer simulation was carried out to examine the behaviour of
coatings under high heat flux conditions. Details of the computer simulation
technique adopted in this study can be found elsewhere [6].

3.

EXPERIMENTAL RESULTS AND DISCUSSION

In the case of isothermal heating in a dynamic vacuum of 5 X 10"4 Pa,
the coated specimens lost their mass by vaporization. The measured vaporization
rates of the TiC and TiCo.67 coatings on various substrates are shown in Fig. 1 asa function of temperature.
First, the vaporization rate depended on the chemical composition of thedeposited films. The TiCo.67 films had a vaporization rate one order larger
than that of a stoichiometric TiC. This is due to the strong dependence of thetitanium vapour pressure on the chemical composition of TiC x . Secondly, thevaporization rate is found to depend strongly on the kind of coating substrate.
The TiC film, which is deposited on graphite substrate, has the lowest vaporization
rate: less than 0.3 jig-cm^-min" 1 at 2300 K. This vaporization rate is about thesame as that of the bulk single crystal TiC. On the other hand, the TiC film on
molybdenum substrate has a vaporization rate of about 180 /ng-cm~2-min~'at 2300 K.

353

IAEA-CN-47/H-ffl-S

2300

TEMPERATURE ( K )2100 1900
1700

4.5

5.0

55

60

1500.

65

7.0

RECIPROCAL TEMPERATURE ( I O V T X K " 1 )FIG. 1. Vaporization rate of TiC and TiC061 films. Full circle: TiC061/Mo; open circle:
TiC/Mo; triangle: TiC/W; square: TiCI graphite.

TEMPERATURE ( K )
1900
1700

2100

E
a.

—

UI

o
2
O

75

t

'- +
46

'\
I

I

I

I

I

50
54
58
RECIPROCAL TEMPERATURE

6-2

Z

66

FIG. 2. Diffusion depth of Mo and W into TiC and TiC15. Full circle: TiCl5/Mo;
open circle:TiC/Mo; triangle: TiC/W.

354

OKADA et al.

This high vaporization rate of TiC on molybdenum substrate is due tothermochemical interaction between the TiC film and its molybdenum substrate.
At elevated temperatures, molybdenum atoms diffuse into TiC films to form
(Ti,Mo) x Ci_ x , which has a far higher titanium vapour pressure than TiC.
As a result, the molybdenum diffusion into TiC enhanced the vaporization oftitanium substantially.
Niobium, vanadium and some other carbide forming refractory metals
were also found to interact with TiC thermochemically. In this study, only
tungsten showed good resistance to the thermochemical interaction with TiC
at elevated temperatures. The diffucion depth of tungsten into TiC was about
one order less than that of molybdenum at 2100 K as is shown in Fig. 2. As aresult, TiC on tungsten substrate had a vaporization rate one order smaller thanthat of TiC on molybdenum as is shown in Fig. 1.
The rate of thermochemical interaction between TiC and its substrate wasfound to depend strongly on the microstructures of TiC. In particular, theexistence of excess titanium vacancies is considered to enhance the thermochemical interaction. The molybdenum diffused into a TiCj 5 film with anapparent activation energy of about 230 kJ-mol" 1 as is shown in Fig. 2. The
TiCj 5 film was analysed and showed a large amount of titanium vacancies.
On the other hand, the stoichiometric TiC interacted with its substratemolybdenum with an apparent activation energy of about 600 kJ-mol" 1 as isshown in Fig. 2. In general, higher temperature coating methods such as CVD
provided TiC films which had less titanum vacancies and a lower rate ofthermochemical interaction with their substrate.
With the increase of heat flux, melting of the coatings takes place. Figure 3
shows the melted area of TiC coating as a function of deposited energy, with aheat flux of 5.3 kW-cm"2 in the test of the second type. The TiC coatedmolybdenum began to melt at lower deposited energy, i.e. at 5-6 kJ-cm" 2 , thanthe uncoated molybdenum which began to melt at about 8 kJ-cm" 2 .
This premature melting of TiC coated molybdenum is due to thermochemicalinteraction between TiC and molybdenum. MoCj _ x and/or (Ti,Mo)Ci_ x isformed at the interface between TiC and molybdenum. As MoCi_ x and (Ti,Mo)C!_ x
have lower melting points than TiC and molybdenum, the initiation of melting
takes place at the interface, in the case of TiC coated molybdenum. Once melting
takes place, the produced liquid phase further enhances the chemical interaction
between TiC and molybdenum, through a liquid phase reaction. Thus, melting
propagates fast, a substantial amount of titanium vapour being produced.
The insertion of a 5 m thick intermediate tungsten layer between the TiC
film and its molybdenum substrate hindered the thermochemical interaction andthe formation of low melting point compounds at the interface. As a result, TiC
coated molybdenum with an intermediate tungsten layer (TiC/W/Mo) demonstrated
excellent resistance to melting, even a little superior to the uncoated molybdenum
as is shown in Fig. 3.

355

IAEA-CN-47/H-m-5

6

8
10
12
,
TOTAL DEPOSITED ENERGY (kJ-cm' z )FIG. 3. Melt area of TiC coatings at a heat flux of 5.3 kW- cm'2. Open circle: TiC/Mo;
open hexagon: uncoated Mo; open square: TiC/W/Mo.

A

E
in

FIG. 4. Optical photograph of graphite after exposure to a heat flux of6kW-cm(a) uncoated graphite; (b) TiC coated Mo.

2:When a heat flux of 5-6 kW- cm 2 was applied to the graphite, a substantial
amount of sublimation took place, but no mechanical fracture was observed. On
the contrary, the TiC coated graphite was cracked to fracture by application of
the same amount of heat flux, as is shown in Fig. 4. Computer simulation indicatedthat strong shear stress is generated at the interface between TiC and graphite, due
to thermomechanical interaction between the TiC film and its graphite substrate.
When the applied heat flux exceeds 30 kW- cm" 2 , flaking of the TiC film is
expected from the computer simulation. In this case, the TiC film will breakwithin its matrix when the TiC film is thick. When the TiC film is thin, exfoliationof TiC from its substrate will take place. The actual test revealed occurrence of thistype of exfoliation of the TiC film on graphite. The experimental results also
indicated that pre-injection of hydrogen isotope atoms would enhance this typeof exfoliation.

356
4.

OKADA et al.
CONCLUSIONSHigh heat flux problems are seriously demanding the development of newmaterials, which will be composed of multielements and multimaterials. These
complex materials will behave differently in different heat load conditions.
One example for the complexity of this behaviour has been treated in thispaper. Further studies an needed using new high heat flux test facilities, which
will simulate the actual heat load conditions in fusion devices; they will contribue
to the construction of reliable database for high heat flux materials [7].

REFERENCES
[1]
[2]
[3]
[4]
[5]
[6]
[7]

KRAKOWSKI, R.A., J. Nucl. Mater. 122 & 123 (1984) 37.
WHITLEY, J.B., J. Nucl. Mater. 133 & 134 (1985) 39.
SHIKAMA, T., SHINNO, H., FUKUTOMI, M., FUJITSUKA, M., KITAJIMA, M.,
OKADA, M., Thin Solid Films 101 (1983) 233.
FUKUTOMI, M., FUJITSUKA, M., SHIKAMA, T., OKADA, M., J. Vac. Sci. Technol. A3(1985)2650.
FUJITSUKA, M., FUKUTOMI, M., OKADA, M., J. Jpn. Inst. Metals (to be published).
SHINNO, H., KITAJIMA, M., J. Nucl. Mater, (to be published).
GAUSTER, W.B., OKADA, M., Report of Japan/US Workshop on High Heat Flux
Materials, National Research Institute for Metals, Tsukuba (1986) 15.

IAEA-CN-47/H-IU-6

DEVELOPMENT OF
SUPERCONDUCTING MATERIALS AND
MAGNET TECHNOLOGY FOR THE
NEXT GENERATION OF
FUSION MACHINES IN JAPANS. SHIMAMOTO, T. ANDO, T. HIYAMA, H. TSUJI,
Y. TAKAHASHI, M. NISHI, E. TADA, K. YOSHIDA,
K. KOIZUMI, K. OKUNO, H. NAKAJIMA, T. KATO,
K. KAWANO, M. HOSHINO, M. OSHIKIRINaka Fusion Research Establishment,
Japan Atomic Energy Research Institute,Naka-machi, Naka-gun, Ibaraki-ken
H. WADA, T. KURODA, K. INOUE, K. TACHIKAWA
National Research Institute for Metals,
Sakura-mura, Niihari-gun, Ibaraki-ken
Japan

Abstract
DEVELOPMENT OF SUPERCONDUCTING MATERIALS AND MAGNET TECHNOLOGYFOR THE NEXT GENERATION OF FUSION MACHINES IN JAPAN.
For the construction of the Fusion Experimental Reactor (FER), which will follow thealready operating JT-60, superconducting magnet technology is making significant progress inJapan. The Japan Atomic Energy Research Institute (JAERI) has systematically developed
toroidal and poloidal coils since 1977 and is ready to construct a 12 T prototype toroidal
coil for the FER. JAERI has also carried out developmental work for pulsed coils. Furthermore, a new technology for advanced high field superconducting materials has been developed
at the National Research Institute for Metals. A 16 T superconductor is ready for large scaleverification and new technologies for 20 T superconductors are being developed.

1.

INTRODUCTION

A leading fusion programme in Japan was the development work for the
Fusion Experimental Reactor (FER). The JT-60 is now in operation and workfor the construction of the FER is progressing. Major tasks for the FER are
the demonstration of the self-ignition state of the D-T plasma for several
hundred seconds and the demonstration of the engineering feasibility of large
tokamak reactors.

357

358

SHIMAMOTO et al.

Maximum Magnetic Field(Bmax)T M C . j

I 2 -,Bmox
12 T
Operation 1985

I I -TOROIDAL COIL OF FUSION
EXPERIMENTAL REACTOR
(FER)
Bmax
12 T
Size
9 x 11 m

TMC-I
Bmox
11T
Operation 1984

NbjSn
Nb3Sn
10 -

9 -

O

8 jDav : Average i
Diameter i

7 Jm^#NbjSn

CLUSTER BACKUP TEST MODULE COIL (TMC-I)
COIL (CBC)Bmai 10T
Bmox 10TDav
1.1 m
Dav 0.6m
Operation ( 9 8 2
Operation 1982

CLUSTER TEST COIL (CTC)
Bmax
7T
Dav
1.5m
Operation 1 9 8 0

Im
r

NbTi

^^^^NbTi
IEA-LCT COIL
Bmax
8T
Size
3.5x 4.5m
Operation 1982

FIG. 1. JAERI's superconducting toroidal programme.

It has been widely recognized that the technology of superconducting
magnets is absolutely indispensable for the FER. If copper magnets would be
used for the FER, the electric power consumption would be much higher than is
feasible and it would be completely impossible to achieve pulsed plasma operation with long pulses [1]. Therefore, the Japan Atomic Energy ResearchInstitute (JAERI) has systematically developed toroidal and poloidal coils
since 1977 [2] and remarkable progress has been made.
In addition to the development of large magnets, a new technology foradvanced high field superconducting materials has been developed at the
National Research Institute for Metals (NRIM). Major achievements for application in the range of 16-20 T are described.

IAEA-CN-47/H-DI-6

359

FIG. 2. The JA-LCT coil developed by JAERI.

2.

LARGE HIGH FIELD TOROIDAL COILSThe key technologies required for toroidal coils are those for the construction of large coils (about 10 m high) and for the realization of high field coils(12 T). JAERI has carried out two major programmes for the development ofthese technologies: the Large Coil Task, with the aim of developing large coils(about 5 m high), under the auspices of the International Energy Agency, andthe Cluster Test Programme, with the aim of developing high field coils(see Fig. 1).

360

SHIMAMOTO et al.

0

1

2
TIME

3
(h)

FIG. 3. Standard-I test of the JA-LCT coil; 100% charging with the backup field of the coils
developed in the USA (GD, GE, WH), in Switzerland and by Euratom.

2.1. The Large Coil Task

The Large Coil Task (LCT) is an international programme for the studyof problems in connection with the design and construction of toroidal coils
for large tokamaks. As a participant in this programme, JAERI has developed
the Japanese LCT coil (JA-LCT coil) [3].
The JA-LCT coil, shown in Fig. 2, was the first coil to be charged up to
100% current. A domestic test was performed in 1982 at JAERI. After thecompletion of another five LCT coils (three of which were developed in theUSA, one in Switzerland and one by Euratom), the JA-LCT coil was charged
up to 100% current. The other five coils were charged up to about 80% current
in August 1986 at Oak Ridge National Laboratory. The maximum magnetic field
generated in the testing of the JA-LCT coil was 8 T (see Fig. 3). The JA-LCT
coil showed excellent performance, as designed [4], and the applicability of thetechnology for the construction of a large coil (about one-half of a toroidal
coil for the FER) was proven. The major technologies developed in thisprogramme are listed below.
- The critical current density of Nb-Ti was doubled.
- The maximum heat transfer capability from the conductor to liquid helium
was increased to 1.3 X 104 W/m 2 , which is about four times that achieved
with a flat metal surface [5].
- The design and fabrication of high current conductors (more than 10 kA)
were realized and winding techniques were developed.

IAEA-CN-47/H-IH-6

361

These techniques were the basis for the development of a 50 kA pulsed
conductor for poloidal coils.
— Design and fabrication methods for a reliable pool boiling water cooled
large coil were established.
— An experimental comparison between a pool boiling water cooled coiland a forced flow cooled coil was performed and guidelines for further
application and development were set up [6].
— High strength structural materials (yield strength 800 MPa) were developed
and used for the LCT coil; also, a new cryogenic steel was developed(yield strength more than 1200 MPa [7]).
— Insulating materials were developed and evaluated [8].
— A mechanical design standard for superconducting large coils wasinvestigated and the stress data obtained were evaluated with a newconcept [9].
— The validity of the pulse loss analysis of large coils was confirmed bydump loss measurements.
— Techniques for the fabrication and operation of large helium liquefiers/
refrigerators (350 L/h, 1.2 kW) as well as magnets were established [10].
2.2. The Cluster Test Programme
JAERI has carried out the Cluster Test Programme with the aim of developing
high field toroidal coils. JAERI has already developed cluster test coils for 7 T, cluster
backup coils for 10 T [ 11 ] and a test module coil for 12 T. This test module
coil is an Nb 3 Sn, pool boiling water cooled coil with 1.6 m o.d. It has beensuccessfully charged up to 33 A/mm2 at 12 T and up to 45 A/mm 2 at 10 T [12].
On the basis of this achievement the designed current density of the FER was
raised to 30 A/mm 2 at 12 T.

3.

POLOIDAL COILS FO R PULSED OPERATION

One of the most important requirements for poloidal coils is the operation
under pulsed current and field. LCT coils designed for DC operation are charged
up to their rated current in about one hour, but pulsed poloidal coils have to be
charged up to their rated current in a few seconds; therefore, a much higher
technology is needed for pulsed poloidal coils, even when the achievements of
the LCT and the Cluster Test Programmes are considered.
Major requirements for the pulsed coil technology are: (a) high operating
current (more than 30 kA), (b) high voltage insulation in order to withstandthe operating voltage of about 10 kV, and (c) low pulse loss in order to keep
the coil stable and to keep the refrigeration load low during pulsed operation;
the pulsed loss has to be decreased to about 1/1000 of that of DC coils.

362

SHIMAMOTO et al.

JAERI has carried out developmental work for pulsed coils, as shown in
Fig. 4 [ 13, 14]. After the development of pulsed conductors of a capacity of
up to 50 kA and the development of a pulsed coil (Poloidal Unit Pancake [15],30 kA, 1.5 MJ), JAERI started the programme of Demonstration Poloidal
Coils (DPC) in 1985. This programme aims at demonstrating the pulsed coiltechnology for inner poloidal coils of tokamaks. Three pulsed coils with
2 m o. d. are being developed. The DPC is composed of two Nb-Ti, 30 kA
pulsed coils (DPC-1, DPC-2) designed for 7 T, and an Nb3Sn, 10 kA pulsed
coil (DPC-EX) designed for 10 T. The stored energy of DPC-1 and DPC-2
is 30 MJ; when the DPC-EX is added, the stored energy is increased to 40 MJ.
Forced flow cooling was selected for the DPC so that it can withstand thenominal operating voltage of 6.6 kV [16].
At present, JAERI constructs a test facility for the DPC, the Superconducting Engineering Test Facility (SETF) [17]. This includes a supercriticalhelium pump system with a capacity of 500 g/s at 4.0 K. For the charging of
the DPC, the large power supply system of the JT-60 will be connected to the
SETF and a unit with a capacity of 5 kV and 53 kA will be used [18].

4.

RESEARCH AND DEVELOPMENT OF HIGH FIELD
SUPERCONDUCTORS AT NRIM

4.1. Titanium doped Nb 3 Sn multifilament wire conductor
A modification of the bronze process has been made; Ti is added to the
Nb core and/or the Cu-Sn conductor [19]. The optimal amount of Ti addition
is found to be 2-3 at.% to the core and 0.3-0.5 at.% to the matrix. With the Ti
additions the upper critical field and the critical current density at high fields
are drastically increased, as shown in Fig. 5. Titanium doped Nb3Sn multifilament wire conductors have already been used for choke coils of MFTF-B
at the Lawrence Livermore Laboratory and for the outer coil of a 31 T hybrid
magnet at Tohoku University, as well as for several Nb 3 Sn/NbTi magnets of
the 15 T category at NRIM and other institutes in Japan. The magnet at
NRIM has a bore of 180 mm, and when a V3Ga magnet is inserted, a field of
18.1 T can be generated, which is the highest field ever achieved by a superconducting magnet [20].
4.2. V2 (Hf,Zr) multifilament wire conductor
The superconducting Laves phase compound, V2(Hf,Zr), is known to beresistant to neutron irradiation and insensitive to applied strain. It is thus assumed
to be a promising material for fusion magnets. Multifilament wire conductors of
this material (1634 cores) have been developed by a composite diffusion process.

363

IAEA-CN-47/H-ni-6OperatingCurrent
50

JAERI
Demo Poloidal Coil

30

Poloidal Coils

JAERI
Poloidal Unit Pancake

JAERI
Pulser-F,Gof the FER

o

20
JAERI
Pulser-D,Eo11983)

LANL
300kJ
o

10 IkA).
5

IPPJ Ninon Univ.
RPC-IIO

3

ETL
400 kJ

ANL'
|.5MJ

ANL
3.5MJ
ETL

CO

umber of Coils : 12-22
iameter : 2 - 2 0 m

3MJ4MJ
Pulser-C(1982}Im. ETL 400kJ

2°osaka univ. sookJ
i

t

i

I

i

t

Stored Energy
I

i

lit

I

i

i

0.1 2 3 5 | 0 2 3 5 10 2 3 5 100 2 3 5 |000 (MJ)
FIG. 4. Existing pulsed coils for the development of poloidal coils.

|.Q. Nb 3 (AI,Ge)
o
<
*«
c

O

l
0

5

10
15
20
Magnetic Field (T)25

FIG. 5. High field critical current density of superconductors under development at NRIM.

A critical temperature of 9.9 K, an upper critical field of 22.1 T at 4.2 K, and an
overall critical current density of 1 X 104 A/cm2 have been obtained at 16 T,
as shown in Fig. 5 [21 ]. Since the upper critical field and hence the criticalcurrent density at high fields are increased rapidly with decreasing temperature,these conductors may be especially suitable for use at superfluid helium temperatures.

364

SHIMAMOTO et al.

4.3. Nb3Al and Nb 3 (Al,X) tape conductors
Many of the superconducting materials to which the bronze process cannot
be successfully applied have high temperature phases with a high critical temperature, but they are not in equilibrium at ambient temperatures. The advanced
fabrication techniques have been developed for such materials as Nb3 Al andNb 3 (Al,X). One technique is a modification of liquid quenching. With thistechnique, an Nb3(Al,X) (X = Ge,Si) tape conductor can be produced by
quenching the liquid phase of this material on a moving copper substrate tapeand heating the tape to above 800°C [22]. Values of the critical current densityof about 1 X 10s A/cm2 at 20 T have been obtained, as shown in Fig. 5.
The second technique is an application of the high energy and high density
beam irradiation where either an electron beam or a laser beam irradiates a
travelling composite tape containing powders of Nb, Al and Ge compacted inan Nb sheath. Because of the highly concentrated energy beam irradiation,
extremely localized heating and subsequent cooling occur in the travellingtape, and the high temperature A-15 phase can be formed and preserved. The
maximum critical temperatures obtained are 18.5 K for Nb3 Al and 20.2 K
for Nb3(Al,Ge). The measured critical current density characteristic of Nb3 Al
is shown in Fig. 5 [23, 24].

5.

CONCLUSIONSFor the Fusion Experimental Reactor, which is being constructed now
that JT-60 has been put in operation, superconducting magnet technology has
made significant progress in Japan. JAERI is ready to construct a 12 T prototype toroidal coil for the FER. For poloidal coils, in addition to the ongoingdevelopment of the Demonstration Poloidal Coil at JAERI, development andtesting of a large diameter ring prototype coil are necessary. Also, efforts
to develop large cryogenic systems for the FER will be continued by JAERI.
Research and development for advanced superconductors have been
actively carried out at NRIM. Titanium doped Nb3 Sn is now ready for large
scale verification, using a 16 T.coil with a coil bore of several metres. This
ambitious project can bring about a breakthrough in the development of theadvanced 16 T tokamak [25].

REFERENCES[1] SHIMAMOTO, S., "The present role of superconductivity in fusion", Proc. 11 th Int.
Cryogenic Engineering Conf., Butterworths, London (1986) 127.
[2] SHIMAMOTO, S., Cryogenics 25 (1985) 171.

1AEA-CN-47/H-IH-6

365[3] SHIMAMOTO, S., et al., Cryogenics 25 (1985) 225.
[4] OKUNO, K., et al., "Testing of the pool-boiling-cooled Japanese LCT coil in theInternational Fusion Superconducting Magnet Test Facility", Proc. 1986 Applied
Superconductivity Conf.
[5] NISHI, M., et al., IEEE Trans. Magn. 17 1 (1981) 904.
[6] TSUJI, H., et al., Cryogenics 25 (1985) 537.
[7] NAKAJIMA, H., et al., Adv. Cryog. Eng. 32, p. 347.
[8] KOIZUMI, K., et al., Adv. Cryog. Eng. 28, p. 223.
[9] YOSH1DA, K., et al., in Proc. 9th Conf. Magnet Technology, SIN, Switzerland(1985)435.
[10] TADA, E., et al., in Proc. 9th Int. Cryogenic Engineering Conf., Butterworths,
London (1982) 93.
[11] ANDO, T , et al., in Proc. 10th Int. Cryogenic Engineering Conf., Butterworths,
London (1984) 173.
[ 12] ANDO, T , et al., "Development of high-field superconducting coil for the tokamak
fusion machine in JAERI", Fusion Engineering (Proc. 1 lth Symp. Austin, TX, 1985),
Cat. No. CH-2251-7, IEEE, New York (1985) 991.
[13] SHIMAMOTO, S., et al., in Technology of Fusion Energy (Proc. 5th Top. Mtg, 1983),
American Nuclear Society (1983) 924.
[14] TSUJI, H., et al., Adv. Cryog. Eng. 29, p. 105.
[15] TAKAHASHI, Y., et al., "Experimental results of 30-kA 70-cm bore Poloidal Unit
Pancake", Proc. 1986 Applied Superconductivity Conf.
[16] TSUJI, H., et al., "Design selections for the fabrication of the Demonstration Poloidal
Coil", Proc. 7th Top. Mtg Technology of Fusion Energy (1986).
[17] KATO, T., et al., "Cryogenic system component development for Fusion Experimental
Reactor at JAERI", Proc. 7th Top. Mtg Technology of Fusion Energy (1986).
[18] SHIMAMOTO, S., et al., "System design of the Demonstration Poloidal Coils and theTest Facilities", Proc. 1986 Applied Superconductivity Conf.
[19] TACHIKAWA, K., et al., J. Nucl. Mater. 133-134(1985) 830.
[20] TACHIKAWA, K., et al., to be published in IEEE Trans. Magn.
[21] INOUE, K., et al., IEEE Trans. Magn. 21 (1985) 467.
[22] TOGANO, K., et al., IEEE Trans. Magn. 19 (1983) 414.
[23] KUMAKURA, H., et al., Appl. Phys. Lett. 49 (1986) 46.
[24] KUMAKURA, H., et al., Appl. Phys. Lett. 48 (1986) 601.
[25] INTERNATIONAL ATOMIC ENERGY AGENCY, Tokamak Concept Innovations(Proc. Specialists Mtg Vienna, 1986), IAEA-TECDOC-373, IAEA, Vienna (1986) 243.

IAEA-CN-47/H-IO-7

ENGINEERING STUDY OF THE FEASIBILITY
OF A FUSION POWER REACTOREMPLOYING THE DENSE Z-PINCHPLASMA CONFINEMENT GEOMETRY
H.R. BOLTON, P. CHOI, A.E. DANGOR,
M.G. HAINES, M. JAVADI, S.J. PEERLESS,
A. POWER, A.E. ROBSON, S.P. WALKER
Centre for Fusion Studies,
Imperial College of Science and Technology,
London, United Kingdom
Abstract
ENGINEERING STUDY OF THE FEASIBILITY OF A FUSION POWER REACTOR EMPLOYINGTHE DENSE Z-PINCH PLASMA CONFINEMENT GEOMETRY.
The findings of an engineering assessment of a commercial fusion reactor based on the dense
Z-pinch (DZP) confinement geometry are reported. A major requirement is a system for supplying electrical energy to the pinch with the required pulse form and at high repetition rates; the very high rateof current rise required imposes a low upper limit on the inductance of the primary circuit. A second
requirement is that the electrodes supplying the electrical power to the pinch must be capable ofcontinuous operation over long periods with acceptably low rates of damage from repeated pinch
discharges; for this, liquid cooled electrodes or even all-liquid electrodes may be needed. The studies
showed the importance of ensuring the recovery of the electromagnetic field energy from the pinch zone
after discharges, and that asymmetric configurations of the discharge zone are unlikely to be satisfactory. It can be expected that a DZP reactor will operate at moderate or high pressure and will therefore
require a pressure vessel similar to that of pressurized water fission reactors. Present indications arethat neutron damage to the containment vessel can be kept at acceptable levels and that power densities
similar to those of PWRs will be achieved. Therefore, the DZP concept could provide an economically
viable power source. Three tentative reactor layouts are presented; in each the plasma column isapproximately 100 mm long, 30 ^m in diameter and of 7 x 1027 m~3 number density carrying a peakcurrent of 1.5 MA.

1.

INTRODUCTION

The many studies of tokamak reactors have led to the realization of the needfor reactors of lower overall size and complexity, and higher power density [1-3].
The dense Z-pinch (DZP) offers the possibility of achieving these.
The two commonly quoted shortcomings of the DZP are MHD instability andend energy losses. There is now firm experimental evidence [4] that a compressional
Z-pinch can exhibit stability for up to 20 radial Alfve"n transit times, and losses havebeen shown theoretically to be a small fraction of fusion energy output under the highdensity plasma conditions envisaged for a reactor [5]. Typically, these conditions are
367

368

BOLTON et al.

a number density of 7 X 1027 m"3, a peak current of 1.5 X 106 A, a temperature of
5 X 108 K, a pulse duration of 300 ns and a plasma length of 100 mm, giving aplasma diameter of 30 /xm. Such a pinch would release about 4 MJ of fusion energy
per discharge. The conditions proposed here are similar to those used previously [6],except that we aim at reducing significantly the pulse length required for alarge fractional burn by employing radiative collapse to largely increase plasma
density [7].
Questions do remain regarding the attainment of the conditions required forfusion and the stability of the pinch, but they are not discussed further in this paper.

2.

ENGINEERING CONSTRAINTS ON REACTOR DESIGNThis study has identified the principal physical requirements of a DZP reactor
and determined the constraints within which such a reactor must be designed. We
have tried to remain aware of the need for the reactor to satisfy economic as well as
technical criteria. The following topics have been identified as crucial.
2.1. Repetitive pulse feed
The repeated production of the current and voltage waveforms necessary for thecreation of the pinch conditions has been studied. The use of circuits employing
wholly passive elements, including non-linear saturable inductors as magnetic
switches, has been shown to be attractive, particularly with regard to efficientmagnetic energy recovery. An order of magnitude more energy is stored in themagnetic field around the pinch than is stored as thermal energy in the pinch itself,and efficient recovery of this energy is necessary. Simulations of a pinch of varying
radius coupled to a circuit employing saturable inductors indicate that the required
high fractional recovery is possible in principle. The reactor must present to thepulsed power supply an inductance of the order of only 1 /iH, to permit the necessaryrate of current rise of 1013 A-s"1 to be achieved with practicable supply voltages.
Also, unless the electric current flow is symmetrical about the axis of the pinch, large
net electromagnetic forces act sideways on the plasma. Simulations of the plasma
response predict a terminal velocity of 2 x 105 m-s" 1 . This is unacceptable becauseof both the accretion of the cold gas and the enhancement of instabilities. The needfor symmetry and low inductance imposes geometrical constraints on the reactor
design.
2.2. Electrode wear
Energy losses from the ends of the plasma result largely in thermal energy
generation at the electrode tips. Although constituting only a small fraction of reactor
output, these end losses represent a significant absolute power, and some evaporation

1AEA-CN-47/H-HI-7369

of electrode material is inevitable. Estimates of the loss rate have been made andcompared with experimental data [8] from laser targets subject to similarly highpower densities (of the order of 1018 W-m" 2 for 10~7 s). While this loss is perhaps
tolerable, it can be avoided by the use of an electrode with a liquid metal surface.
2.3. Tritium breeding and inventory
The breeding of tritium in a DZP reactor should present no particular difficulty
and because of the reduced complexity of this reactor it would be easier than in
tokamaks. The use of heavy water as moderator and coolant has attractions, and adetailed study of this has been made [9].
The use of a high pressure reaction region increases the difficulty of a significant tritium inventory which is present both as a gas and dissolved in materials(especially water or liquid metal) bordering this region. This is of concern regarding
the initial availability and the doubling time. Solutions considered include prior saturation with a chemically similar gas such as deuterium, and the use of filaments ofa liquid or solid D-T mixture as fuel.
2.4. Heat removal
Water, lithium and lithium-lead mixtures have been considered as moderatorsand coolants. In the DZP reactor, most of the thermal energy is generated within the
coolant, rather than being transferred to it by conduction and convection. The
reduced need for large heat transfer surface areas of the DZP helps to attainattractively high power densities.
2.5. Pressure vessel damage
The DZP is unique among the proposed fusion reactors in its probable use ofa high pressure (—10 MPa) reaction region. The reactor thus will probably requirepressure vessel technology akin to that of the PWR, and similarly stringent constraints on neutron radiation damage to the vessel would need to be applied. PWR
vessels experience typically 10~3 times the damage anticipated before the replacement of tokamak first walls. Calculations [9] show that adequate internal shielding
and breeding materials can be included in the DZP to reduce vessel damage to thevalues found for PWR vessels, while power densities similar to those of PWRs andmore than an order of magnitude greater than those of tokamaks will still be achieved.

3.

REACTOR CONFIGURATIONS CONSIDERED

No definitive reactor configuration has been established, but, to assist with theidentification of the main engineering problems, we have set up tentative layouts asbases for preliminary assessment. Two of these layouts are shown in Figs 1 and 2.

370

B O L T O N et al.

High voltage
electrode
Lithium _H
coolant _TTfl

To gas• ^ S processing
C f = plant
Combined electricalinsulator andporous plug
To tritium
extraction plantand steam
generator

Main
Lithium
inflow=?:— Lithium coolant
Lowerelectrode

FIG. 1. Reactor layout; solid first wall.

To gasprocessing
plant

Lithium
coolant•High-voltage
electrodeMain- lithium
inflow

Free vortex
surface

Lithium
coolant
and D-Tgas supply

Outflow to lithium
processing plantand steam generator

FIG. 2. Reactor layout; vortex separation.

371

IAEA-CN-47/H-HI-7

PRESSURE.
VESSEL 'INSULATOR
BUBBLECATCHER

BUBBLE
GENERATOR
D 2 IN D-T FUEL, COLD WATER IN_1m—.
FIG. 3. Layout of a water-cooled reactor (from Ref. [10]).

In Fig. 1, the liquid lithium blanket is separated from the central (gas filled)zone by a metal wall. Deuterium and tritium are supplied to the reaction zone, anda mixture of D, T and He and lithium vapour is withdrawn to the processing plant
for separation and cooling as required before recycling. The liquid lithium would also
be recirculated through the external plant for cooling and extraction of tritium. The
upper electrode would be cooled by an internal flow of, for instance, liquid lithium.
Some liquid would pass through a porous region at the lower end of the electrode,providing protection from the intense local heat flux from the plasma. The lower
electrode would be protected by a layer of liquid lithium. In Fig. 2 the gas zone isseparated from the liquid lithium body by maintaining vortex motion in the latter. The
electrode arrangements would generally be similar to those shown in Fig. 1.
An outline of a DZP design was published [10] shortly after our study began,
and is reproduced here as Fig. 3. We have used both this proposal and our own ideas
primarily as vehicles for the identification of reactor requirements and constraints.

4.

CONCLUSIONS

It is felt that the DZP is a very promising candidate for a fusion power reactor.
Experimental and theoretical progress with the physics is encouraging, particularlyin the areas of stability and radiative collapse. The assessment of the engineering
feasibility has not shown the concept to suffer from any insuperable problems, butsuggests that relatively small, simple DZP reactors could be built with high power
densities.

372

BOLTON et al.

ACKNOWLEDGEMENTS
The following persons have contributed to this study: W. Bessel, A. Folkierski(Centre for Fusion Studies), K. Malik (Central Electricity Generating Board), and
M. Goodman (Atomic Weapons Research Establishment).
The support of M. Javadi and A. Power of the Science and Engineering
Research Council is also acknowledged.

REFERENCES
[1]

[2]
[3]

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

SCHMITTER, K.H., "The tokamak: An imperfect frame of reference", Unconventional
Approaches to Fusion (BRUNELLI, B., LEOTTA, G.G, Eds), Plenum Press, New York(1982).
LIDSKY, L.M., The trouble with fusion, Technol. Rev. 86 (1983) 32.
CARRUTHERS, R., "Criteria for the assessment of reactor potential", Unconventional
Approaches to Fusion (BRUNELLI, B., LEOTTA, G.G., Eds), Plenum Press, New York(1982).
CHOI, P., et al., in Plasma Physics and Controlled Nuclear Fusion Research 1978 (Proc. 7th Int.
Conf. Innsbruck, 1978), Vol. 2, IAEA, Vienna (1979) 69.
HAINES, M.G., Proc. Phys. Soc. 77 (1961) 643. (See also Ref. [7].)HAGENSON, R.L., et al., The Dense Z-Pinch (DZP) as a fusion power reactor: preliminary
scaling calculations and systems energy balance, Nucl. Fusion 21 (1981) 1351.
HAINES, M.G., Phys. Scr. T2/2 (1982) 380.
HAINES, M.G., et al., in Plasma Physics and Controlled Nuclear Fusion Research 1982 (Proc.
9th Int. Conf. Baltimore, 1982), Vol. 1, IAEA, Vienna (1983) 163.
WALKER, S.P., JAVADI, M., ROBSON, A.E., in Fusion Technology (Proc. 14th Symp.
Avignon, 1986), Pergamon Press, Oxford (1987).
ROBSON, A.E., "The high density z-pinch as a fusion reactor", Proc. 1st Conf. on Dense
Z-Pinches for Fusion, Naval Research Laboratory, Washington, D.C. (1984) 2.

IAEA-CN-47/H-HI-9

DEUTERIUM BASED FUEL CYCLES
The benefits of tritium leanness*
W. KERNBICHLER, R. FELDBACHER,
M. HEINDLER, A. NASSRI
Abteilung fur Energie-Physik,
Institut fur Theoretische Physik,Technische Universitat Graz,
Graz
K. SCHOPF
Institut fur Theoretische Physik,
Universitat Innsbruck,
Innsbruck
Austria
Abstract
DEUTERIUM BASED FUEL CYCLES: THE BENEFITS OF TRITIUM LEANNESS.
The correlation between the tritium fuelling rate relative to that of deuterium and the tritium
breeding requirement in the blanket is investigated. The entire domain between a 1:1 deuterium-tritium(D-T) fuel mixture and D-D fusion (with and without recirculation of tritium and/or helium-3) is studied
for plasma confinement and reactor parameter values which are believed to be representative of concepts
associated with D, T and 3 He based fuel cycles. The paper describes the model that is used to simulate(1) the energy and particle flows into, within and out of the plasma, (2) the plasma confinement, and(3) the reactor concept. Special emphasis is put on the behaviour of suprathermal ions and the synchrotron radiation loss. A short description of the in-house data library developed for this study is also given.
The results show the strongly non-linear relationship between the tritium breeding capacity of theblanket, the tritium fraction in the plasma and the fusion power density. The possible operational
domain, defined by ignition or by the positive net electrical power output, is calculated as a functionof the T/D fuelling rate ratio and the plasma temperature. The neutron power fraction and the number
ratio for slow and fast neutrons as a function of these variables is also presented. Sensitivity studies havebeen performed for a large part of the externally given parameters in order to ensure that the results
shown in this paper can be considered as representative. On the basis of these and similar results itshould be possible to optimize a D-T reactor with respect to a broad range of physical, technologicaland other aspects.

* Work supported by:International Atomic Energy Agency, Vienna (Research Contract No. 4082/R1/RB).
Osterreichisches Bundesministerium fur Wissenschaft und Forschung, Vienna (Zl. 77.003/3-25/85
and 77.651/2-25/86).
Steiermarkische Landesregierung, Abteilung fur Wissenschaft und Forschung, Graz
(AAW 12He 15-80/28).
Jubilaumsfonds der Osterreichischen Nationalbank, Vienna (No. 2726).
Arbeitsmarktfdrderung, Landesarbeitsamt Steiermark, Graz.

373

374

1.

KERNBICHLER et al.

INTRODUCTION

In principle, there is a much larger variety of fuel options for fusion than forfission. However, the only credible candidate as a 'near term' fusion fuel seems to
be a mixture of deuterium (D) and tritium (T) with a density ratio of 1:1. This choicereflects the prevalence of physical arguments over technological arguments because
low ignition temperature and high power density are traded in for particularly severe
and largely unresolved problems regarding reactor design and operation as well asradiological protection and safety, which problems are intrinsically associated withthe use of tritium.
While these reactor issues attract little but increasing attention by the fusion
community, there is renewed interest in alternative fusion fuels such as D- 3 He
[1, 2]. Much research would be required, however, for these advanced fuel concepts
to reach a level of maturity comparable to that of concepts in the mainline research.
Therefore it may be of interest to assume that D-T reactors will come first andto investigate the possible consequences of their failure to meet the design criteria.
One especially sensitive component of the reactor seems to be the blanket. For this,the challenge is to find a combination of materials that are compatible, offer a proper
environment for energy conversion and transport, and ensure a high degree ofneutron multiplicaton and tritium breeding via n-Li fissioning [3].
At temperatures typical for D-T reactors, virtually all tritium breeding would
take place in the blanket. An increase of the operating temperature would, however,allow the deuterium in the plasma to assist the breeding of tritium via one of the two
equally likely D-D channels ('plasma breeding').
So far, reactor concepts relying only on blanket breeding (D-T, 1:1 ratio)or, alternatively, relying only on plasma breeding (D-D, (semi-)catalysed D) haveattracted attention. Here we suggest an investigation of the entire spectrum of possibilities. This should eventually permit optimization of the design with respect to theoverall competitiveness of the reactor.

2.

MODELS FOR PLASMA CONFINEMENT AND
REACTOR DESCRIPTION

Previous versions of our zero-dimensional steady state plasma and reactor simulation model and the incorporation of confinement specification parameters aredescribed in Refs [4-6].
The improved version of the plasma model used here is based on the simultaneous solution of energy and particle balance equations for a large number of ion
species (D, T, 3He, p and a) and reaction channels (14 primary, progeny and side
reactions). This model includes several formulae for synchrotron power loss (seeIAEA-CN-47/H-m-9375

Section 2.2); also, it accounts for fast-ion effects, using a continuous slowing-downtheory modified to include nuclear elastic scattering (NES), leakage and fusion
during slowing-down. An improved treatment of synchrotron losses, based on a
recently developed formula for the absorption coefficient [7], and a discontinuous
treatment of fast-ion slowing-down (see Section 2.1) are being developed.
Each isotope is associated with fuelling rate and/or recirculation efficiency
specifications. The energy flow is described by the various energy transformation
efficiencies linking the fusion power with the net electrical power that can be supplied
to customers.

2.1. Fast ion energy deposition
For a fusion plasma it is essential to know the energy fractions of energetic ionsgoing to the various ion species and to the electrons. We have developed a kinetic
equation for a modified continuous slowing-down (MCSD) description that retains
the anisotropy of scattering events and the finite temperature of the backgroundplasma; this equation replaces the more common isotropic, cold target and single
speed models [8, 9].
The kinetic equation reads as follows:f >."<*> + i f(1)>J (E) + l/r,(E)k(E,t);fs(E) = 0for

E > Max (E^)

where i and j denote Maxwellian particle species, s denotes the suprathermal ion
species, < ) indicates Maxwell-Maxwell averaged quantities and I > indicatesbeam-Maxwell averaged quantities. Note that fs is normalized to the particle densityof the suprathermal ion species s.

376

KERNBICHLER et al.

For steady state operation the distribution function is found to be as follows:,s - E)^)sLET(E)dt(2)
x exp ^ - 1

dE'
dE . LETdt )s (

)E

dt}

where the first factor gives the slowing-down distribution associated with the
products of fusion for i- and j-type ions, and the exponential accounts for losses due
to fusion and other processes (rs) occurring during deceleration from Eo,y to E. The
denominators contain the small and large energy transfer (SET, LET) rates. While
the SET term is given by the well known Sivukhin formula (given in Ref. [8]), theenergy dependent LET term is given by
LET

dE .LETdt(vs = v, Ti = 0)

V
V,

-

1 + m,

+

1+

i

1 +

(V"

exp

-

exp

- 2kT, (v+vs)

2kT;

(3)

;

where the first factor in the integrand is the common expression for the energy loss
rate to a cold plasma and vs = V2E/ms.
As an illustration, an example of an improved ion energy distribution function
calculated from Eq. (2) is shown in Fig. l(a). This formula is incorporated into theplasma simulation model, which can then be used to calculate for each fuel mix thefraction of fusion energy transferred to ions and electrons, as shown in Fig. l(b) fora typical case.
The LET effects encountered in the whole range from conventional D-T fuels
to the D-D fuels considered here are small enough to justify this improved MCSD

377

IAEA-CN-47/H-m-9
(b)

(a)
1.0

4.5

\
•

4.0 -

•

/

•

•/

!

•

/ .•
j

3.5-

C+N
C+F
C+L
C+N+F+L

3

a

3.0-

•H

H
V
IB

0

500
1000
Triton Energy, E T (keV)

1

5 10
50 100Ion Temperature, T^ (keV)

FIG. 1. (a) Energy distribution of D-D generated tritons in a catalysed deuterium (CAT-D) plasma(fuel: D; fusion products T and 3He assumed to be recycled with 99% efficiency). Also shown is theeffect of nuclear elastic scattering (N) as well as of fast particle removal by fusion (F) and by other loss
mechanisms (L) in addition to Coulomb scattering (C).
(b) Fraction of fusion energy transferred to ions and electrons for a CAT-D plasma. Note thatsfi.i + sfii,e < !• because of the energy loss associated with the loss of fast particles.
Notation as in (a).
10.0

0.1
10
20
Ion Temperature,
FIG. 2. Importance of synchrotron radiation loss (sy) from a CAT-D plasma, as compared withbremsstrahlung (br) andfusion power (fit) carried by neutrons (n) and charged particles (c). The spreadof the synchrotron losses reflects the uncertainties associated with (from left to right along the line):
Refs [15, 18] for tokamak geometry, Ref. [17] for slab geometry, Ref. [18] for slab and cylinder geometry and Ref. [14] for cylinder geometry. The parameter values are those described in the reference
parameter set.

378

KERNBICHLER et al.

treatment. For fuel mixtures with high operating temperatures and abundant multiMeV fusion products, a new formula based on a Boltzmann-Fokker-Planck typeequation and on transfer matrices is being prepared.
2.2. Synchrotron radiation loss
Energy loss by synchrotron radiation is considered a minor contribution to
the energy balance in the case of 1:1 D-T plasmas. However, at higher operation
temperatures this loss mechanism requires careful consideration. We intend to avoid
simple formulae with low accuracy (as, for instance, presented in Ref. [10]), but on
the other hand we do not wish to have to rely on very elaborate computer codes
requiring a detailed specification of the confinement scheme [11, 12].
We have evaluated all available formulae for synchrotron loss calculations.
We have extended them, where necessary, to finite magnetic field utilization beta
(B is replaced by B V l - j 3 [13]) and to finite wall reflectivity (the surface radiation intensity I is replaced by 1(1 -R)), and we have chosen the most accurate formulae for our plasma model [14-19]. Figure 2 illustrates the relative importance of thesynchrotron power loss spread of the various formulae used.
2.3. Nuclear data library
A complete and consistent set of cross-section data and derived quantities (forinstance, ion distribution averaged reactivity parameters) is required for these investigations, in a standardized, computer readable form. Therefore, the cross-section
library DATLIB has been developed [20]. It is based on existing data files [21-26]
as well as on an extension, compilation and evaluation of previously publishedbibliographies, the most important of which are Refs [27, 28]. DATLIB includes
also an evaluation of the literature quoted in these publications.
DATLIB contains data for 16 fusion reaction channels and for nuclear elastic
scattering among the isotopes p, D, T, 3He and a, which are of relevance here.
Moreover, DATLIB contains the fusion-relevant cross-section data for all isotopes
up to U B and is subject to continuous extension and development.

3.

CASE STUDY SPECIFICATIONS

The large number of parameters for which data must be provided by the user
permit the simulation of a great variety of fuel cycles, confinement schemes andreactor concepts. To be able to present a small selection of results, it is necessary
to choose one particular case; this case is defined by our 'reference' set of parametervalues. Evidently, there is much freedom in the choice of these parameter values, inparticular because no operating reactor of the kind investigated here can be taken as'standard'. On the other hand, each parameter can be subjected to a sensitivity study

IAEA-CN-47/H-IH-9379

to reveal the size of its impact on the results. We have selected parameter values fortokamak-like concepts that are typical for studies of this kind [29, 30]. The sensitivity
studies gave evidence that the main conclusions drawn in this paper are not sensitive
to the particular choice of parameters; exceptions from this are mentioned explicitly.
3.1. Reference parameter set
In the following, we give a set of reference data for the various parameters. The
range of values used for the sensitivity studies is given in brackets, { }. The cases
for which no strong dependence on the variation of the parameters was found are notmentioned in Section 4. When a sensitivity study pertinent to a parameter has beenreported elsewhere, the reference number is given in parentheses.
— Magnetic field strength: B = 10 T; utilization jS = 20% {5%}.
— Particle confinement: T = 2 s and scaling with temperature according to
T = const, {various scaling laws [4]}; ash ions: same confinement;
impurities: none {0-5% of ion density, A = 2Z = 12}.
— Energy confinement:• kinetic energy of escaping charged particles: 0.045 kT per lost particle;• synchrotron loss: reflectivity 95% {0.90-0.99}, plasma dimension 2 m;
formula from Ref. [17] for cylinder geometry {formulae from
Refs [14-18]};• other losses (heat conductivity, line radiation, etc.):
no loss {1-1000 s confinement time}.
— Interface between suprathermal ions and Maxwellian ions: 3 kT.
— Heating: radiofrequency heating; all energy to ions {60-100% to ions, rest toelectrons}.
— Energy conversion efficiencies:• thermal to electric 35%, electric to RF 35%;
• blanket power multiplication: 1.3 for 14 MeV neutrons,
4.0 for 2.4 MeV neutrons;• auxiliary power consumption (pumps, coils, fuel cycle, etc.): 10%.
— Fuel cycle:• efficiency for recovery of tritium from blanket: 98% {95% in Ref. [4]};• efficiency for recovery of 3He and T from plasma exhaust: 99%{95% and 98% for T in Ref. [4]}.
3.2. Fuel cycle description
For the fuel cycle description the plasma simulation code offers two options:— One option is to use the fuelling rate of the various fusile isotopes as datum.
Since the density of the plasma is assumed to be /3B2 limited, only the fuelling380

KERNBICHLER et al.

rate ratios may be fixed by the user of the code, for instance IT/ID as the relative fuelling rate of tritium to that of deuterium.
— The other option is to assume that for some or for all of those ion species which
are generated in the plasma the injection rate is determined by the rate at which
the respective ions are lost from the plasma, accounting for the efficiency withwhich this isotope can be recovered and recycled.
With these fuelling options the plasma simulation code determines self-consistently
the associated density ratio in the plasma. It turns out that in all cases considered here
the density ratio nT/nD is equal to the fuelling rate ratio IT/ID for tritium-rich plasmas
and that this density ratio is only slightly lower than the fuelling rate ratio for tritiumlean plasmas. Therefore, in the following we present our results in terms of the injection rate ratio, which can be approximately equated to the associated density ratio.
4.

RESULTSWe discuss here three areas of interest: (1) the relaxation of the tritium breeding
requirement in the blanket which is associated with the reduced tritium fuelling of
the plasma; (2) the fusion power density and the net power density associated withthis reduction; and (3) the power associated with, and the energy source distributionof, fusion neutrons as a function of tritium availability and plasma temperature.
4.1. Tritium breeding requirement
A tritium to deuterium fuelling rate ratio can only be imposed if the associatedsupply of tritium is ensured by breeding in and recovery from the blanket surrounding
the plasma. Therefore, for each plasma and confinement parameter set there is an
unambiguous relation between tritium fuelling rate and associated breeding requirement or, alternatively, between a given blanket breeding capacity and the relative
T/D density ratio that can be supported.
Figure 3 shows this correlation for a typical case. The important result to noteis that even a marginal decrease of the breeding capacity of the blanket to a value
below that required for a 1:1 D-T plasma results in a dramatic decrease of the tritiumcontent in the plasma (a 10% reduction of the breeding ratio results in a reductionof the T/D ratio by an order of magnitude). The apparent decrease of the blanket
breeding requirement with increasing plasma temperature can be fully explained by
the increasing importance of tritium breeding D-D fusion reactions in the plasma.
However, this bonus becomes effective only at tritium density ratios well below 1:1.
These results are independent of the extent to which 3He is recycled.
It can be concluded that even a modest decrease of the blanket breeding capacity
below that required for 1:1 D-T plasmas results in a dramatic decrease of the tritiumcontent in the plasma. This suggests the necessity of an investigation of the resulting
effect on important plasma parameters such as power and neutron production.

381

IAEA-CN-47/H-III-9

1.5

10

20
50
100
Ion Temperature, TV (keV)

FIG. 3. Overall blanket breeding ratio required to support the imposed T/D fuelling rate ratio (1/1,
1/2, 1/4, 1/8, etc.). Note that the actual local breeding rate must be higher in order to compensate forthe blanket coverage factor, and that BT = 0 defines the D-D cycle. The results are obtained for thereference parameter set.

1000

100

10

I

1

=

0.1

n

1

i i i 11111

i

5
10
Ion Temperature, T-

i v\ 11

50
100
(keV)

FIG. 4. Fusion power densities associated with various T/D fuelling rate ratios, with and without 3He
recycling (dashed line and dash-dotted line, respectively} for 0 = 20%. T/D: (a) 1/1, (b) 1/4, (c) 1/16,(d) 1/64, (e) 1/256, if) zero. At the left margin, the maximum power density values associated with thisset of fuelling rate ratios are indicated for /3 = 5% so that the effect of tritium reduction can becompared with the effect of an increase of the field utilization beta. Note that the dip in the curves around20 keV is due to the effect of accumulation of alpha particles.

382

KERNBICHLER et al.
I

1/1

I

I I I U

1

0 4

1

1

I

1

1

(b)
1/2

II

i ..

Beta=20*

>H
P«

^,

4->

:ion fOJ
c 0.3 (^

-

0.2 -

H

_

<U
O

0.1 -

Resoui

<d
M
H

-

0.0 -

1

5 10
50 100Ion Temperature, T^ (keV)

Beta=St, 3He=0'iBeta=20t, 3He=<n(I

1 1 i>

T/D Fuelling Rate Ratio,

FIG. 5. (a) Temperature window for ignited operation and for net electric power output for various
T/D fuelling rate ratios. The dotted line gives the temperature dependence of the T/D fuelling rate ratio
corresponding to zero blanket breeding and a 99% recycling of both T and 3He (CAT-D operation) at
0 = 20%. The notation 3He = 0% and 3He = 99% indicates that 0 and 99% of the leaking 3He is
recycled.
(b) Net electric power that can be supplied to external customers per unit fusion power produced in theplasma, for various T/D fuelling rate ratios.

4.2. Fusion power density, ignition, net power density
The fusion power densities associated with various tritium to deuterium fuellingrate ratios with and without 3He recycling is shown in Fig. 4. It is interesting tonote that 3He recycling does not increase the maximum power density, although it
will prove to be vital for the energy balance. Using a combination of Figs 3 and 4,
the penalty in power density connected with the relaxed breeding performance of theblanket can be calculated. For example, a reduction of the breeding capacity by 50%
is found to reduce the power density by an order of magnitude; this is approximately
the amount that could be recovered if it were possible to increase the magnetic field
utilization from 5% to 20%. On the other hand, a reduction of the blanket breeding
capacity BT by only 10% entails a reduction of the power density by as much as afactor of five.
However, the effect of a reduced tritium inventory is much more severe thancan be seen from Fig. 4. This becomes evident from Fig. 5. Figure 5(a) gives, foreach fuelling rate ratio, the temperature window within which ignition is possible orwithin which net power is supplied to the electrical grid. Clearly, the difference
between the condition for ignition and the condition for net power production is not

383

IAEA-CN-47/H-IH-9
1.0

i

i

i

i i

100

in

i

i

i

i

i

i

(a)
'0.8

c

"(a)

80 .

(b)
-(c)

0.6

-•xwv •-(d)

60 .

0.4

-(e)
(f)
(g)
(h)
(i)

40 -

0.2 .

0.0

TTT—

-]—I

2

20-

I I 1110

so 100Ion Temperature, T. (keV)
T/D Fuelling Rate Ratio,
FIG. 6. (a) Fraction offusion power carried by neutrons as a function of ion temperature for various
T/D fuelling rate ratios. T/D: (a) 111, (b) 118, (c) 1/16, (d) 1/32, (e) 1/64, (f) 1/128, (g) 1/256,(h) 1/512, (i) zero. The dashed curves were obtained for 99% recycling 3He. The dotted line indicates
CAT-D operation (T = 99%, 3He = 99%, no T breeding in blanket).
(b) Neutron spectrum, consisting mainly of 2.4 MeVand 14 MeV neutrons with negligible contributionof other energies, as a function of T/D fuelling rate ratio and ion temperature.

very pronounced. Furthermore, the operating temperature window closes for thosetritium-lean fuels which correspond to blanket breeding ratios below about one-thirdof the value required for 1:1 D-T operation, unless 3He is recycled. If impurities
are taken into account (here A = 2Z = 12) a dramatic change to the worse may
occur for tritium-lean fuels. For CAT-D plasma (the corresponding injection rate fortritium is indicated in Fig. 5(a) by the dotted line) there is no ignition at 2% impurity
concentration and no net power at 3% impurity concentration.
4.3. Neutron power fraction and spectrum
As the tritium inventory in the plasma decreases relative to the deuterium inventory, the fraction of fusion power carried by neutrons decreases, as shown in
Fig. 6(a). This decrease is more pronounced when 3He is recycled, since theneutron-free D- 3 He fusion strongly contributes to the overall power, in particular athigh temperatures. Correspondingly, as shown in Fig. 6(b), the contribution of low
energy neutrons increases, at the expense of the 14 MeV D-T neutrons, with increasing ion temperature and decreasing T/D fuelling rate ratio. Clearly, these are twocompeting effects: when the neutron power fraction decreases, the neutron spectrum
softens and this may result in an increase in the number of neutrons per unit power.

384

KERNBICHLER et al.

ACKNOWLEDGEMENTS
The authors acknowledge the contribution to their nuclear elastic scattering
calculations by G. Heinrichs, and the very efficient computational support by
C. Buchegger and K. Malli.
REFERENCES[1] WITTENBERG, L.J., SANTARIUS, J.F., KULCINSKI, G.L., Fusion Technol. 10 2 (1986)
167.
[2] AMERICAN NUCLEAR SOCIETY, Trans. Am. Nucl. Soc. 52 1 (1986) 658.
[3] Session on Blanket Technology, in Fusion Technology (Proc. 14th Symp. Avignon, 1986),
Pergamon Press, Oxford (1987).
[4] KERNBICHLER, W., FELDBACHER, R., HEINDLER, M., "The promise of tritium leanfuels", ibid.
[5] KERNBICHLER, W., HEINDLER, M., "D-based fuel cycle in MCF-reactors: Trade-offs
between neutron-radioactivity and power-production", in Emerging Nuclear Energy Systems(Proc. 4th Int. Conf. Madrid, 1986), World Scientific Publishing, Singapore (1987).
[6] KERNBICHLER, W., HEINDLER, M., "Compound light ion fuel cycles: An approach to
optimization", in Alternative Energy Sources (Proc. 7th Int. Conf. Miami, FL, 1985), Clean
Energy Research Institute, Miami (1987).
[7] NASSRI, A., HEINDLER, M., Phys. Fluids 29 (1986) 3275.
[8] NAKAO, Y., OHTA, M., NAKASHIMA, H., Nucl. Fusion 21 (1981) 973.
[9] PERKINS, S.T., CULLEN, D.E., Nucl. Sci. Eng. 77 (1981) 20.
[10] DAWSON, M., CTR using the p-"B Reaction, Rep. EPRI-ER-536-SR, Electric Power
Research Inst., Palo Alto, CA (1977) 403.
[11] McNALLY, J.R., Nucl. Technol./Fusion 2 (1982) 9.
[12] TAMOR, S., Nucl. Technol./Fusion 3 (1983) 293.
[13] GORDON, J.D., SAMEC, T.K., FREUE, S.A., HAUSS, B.I., Nucl. Technol./Fusion 4 (1983)
348.
[14] ROSE, D.J., On the Feasibility of Power by Nuclear Fusion, Rep. ORNL-TM-2204, Oak Ridge
National Lab., TN (1978).
[15] YANG, T.F., FORSEN, H.K., EMMERT, G.A., Rep. UWFDM-49, Dept. of Nuclear
Engineering, Wisconsin Univ., Madison (1973).
[16] CANOBBIO, E., GIUFFRE, S., Nucl. Fusion 10 (1970) 189.
[17] BORNATICI, M., CANO, R., DeBARBIERI, O., ENGELMANN, F., Nucl. Fusion 23 (1983)
1153.
[18] TRUBNIKOV, B.A., "Universal coefficients for synchrotron emission from plasma configurations", Review of Plasma Physics (LEONTOVICH, M.A., Ed.), Vol. 7, Consultants Bureau,
New York (1979) 345.
[19] TAMOR, S., Nucl. Fusion 23 (1983) 1704.
[20] FELDBACHER, R., The AEP-Barnbook DATLIB, Nuclear Reactor Cross Section andReactivity Parameter Library and Files, Rep. AEP-86.007 (1986).
[21] ECPL-82: The LLNL Evaluated Charged-Particle Data Library, Nuclear Data Section,
Documentation Series IAEA-NDS-56, IAEA, Vienna (1983).
[22] EXFOR: Exchange Format Nuclear Data Library, Nuclear Data Section, Documentation Series
IAEA-NDS-1, 2, 3, IAEA, Vienna (1983).
[23] MILEY, G.H., TOWNER, H., IVICH, N., Fusion Cross Sections and Reactivities, Rep.
COO-2218-17, Chicago Operations Office, Department of Energy, Chicago, IL (1974).

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[24]
[25]
[26]

[27]

[28]

[29]
[30]

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McNALLY, J.R., Jr., ROTHE, K.E., SHARP, R.D., Fusion Reactivity Graphs and Tables for
Charged Particle Reactions, Rep. ORNL-TM-6914, Oak Ridge National Lab., TN (1979).
FOWLER, W.A., CAUGHLAN, G.R., ZIMMERMAN, B.A., Ann. Rev. Astronomy
Astrophys. 13 (1975) 69.
HOWERTON, R.J., An Integrated System of Production of Neutronics and Photonics Calculadonal Constants, Rep. UCRL-50400, Vol. 21A, Lawrence Livermore Lab., California Univ.,
Livermore (1979).
SHUY, G.W., CONN, R.W., "Charged particle cross section requirements for advanced fusion
fuel cycles", in Nuclear Cross Sections for Technology (Proc. Int. Conf. Knoxville, TN, 1979),
Special Publ. 594, National Bureau of Standards, Washington, D.C. (1980) 254.
HOLDEN, N.E., RAMAVATARAM, S., DUNFORD, C.L., The Bibliography of Integral
Charged Particle Nuclear Data, Rep. BNL-NCS-50640, Brookhaven National Lab., Upton, NY(1979).
HOLDEN, N.E., RAMAVATARAM, S., DUNFORD, C.L., Integral Charged Particle Nuclear
Data Bibliography, Rep. BNL-NCS-51771, Brookhaven National Lab., Upton, NY (1986).
EVANS, K., GILLIGAN, J.G., JUNG, J., Tokamak Fusion Reactors with Less than Full
Tritium Breeding, Rep. ANL/FPP/TM-169, Argonne National Lab., IL (1983).
BROMBERG, L., et al., Nucl. Technol./Fusion 4 (1983) 264.

IAEA-CN-47/H-m-lO

SUPERfflGH DENSITY OPERATING CONDITIONSFOR A POLOIDAL DIVERTOR
IN A TOKAMAK REACTOR
S.I. KRASHENINNIKOV, A.Yu. PIGAROV
I.V. Kurchatov Institute of Atomic Energy,
Moscow, Union of Soviet Socialist Republics

Abstract
SUPERHIGH DENSITY OPERATING CONDITIONS FOR A POLOIDAL DIVERTOR IN
A TOKAMAK REACTOR.
It is shown that, in an INTOR-type reactor, it is possible to go over to a 'gas blanket'type of divertor operating regime by increasing the plasma density at the separatrix to2 X 1014. The new situation emerging with the La radiation cut-off as a result of the highdensity of the plasma and the neutrals is discussed.

1. The establishment of a gas blanket regime in the divertor volume is probably
the most generally accepted solution to the problem of removing the energy
released in a tokamak reactor in the process of fusion. In this case, virtually all
the energy reaching the divertor plates is transferred by radiation, and almost
complete volume recombination of plasma takes place, ensuring minimal
sputtering of the plates. Moreover, 'shifting' the zone of radiative losses awayfrom the plates significantly reduces the heat load on them.
However, the results of 2-D calculations [1] show that the operating
conditions for the divertor plates in INTOR are rather tough. The heat fluxes•to the plates reach q 0 ^ 0.3 kW-cirT2, and a rather high plasma temperaturenear the plates (~10 eV) can result in their sputtering, contamination of themain plasma, reduced operating efficiency and diminished reactor lifetime.
In this paper we shall analyse the possibility of establishing a 'gas blanket'
regime in the divertor of an INTOR-type tokamak.
As the rate constant of three-particle recombination, which is dominant
in these conditions, has a strong temperature dependence (Kfp = 6 X 10" 2 8 /(TeV) 9 ' 2 cm^s" 1 [2], where T is the plasma temperature), the plasma temperature should be rather low (< 1 eV) for volume recombination to take place.
Such strong cooling can be brought about only by an increase in radiative
losses, either in a pure D-T plasma (for which the plasma pressure within thezone of radiative losses should be rather high) or by additing impurities. Thus,
plasma recombination processes and re-radiation of the energy entering thedivertor volume are strongly interdependent. Since impurity transport processes
387

388

KRASHENINNIKOV and PIGAROVFIG. 1. Geometry of the situation.

have not yet been studied throughly, and there is a danger of impurity accumulation in the main tokamak plasma [3], we limit our examination to the pure
D-T mixture.
An increase in the plasma pressure within the region of radiative losses, where
the conversion of the conductive heat flux into radiation occurs, leads to anincrease in plasma pressure and density at the separatrix, n s (Fig. 1), because of
fast equalizing of the plasma pressure along the magnetic field lines. As theradial plasma flux in the main tokamak volume is proportional to 3n/9r> 0 (r is thecurrent minor radius of the plasma), ns is limited by the inequality ns < n(0),where n(0) is the plasma density at the axis, r = 0 ; n ( 0 ) ! * 2 X 10 14 cm" 3 for
INTOR parameters. Variation in n s results in a change of the heat fluxes q k ,Qo (lo = Qk s i n ^s ; Fig- !)• The dependence of q 0 and other quantities ofinterest to us on the plasma density, ns, is examined in Section 2. The conductive heat flux re-radiation process is analysed in Section 3. Sputtering of thefirst wall in a tokamak by charge exchange neutrals in the 'gas blanket' regime
is studied in Section 4. The results are discussed in Section 5.

2. First of all, we are interested how the scaling of the plasma pressure at
the separatrix, Ps (Fig. 1) and the heat flux densities in the core, q k , where thebulk of the energy is transferred, depend on the density, ns, and on the power,
WQ, entering the divertor from the main plasma. Normalizing the results withrespect to the results of the 2-D calculation, we easily obtain the absolutevalues of the parameters of interest. The core width, A k , and the quantities Ps
and qk are determined by the transport processes in the core, away from theradiative region, where the density of neutrals in negligible, and the heat
conduction equation has the form:
V||K||V||T=0(1)

IAEA-CN-47/H-m-10389

where K. \\ <* T 5 ' 2 is the plasma heat conduction coefficient along the magnetic
field; K± = K^n b T c is the plasma heat conduction across the magnetic field;
the constant b and c are determined by the scaling law for Kj_.
Taking into account the boundary condition at the separatrix:
KL VLT ~ «xT s /A k = W0/(2;r)2 a R(R and a are the major and minor plasma radii), we obtain, from Eq. (1), thefollowing set of equations:£ i n b T c+l/A k =W 0 /(27r) 2 aR

(2)where L is the mean length of a magnetic field line in the core, and K\\ = n\\ / T s / 2 .
From Eq. (2), we find:(3)
Ps a (W£n(9+2c-2b))l/(9+2c)(4)

From the Bohm scaling law for KL, it follows from expressions (3) and (4) thatq 0 <x n ~ 7 / " , Ps cc nj/ 9 . From a scaling law K± <* n, we have q 0 a n7 7 / 9 .
Since, for n s *> 0.5 X 1014 cm" 3 , q 0 as 0.3 kW-cm"2 and P S * 1 . 5 X 1016 cnT3-eV= 0.02 atm [1], an increase in n s up to a level of ~2 X 1014cm~3 allows adecrease in q 0 to q 0 «* 0.1-0.12 kW-cirT2 in these cases; then,Ps » p m a x as 0.05-0.07 atm. However, with Alcator scaling (b = 0), q 0 does
not decrease as n s , but Ps « n s £ 0.08 atm. If a non-inductive current drive is used in
the reactor and Wo is increased by about 1.00 MW [1], i.e. virtually doubles
by comparison with the Ohmic regime, then qg will increase by three times,
and Ps by about 1.3 times.
3. Since plasma recombination, together with the conversion of the conductive heat flux into radiation, occurs at temperatures T R ~ 1 eV, the plasma density,n, and the neutral particle density, N, in this region of the divertor are of the orderof Pmax/3T;R ~ 1016 cm" 3 . As is well known [4], the main channel of radiative
losses in an optically transparent hydrogen plasma is radiation through the
L a line, at temperatures T ~ 1 eV. However, with rising plasma and neutral
densities, the L a line is 'cut off, and its relative contribution to all radiation
losses drops substantially. Let us estimate the degree of cut-off of the L a linein the case under consideration. The parameter of L a damping, j3r = n e k 2 i/T r(k 2 j is the damping rate constant of the first excited level in hydrogen by

390

KRASHENBVNIKOV and PIGAROV

electron impact and yr is the radiative width of the line) is foundto be ~0.1. The L a quantum range in the plasma, £r «» (27r(c/coa)2 (7 r /r)N)~ 1 ,determines the width F of the La line {oja is the radiation frequency). For
given parameters, the F of the L a line is represented [5] by the ion Stark
broadening (FjCs"1) = 40(nj (cm" 3 )) 2 '' 3 ) and by Doppler broadening(F D = 2co/c(2£n 2T/M)^ 2 , M is the atomic mass), which are approximately
equal: Fj *» 2 X 10 1 2 s - 1 , F D s= 10 I 2 s"'. In this case, £r «* 2 X 1CT2 cm. It canbe shown that the characteristic length of the re-radiation region, £ q , determinedby plasma heat conduction and heat flux, qg, is ~I cm. Then, the degree of L a
cut-off is determined by the parameter 0g = £ q /£ r (l + l/2jS r ) 1/5 [5], where
6 = 0.5 and 1 for dispersion and Doppler line shapes, respectively. Substituting
the estimates obtained into the expression for 0g, we find j8g «* 10-20.
For the H, P, etc. line series, since their radiation widths are so small, j3r isgreater than unity, but for the characteristic ranges, owing to the small
population of these levels, we have £r > £ q . Thus, for Ps ~ P m a x , the plasma
in the region where the conductive heat flux is converted into radiation is ina state of local thermodynamic equilibrium (LTE) [5].
Let us restrict ourselves to a temperature range of T 5= 0.5 eV, which allows
us to neglect the effects of negative ions and the molecular fraction. Then, themain channels of radiative losses are: radiation transport through the L a line (Q a ),radiation transport through the H, P, etc. line series (Qg), and recombination
radiation ( Q R ) . Assuming the detachment of the radiative region from the plates tobe small, ~1 cm, which is much less than the contact width between core andplate (~40 cm) [1], one can use a one-dimensional heat conduction equation:(5)
where £ is the co-ordinate normal to the plate surface and Kj is the heat
conduction coefficient for a plasma in the LTE state [6].
Since the photoionization cross-section of atomic hydrogen is given by
a\ < 6.3 X 10"18 cm"2 [2], the region of radiative losses, for £q ~ 1 cm, can beconsidered to be transparent to recombination radiation so that Q R , besides
Qg, is three-dimensional in character. Then, the expressions for Qg and Q R
have the form [6]:Q g = 4 . 3 X 1 0 s P s [ a t m ] ( l - 2 x e ) e - 9 ( e e / 9 - l - e / 9 ) / e i [kW-cm"3](6)

Q R = 175 P 2 [atm] X2e/(T (eV)) 5 / 2 [kW-cm"3](7)where 0 = I/T (I is the ionization potential of hydrogen), x e = n e T/P s is therelative electron pressure determined by Saha's equation.

IAEA-CN-47/H-in-10

391

The quantity Qa is determined by the expression [7]:Q a = /dftdcok'(co)[I p (co,ft)-I(a;,fi)](8)where k'(co) is the effective absorption coefficient of a quantum with frequency
frequency co; I(CJ,SI) is the intensity of the radiation of frequency CJ per unitof solid angle, and I p (o;,£2) = hco3/47r3c2 [exp(hco/T) - 1] is the corresponding
Planck intensity.
The quantity 1(60,£2) is found from [7]:- ip («,s5)]

dS

(9)

with appropriate boundary conditions (S is the direction of quantum propagation).
To obtain simple estimates, we neglect Q a in (5). Then, it is not difficult tofind the density of the energy flux reaching the divertor plates as a result ofconductive heat transport, qjj:
q3=q0(l-(qT/q0)2)1/2

(10)

where

Tj is the plasma temperature near the plate.
The temperature T d should be rather low and q£ < q 0 for divertor
operation in the gas blanket regime. These conditions, as follows from (10),can be satisfied forq o < w = (2As(QR + Q«)dTy/2

di)0
The dependence of q max (P s ) as calculated from Eqs (6 and 7), with datafrom [6], is shown in Fig. 2. Since q o ~0.1-0.12 kW-cm"2 (Bohm scaling forKi and jq a n) is even smaller than qmax(Pmax) f ° r *s ^Pmax = 0.05-0.07 atm
and since we neglect additional energy losses through the L a line, we may
conclude that it is theoretically possible to realize a gas blanket in the INTOR
divertor.

392

KRASHENINNIKOV and PIGAROVFIG. 2. <7max versus P.

In the case of Alcator scaling or when a non-inductive method of toroidal
current drive in a tokamak is used (where, as has been shown above, qg ~0.3 kW- cm" 2 )condition (11) is not satisfied for P ~ P m a x . However, we should allow for the factthat we restricted our examination to considering an approximate, one-dimensional
representation of the processes in the divertor. In real conditions, q k is non-uniform
and, since ~0.3 kW-cm"2 corresponds to a peak of qg [1], it is possible that amore correct allowance for the geometry of the problem for L a radiation transportand other factors lying outside the scope of this work will reduce the pressure,Ps, necessary for the gas blanket to be established to an acceptable level in thiscase, as well.

4. The LTE approximation, which is valid in the region of radiative losses,
ceases to be valid as we move away from the radiative region towards the main
plasma, where there is a considerably greater number of neutrals than in the
LTE region.
Meanwhile, a high neutral density in the gas blanket regime can result inhigher losses of neutrals both into the main plasma, F^ (Fig. 1), and onto
the first wall of the tokamak, leading to sputtering.
The following model was examined in order to study this problem. It was
assumed that the spatial distribution of plasma temperature and density in the
scrape-off layer is determined by a dense, hot core and a cool, relatively rarefied'coat'. This picture of the situation corresponds to the results of a 2-D simulationof the scrape-off layer [ 1 ]. The temperature distribution in the core is foundfrom the heat conduction mechanism for energy flux density transport,
q k W = Qk e *P(~(x/A k ) 2 , where x is the distance from the separatrix. It is
assumed that two-thirds of the heat flux are transferred by the electrons.
The plasma density in the core is found from the pressure distribution,
P = Po exp(-(x/A k ) 2 ) and from the temperature distribution. The plasma
temperature in the 'coat' is taken to be 1 eV, and its density varies within wide
limits. It is assumed that complete plasma recombination takes place at a core

393

IAEA-CN-47/H-in-10

KAk

T = 1 eV

FIG. 3. &x, 2 W , <7k> A k and 5 w rte plasma.

10n s h (10 1 3 cm- 3 )20
n $ h (10 1 3 cm- 3 )F/G. 4. Sputtering rate versus n, h /or two different cases (a, b).

temperature of T = 1 eV and thus the pressure of the neutral gas produced
balances the plasma pressure; the neutral gas temperature is assumed to be 1 eV
(Fig. 3). The distances between the separatrix and the first wall, £ w , and between
the recombination zone and the effective x-point, £ x , are assumed to be 10 and
10 to 20 cm, respectively.
Neutral transport was claculated by a Monte Carlo method for theparameters Po = 0.05 atm, q k = 2 kW-cm" 2 , and Ak = 2 cm. The maximum
sputtering rate for the first wall and the flux T^ versus n sh are shown in Fig. 4.
It is easy to see that the sputtering of the first wall and F^ can be neglected1014 cm" 3 foreven when n sh <; 1.5 X 1014 cm"3 for Cx = 10 cm and when n sh
£x = 20 cm. Note that the values of n sh given here are consistent with the resultsof 2-D simulation.

394

KRASHENINN1KOV and PIGAROV

5. Thus, we have shown that it is possible to go over to divertor
operation with a gas blanket in an INTOR-type reactor, with Bohm scalingof K± and K± « n , increasing the plasma density at the separatrix up to ~ 2 X 10 14 .
More detailed studies are needed for Alcator-Iike scaling of K^ or for noninductive current drive in the tokamak. A new aspect, by comparison with thecalculations in Ref. [1], is the L a radiation cut-off due to the high density ofplasma and neutrals (~4 X 10 !6 cm" 3 ), which results in the emergence of local
thermodynamical equilibrium in the radiating plasma.

ACKNOWLEDGEMENTS
The authors thank V.I. Kogan, V.S. Lisitsa and V.A. Abramov for
discussions and useful advice.

REFERENCES
[1]
[2]
[3]

[4]
[5]

[6]

[7]

INTOR GROUP, International Tokamak Reactor, Phase Two A, Part II (Rep. Int.
Workshop Vienna, 1984 and 1985), IAEA, Vienna (1986).
ELETSKIJ, A.V., SMIRNOV, B.M., in Osnovy fiziki plazmy (GALEEV, V.A., SUDAN, R.,
Eds), Vol. 1, Ehnergoatomizdat, Moscow (1983) (in Russian).
ZHDANOV, V.M., Yavleniya perenosa v mnogokomponentnoj plazme (Transport
Phenomena in a Multicomponent Plasma), Ehnergoatomizdat, Moscow (1982)(in Russian).
BIBERMAN, L.M., VOROB'EV, V.S., NORMAN, G.E., Optika i spectroskopiya,
Vol. 14 (1963) 328 (in Russian).
SOLOUKHIN, R.I., YAKOBI, Yu.A., KOMIN, A.V., Opticheskie kharakteristiki
vodorodnoj plazmy (Optical Characteristics of a Hydrogen Plasma), Atomizdat,
Moscow (1975) (in Russian).
ELETSKIJ, A.V., PALKINA, L.A., SMIRNOV, B.M., Yavleniya perenosa v slaboionizirovannoj plasme (Transport Phenomena in a Weakly Ionized Plasma), Atomizdat,
Moscow (1975) (in Russian).
ABRAMOV, V.A., LISITSA, V.S., KOGAN, V.I., in Voprosy teorii plazmy, Vol. 12,
Ehnergoatomizdat, Moscow (1982) (in Russian).

Session I
FUNDAMENTAL PROCESSESAND NEW TRENDS(Posters)

IAEA-CN-47/I-M

IGNITION PHYSICS PROCESSESAND COMPACT EXPERIMENTS
B. COPPI, R. ENGLADE, S. MIGLIUOLO,
F. PORCELLI, L. SUGIYAMA
Massachusetts Institute of Technology,
Cambridge, Massachusetts,
United States of America
Abstract
IGNITION PHYSICS PROCESSES AND COMPACT EXPERIMENTS.
The main processes that, on the basis of present knowledge, are expected to determine the abilityof toroidal, magnetically confined plasmas to reach ignition conditions are the anomalous thermal transport of the electrons and the ions, and the effects of internal m° = 1 modes, epitomized by sawtoothand fishbone oscillations. To study these effects in compact ignition experiments, an anomalous electron
thermal transport model, based on the 'principle of profile consistency', has been used to reproduce theresults of an extensive series of Ohmic experiments on TFTR, in a 1-D transport code with a sawtoothoscillation model. As alpha heating becomes important, it is shown that for the ion anomalous thermaltransport resulting from the excitation of TJ; modes a minimum density is required so as to ensure that
it is offset by the alpha particle heating, once the temperature exceeds that where alpha heating compensates both the bremsstrahlung radiation and the collisional ion thermal conductivity. Results from al'/i-D transport code verify this result for time dependent ignition scenarios. 'Sharktooth1 oscillations
resulting from the centrally peaked alpha heating and fishbone oscillations due to resonant interaction
with the fusion alpha particle distribution can crucially affect ignition. For the latter, the resonant interaction with the alpha particles produces an effective ion viscosity which destabilizes the mode. The
special design characteristics of compact ignition experiments can be used to avoid or to minimize theeffects of these modes.

1.

INTRODUCTION

Compact ignition experiments are expected to reach substantial levels of fusion a-particle heating while requiring theleast extrapolation from presently explored regimes of toroidal
magnetic confinement. The degree of electron colllsionality,
exploitation of Ohmic heating and its favorable confinement
properties, and ignition at the lowest possible temperature withminimum thermal loading of the first wall all favor compact,high density designs [1] . Present results suggest that the mainprocesses that determine whether ignition can be attained are
the anomalous thermal transport of the electrons and the ions,and the effects of internal modes with poloidal mode number
m°=l, such as those associated with sawtooth and fishbone typeoscillations.

397

398

COPPI et al.

We investigate the effects of these processes on compact
ignition. First we present a model for the anomalous electron
thermal transport and the effects of sawtooth oscillations inpresent plasmas, and discuss general and specific implications
for the prediction of heating in future experiments throughscalings of the energy confinement time. When fusion a-particle
heating begins to take over from Ohraic heating, restrictions are
imposed on the plasma density by excitation of n^ modes that can
produce an anomalous ion thermal transport. Results of 1 1/2 Dtransport code simulations illustrate the combined effects ofthese processes on heating to ignition for a specific proposed
device, Ignitor. Conditions for the excitation of m°=l modes
are discussed as the driving force for sawtooth or "sharktooth"[2] oscillations of the central part of the plasma column andpossible precursor oscillations. Instabilities analogous to thefast ion fishbone oscillations of neutral beam injection experiments, in this case interacting with the fusion a-particle distribution, represent a different root of the same type of dispersion relation. Since compact ignition experiments can be
designed to ignite at relatively low values of B p , the onset of
m°=l modes, to the extent that they are driven by pressure gradient, can be avoided. Since these experiments envision highplasma currents (up to 10MA), another possible way to mitigate
the effects of these modes is to decrease the plasma current
after the Ohmic phase so as to reduce the size of the qil region.

2.

ANOMALOUS ELECTRON THERMAL TRANSPORT

Identification of the form of the anomalous electron
thermal conduction is crucial to the accurate prediction ofignition scenarios, particularly in the initial, Ohmic heatingphase. While the diffusion coefficient x e m a y be expected tovary as the underlying physical processes change, many of thedimensionless parameters describing DT ignition in compact
experiments have values similar to those in present experiments.
Nevertheless, extrapolation should be viewed with caution b e cause a given expression for x e may not produce energy confinement times that scale as a 2 /x e - Conversely, a known confinement
time scaling can be reproduced by many different forms of x •As an example, we have extensively investigated Ohmic discharges
produced in the TFTR machine at Princeton Plasma Physics Laboratory using a one-dimensional (radial) transport code. They canbe reproduced over a wide range of conditions [3] by a form ofX e that is based upon profile consistency [A], when sawtoothoscillations are approximated by periodic flattening of theplasma profiles T e , T ^ n, and, approximately, j z . The diffusion coefficient reduces to the same one previously used tomodel Alcator A, Alcator C, FT, and other machines [3,5]. The
experimental and simulation values of the central electron temperature T e o and the surface voltage V s adhere closely toIAEA-CN-47/I-M

399

scalings predicted from the expression for xe- ^ n the other
hand, the electron energy confinement time does not follow thepredicted scaling T E e ^n e 2 / 3 B " 1 / 3 Z e f f 2 / 3 as closely and, evenwhen the energy balance is dominated by the electrons, thetotal energy confinement time follows a different scaling,TjA
n e q a aR 2 , where a and R are approximately constant.
Sawtooth oscillations do not cause the difference in confinement time scalings in these cases, where the affected region
is at most r-a/2. They do contribute strongly to the electron
temperature profile width measured by T eQ /<T e >, where <Te>
represents the volume average, but turning off the sawtoothcrashes does not change the values of T E or Tge> or the adherence of T e o and V s to the predicted scalings.
On the other hand, in ignition experiments tg may beaffected by sawteeth if they exist and if, in addition, theyextend to larger radii.

3.

PATHWAYS TO IGNITION

In order to attain ignition conditions, it is desirable tomaintain the highest possible value of the confinement parameternx£. The dependence of Tg on the main plasma parameters, inferred on the basis of present experiments, implies operation
at relatively high densities where the collisional ion thermal
conductivity and bremsstrahlung radiation become important in
the global energy balance. At the "minimum ignition temperature" these losses are compensated by the energy deposited by
the fusion reaction products. Near this temperature and above,
the optimal density depends critically on the presence of modes
that enhance the ion thermal conductivity. We consider in particular the so-called n^ modes with transverse wavelengths
longer than the ion gyroradius [6], bearing in mind that theseresults are valid for any modes producing effective diffusion
coefficients with similar scalings. We refer to an approximate
model equation [7] valid near the center of the plasma column,
n 2 K(T)-n T 5 / 2 + (P-CeT) =0(1)

Here K(T)=F(T)-T1/f2, C e is nearly constant, and the dimension7
less quantity n indicates the density, T the temperature and P
the external heating power density. In addition,F(T)°:<apV>,
Op being the fusion cross-section, and T=l corresponds to theminimum ignition temperature. The term n 2 K(T) includes thea-particle heating and the loss due to bremsstrahlung and collisional ion thermal conductivity, while n T 5 ' 2 represents theloss due to Hi-mode induced ion thermal conductivity, and C e T
that part of the anomalous electron thermal conductivity thatcan be represented by the Alcator scaling Xe^l/n.

400

COPPI et al.

0

I

2

3

4

f

FIG. 1. Allowed plasma density range as ignition is approached.

In the absence of an anomalous ion thermal conductivity,
there is a maximum density limit on the plasma heating before
the minimum ignition temperature is reached, while beyond thistemperature no forbidden density range exists. If an anomalous
conductivity exists, a critical external power P must be supplied to open a channel between the allowed density regions at
T<1 and T>1. Figure 1 illustrates a case where P is below critical and the passage is not open. Here, F(T) has been taken
to be proportional to T 5 ' 2 . Notice that P is proportional tothe anomalous ion thermal conductivity. Another point shown by
this analysis is that for $>1 the density has to be progressively
adjusted above a certain minimal value in order to compensate theanomalous ion thermal loss when the actual form of <a-pv> is used.
Complementing the analytic work, we have also performed
extensive numerical simulations of ignition scenarios using a
1 1/2 D version of the BALDUR transport code [8]. A set ofparameters, taken from the actual design of the compact ignition
experiment Ignitor, are R o =111.5 era, a=41 cm, b=72 cm, B^ 0 =125kG,
Ip=10.A MA, and n o %10 15 cm"" 3 . In addition to an anomalous ion
thermal diffusion coefficient X±=i-±\1^2
T ± 3 / 2 B T ~ 2 R ^ h , wheref^ is a numerical factor, h(n.£,ric) is a threshold function computed at the plasma half-radius, and A^ is the fraction of theplasma cross-section within the flux surface ijj, a neoclassical(Chang-Hinton) ion thermal conductivity was assumed. The electron anomalous thermal transport was assumed to have a CoppiMazzucato-Grflber form with an additional degradation factor depending upon B p whose magnitude was chosen to match neutral injection confinement times on PDX and D-III. The results indicatethat Ignitor is able to achieve DT ignition with only Ohmic
heating, if the anomaly factor f^ does not exceed that needed toreproduce recent D-III results [9] with the assumed form of Xi
and the threshold n c is not too small (£2), provided also thatsawtooth oscillations are infrequent and the plasma remains
relatively clean. Scans of n e and f^ confirm the basic analytical properties of the density limit, taking profile variations
into account. Another expression for the ion anomaly based on
the properties of the instability parameter USn^P^/r-r
been investigated.

IAEA-CN-47/I-M

401

The 1 1/2 D calculations support previous ID studies ofmany different ignition scenarios [10] that found a strong dependence of the heating on the spatial profiles of the plasma
parameters and their evolution in time. The maximum Ohmic power
deposition shifts away from the magnetic axis in the presence ofsawtooth crashes or strong central heating, increasing theheating for a given value of the central electron temperature,and increasing the likelihood of nonmonotonic current and q
profiles.
4.

THEORY OF INTERNAL OSCILLATIONS

The crash phase of sawtooth oscillations is closely associated with instabilities of poloidal mode number ra°=l that
represent one branch of the general m°=l dispersion relation,
the other branch describing modes associated with fishbone typeoscillations [11]. The first requires finite resistivity and isstabilized by ion viscosity, while the second is destabilized byan effective ion viscosity provided by the resonant interaction
with fast ions. For the sawtooth branch, the linear theory inthe two-fluid collisional limit provides the basic framework fordeveloping theory applicable to compact ignition. It predicts avariety of regimes [12] , presented schematically in Table I and
Fig. 2. The relevant parameters are Air=-<5W_..-ri/£_,1 ' 3=-Tr6Wm -_/[e n i / J(CcorBQq')z] and <D A =U> A1 / ( C D ^ 1 ' 6) , evaluated at r=rQ whereq(r o )=l, <SWmin is relevant energy principle quadratic form, £„,is the^radial displacement at r=0, e =nnc2/4Trr2)<iJA'VLCr9 -10~ s ,w =v
A A8 s/ ' r o' V Ae =B 6/^ Trm i tl i)> s=d£nq/d£nr is the shear parameter, w ^ is the ion diamagnetic frequency and, for simplicity,
w. =co. J--W . „. The figure assumes a simple model for 6Wm,-r, of
.

v^rt

-I

mill

the type [13]<5W . =-A (r / R ) 2 ( 6 2 - 3 2
. )
min
o o
p
p.cnt'(2)'v

with X Q a finite numerical factor and g the poloidal beta. The
critical 3p )Cr j: t for ideal MHD instability depends on the details of the plasma shaping, anisotropy, and the q and p profiles. We argue that regimes corresponding to regions II, III
and IV of Fig. 2, i.e., near the ideal MHD marginal stability
threshold, are the most likely candidates to cause the crash
phase of the sawtooth. The relevant modes can be considered infact more violently unstable than the "drift-tearing" mode [14]corresponding to regions V and VI. The indications of thelinear analysis are supported by the type of conclusions reachedin Ref. [15], for the nonlinear phase, although these arelimited to the strongly collisional regime (k|IXmjpEk;|v£he/ve<l) •No unstable m°=l modes exist when wA>-XH>aivt-1' 2 (region VII).
In fact, a numerical evaluation of XJJ using TFTR profiles froma ID transport code, shows that X H is negative and 0(1) near
the top of a sawtooth ramp.

402

COPPI et al.
TABLE I. m° = 1 MODES FROM THE TWO-FLUID THEORY

Instability regime

Growth rate
Y

I. Ideal kink

Oscillation frequency
u)'

X w
H A

II. Ion-Larmor-radiusstabilized ideal
regimeIII. Resistive kink

IV. Resistive kinkdiamagnetic limit

e i/V
ri
A

(u. /3)[1-(T /T )*ex ee OJ /oi.
n A *-V. Reconnectingmode

VI. Drifttearingmode.
(E o> 5/3/|xJ^/3 2 / 3 )
n A
H'
*

Definitions:*eA ^ - f i W ^ , ^=0.^^-0.^(>0) , "'=w: mode frequency in LAB framefi:plasma rotation frequency

The entry into an unstable regime with a finite frequency
may determine the appearance of precursor oscillations. For
example, when |XJJ|<1 and the plasma rotation frequency |f2rotl
is small, the m°=l mode is nearly purely growing in thelaboratory frame of reference and no precursor oscillations
would be observed.
At relatively high temperatures where a>.[a:n^'^ T^'^ B~^'^r Q ~ / ]>l, the linear theory predicts a significantly reduced
growth rate. If also -u A ~ 1 / 2 <X H <w it (regions II and IV), the2

3

IAEA-CN-47/I-I-1

403

FIG. 2. Asymptotic regimes, in the X^-u. plane (kH s X^), for m° = 1 modes predicted by thetwo-fluid theory. The functional relation between \H and other physical parameters (u,, uA, ej isindicated near each curve.

transverse viscosity \iL can stabilize [16] the indicated modes[17], a feature that is common for modes where the perturbed B r
is not constant [16] in the transition region. The
viscosity becomes important when Y ^ ^ i i ^ p|, v ^ being the ionion collision frequency, p^ the ion gyroradius and k^. indicatesthe order of magnitude of the eigenfunction derivative in thetransition layer. Its effect can be related to the dimensionlessparameter DEe^/e^/lO ( m ^ / r a ^ ) : / 2 ge(r=ro)^p^/nn , where£ =y1/(r2rn^n^u^). When k x p^<l, a simple linear viscous operator
yields the curves of marginal stability shown in Fig. 3; in thelimit of large u^, Dc^-it^H'"1*) scales as

crit(Note that D^0.20 for a 50-50 deuterium-tritium plasma with T e=T^ and 8e=l%) . When k x p ^ l , ion-ion collisions introduce more
complex operators [16], but on the basis of the conclusions
reached in Ref. [16] we may argue the stabilization found withthe simple viscous operator remains effective.
The influence of ion-ion collisions is not equally beneficial to the electron drift-tearing mode [12]. Under ignition
conditions, oj>V£^(kxP£)2, the ions can be considered collisionless. Therefore a normal mode equation [18] that is valid to
all orders in the ion Larmor radius, but ignores ion-ion collisions as well as ion Landau damping and treats the electrons in404

COPPI et al.

FIG. 3. Marginal stability, Im(u) = 0, curves for the m° = I mode which otherwise causes the sawtooth crash. The parameter D measures the relative contributions from ion-ion collisions (stabilizing)and electron-ion collisions (destabilizing).

the fluid limit (v e >u and k||A m fp<l), has been s o l v e d .
3In)•>£*' ,r_displaceme
the limit where p.>E^-/. the plasma r a d i a l displacement
becomes5(x) = (C 0O /2)| 1- [l-exp(-|x|/6 )](sgnx) (where x=(r-ro)/ro and 6p=6 o (p^/r o ), with 6 O a numerical factorof order unity that depends weakly on to and u.. The thicknessof the transition layer is of the order of p^, as was earlier
indicated [12,16] by the two-fluid theory.
These results have several consequences for the predictionof sawtooth oscillations, to the extent that the crash begins
when |XJJ|<1. First, sawtooth crashes may occur only when <SWm^nexceeds a (possibly negative) threshold value during the temperature ramp, as the pressure profile steepens. Second, sawteeth
may be suppressed even though q<l in part of the plasma. Third,
in high f3p plasmas when the m°=l instability is driven primarily
by the pressure gradient, disruptions may relax the pressure
profile only, leaving the q-profile unchanged.
Compact ignition experiments have the ability to cover arange of 6 . Thus, possible scenarios envision $ =0.2 whiler o £ a / 2 , or g =1 and r 0 considerably smaller. In the first casethe relevant xnstability conditions may not be met and in thelatter case the plasma volume where q<l may not be significant.
In fact, ignition conditions are characterized by k|I ^nfp^Px(s/roR)Xmfp>l, a regime that is not covered by the analysis
presented here, and, in the case of compact, high density plasmas(n o =10 15 cm- 3 , To=15 keV), v e >u^.

IAEA-CN-47/I-I-1
5.

405

FISHBONE OSCILLATIONS AND MeV PARTICLE PRODUCTION

Recently, a model for the excitation of fishbone oscillations [19] by energetic beam particles has been proposed [11].
The mode, whose poloidal structure is dominated by the component
with m°=l, has a frequency w = w d i , the ion diamagnetic frequency,
and is one of the two m°=l modes with ^•{{^en1
H"*^ that are
rendered marginally stable by finite ion Larmor radius effects[12]. The mode excitation energy is related to the plasmapressure gradient, which can be tapped in the presence of adissipative process. In the case of near perpendicular beam
injection, this process arises from the mode-particle resonancew
D n ( r > e >l J ) =tJ that scatters the trapped beam ions with energies
e=m^v 2 /2 and magnetic moments y=m^v^/2B, where w ^ ^ is theaverage magnetic drift frequency along the particle orbit. For
typical beam parameters, a) di /u) D ^K(T 1 /T h ) (R/r n ) (1+ni)vL; in PDX
experiments, Wj.^20 kHz, consistent with the measured value ofthe fishbone frequency [19]. In the realistic limit (ucj^/uA)>(r o /R)& p Yi> where 3 ^ is the fast-ion poloidal beta, we find a
growth rate [11]where w D h = u ^ ) ( r o , e i n - , U i n j ) . £±nj is the beam injectionenergy, and yi n j=£i n j/B o - Note that the instability requires
X H =-6W m i n >0 and that the electrical resistivity tends to damp it.
The same m°=l mode with w ^ w ^ can be destabilized by theLandau resonance, w=kj|V||, due to the energetic circulating
particles. This resonance is important near the q=l surface
where k|| is small. A preliminary analysis indicates thatthe mode has a positive growth rate, at least in regimes where
the effects of finite resistivity are unimportant. The growthrate is proportional to the magnitude of the pressure gradientof the energetic particles. The mode frequency is not anchoredto the average magnetic drift frequency of the circulating
particles. This result offers an explanation of recent observations [20] of fishbones that occur when neutral beams are
injected tangentially to the magnetic field.
In principle, circulating a-particles may also destabilize
an m°=l fishbone through the Landau resonance (o=k||V||. However,the necessary condition SW m ^ n <0 may not hold during most of thesawtooth ramp at the relatively low values of g characterizing
compact ignition experiments.

ACKNOWLEDGEMENTS
It is a pleasure to thank P. Detragiache, F. Pegoraro, andW.M. Tang for their effective collaboration. This work wassponsored in part by the U.S. Department of Energy.

406

COPPI et al.

REFERENCES
[1]
[2]
[3]

[4]
[5]
[6]
[7]
[8]
[9]
[10]
[11]
[12]
[13]
[14]
[15]
[16]
[17]
[18]
[19]
[20]

COPPI, B., Rep. PRR-75/18, Massachusetts Institute of Technology, Cambridge, MA (1975);
Comments Plasma Phys. Controll. Fusion 2 (1977) 47.
COPPI, B., TARONI, A., CENACCHI, G., in Plasma Physics and Controlled Nuclear Fusion
Research 1978 (Proc. 7th Int. Conf. Innsbruck, 1978), Vol. 1, IAEA, Vienna (1979) 487.
SUGIYAMA, L.E., MARTINELL, J., EFTHIMION, P.C., Predictive Transport Simulation of
Ohmic TFTR Discharges, Rep. PTP-87/1, Massachusetts Institute of Technology, Cambridge,
MA (1987).
COPPI, B., Fiz. Plazmy 11 (1985) 83.
GRUBER, O., Nucl. Fusion 22 (1982) 1349.
ANTONSEN, T.M., COPPI, B., ENGLADE, R.C., Nucl. Fusion 19 (1979) 641.
COPPI, B., TANG, W.M., Rep. PPPL-2343, Princeton University, Plasma Physics Lab.,
Princeton, NJ (1986), submitted to Phys. Rev. Lett.
BATEMAN, H., in Proceedings of the 1985 Trieste Spring College on Plasma Physics, International Centre for Theoretical Physics, World Scientific Press, Singapore (1987).
GROEBNER, R.J., et al., Nucl. Fusion 26 (1986) 543.
SUGIYAMA, L., Rep. PTP-84/2, Massachusetts Institute of Technology, Cambridge, MA(1984).
COPPI, B., PORCELLI, F., Rep. PTP-85/13, Massachusetts Institute of Technology,
Cambridge, MA (1985); Phys. Rev. Lett. 57 (1986) 2272.
ARA, G., BASU, B., COPPI, B., LAVAL, G., ROSENBLUTH, M.N., WADDELL, B.V.,
Ann. Phys. 112 (1978) 443.
BUSSAC, M.N., PELLAT, R., EDERY, D., SOULE, J.L., Phys. Rev. Lett. 35 (1975) 1638.
COPPI, B., Phys. Fluids 8 (1965) 2273.
HAZELTINE, R.D., MEISS, J.D., MORRISON, P.J., Phys. Fluids 29 (1986) 1633.
COPPI, B., ROSENBLUTH, M.N., in Plasma Physics and Controlled Nuclear Fusion Research(Proc. Conf. Culham, 1965), Vol. 1, IAEA, Vienna (1966) 617.
PORCELLI, F., MIGLIUOLO, S., Phys. Fluids 29 (1986) 1741.
PEGORARO, F., SCHEP, T.J., Plasma Phys. 28 (1986) 647.
McGUIRE, F., et al., Phys. Rev. Lett. 50 (1983) 891.
HEIDBRINK, W.W., et al., Phys. Rev. Lett. .57 (1986) 835.

IAEA-CN-47/I-I-2

A NEW APPROACH TO NUCLEAR FUSIONUSING A CO 2 LASER
S. NAKAI, Y. KITAGAWA, H. DAIDO, H. FUJITA,
K. SAWAI, M. FUJITA, K. NISHIHARA, K. MIMA,
A. HASEGAWA1, C. YAMANAKA
Institute of Laser Engineering,Osaka University,
Suita, Osaka, Japan

Abstract
A NEW APPROACH TO NUCLEAR FUSION USING A CO a LASER.
The feasibility of a hybrid fusion scheme — magnetically insulated inertially confined fusion —has been demonstrated by experiments with the CO2 laser LEKKO VIII. In these experiments, afocused laser beam of 300 J in 1 ns is injected into a cavity of 1-3 mm in diameter through a small
hole and irradiates the inner surface of the cavity. A hot plasma core is formed at the centre of the cavity.
The measured temperature is 300-600 eV and a self-generated magnetic field of 100 T is obtained. Thelong confinement of the laser produced plasma in the cavity is studied by theoretical and computational
analyses. The plasma pressure is confined by the inertia of the heavy spherical shell and thermal conduction loss is inhibited by the self-generated magnetic field. The results show that it is possible to achieve
ignition with a reasonable laser energy and a simple target.

1.

INTRODUCTION

For the two main approaches to controlled thermonuclear fusion currently
pursued — inertial fusion and magnetic fusion — there are differences in plasma
density and confinement time of twelve orders of magnitude. The new scheme —magnetically insulated inertially confined fusion (MICF) [1] — constitutes a different
approach to fusion, with a plasma density of about 1021 cm' 3 , which is thegeometric mean of the densities achieved with the two main approaches. Combinations of plasma density and confinement time for different approaches are shown in
Fig. 1.
The basic structure of the plasma container consists of a spherical metallic shellwith an outer radius c and an inner radius b; solid D-T fuel is contained in the part
between radius a and radius b, as shown in Fig. 2(a).
The gas plasma at r < a is produced by ablation of the solid fuel caused by laser
beam injection through the hole on the top of the reactor chamber. The laser createsa toroidal magnetic field through the current loop produced by the ejected hot electrons. The magnetic field expands rapidly because of the E x B drift of electrons in
AT&T Bell Laboratories, Murray Hill, NJ 07974, USA.

407

408

NAKAI et al.

Inertia! Fusion

N

Hybrid Fusion(MICF.DCT)
Inertial
Confinement
Pinch
Stellarator,
TokamakConfinement Time (s)FIG. 1. Combination of plasma density n and confinement time r to achieve nr ~ 1&4 s-cm'3 forvarious approaches.

the inner part of the shell, as shown in Fig. 2(b). A large number of experiments[2, 3] gave evidence of laser produced magnetic fields of more than 102 T in an areamuch larger than the size of the laser spot. When the inner surface of the solid fuelis heated by electron deposition and multiple photoreflections, plasma is ejected; itcarries the magnetic field from the surface by ablation and a hot low-density core isformed in the central region of the shell. The main feature of the MICF scheme isthe utilization of both a magnetic field (for confinement of the plasma temperature)and the inertia of the metallic shell (for confinement of the plasma pressure) by acannonball effect.

2.

EXPERIMENTAL RESULTS

Experiments have been performed to check the confinement time, using the
LEKKO VIE CO2 laser system [4], which delivers 300 J in 1 ns, and spherical targets of parylene (C8H7C1), with 15 /tm thickness and diameters ranging from 1 mm

IAEA-CN-47/I-I-2

409

Heavy Metal Shell
FIG. 2(a). Schematic diagram of the reactor chamber. The chamber contains a hot gas fuel at r < a,which is thermally insulated by a magnetic field and used for ignition, a solid fuel at a < r < b, which
is used to increase the q-value of the reaction, and an electrically conductive heavy metal shell at
b < r < c whose inertia is used to maintain the magnetic field and the plasma pressure.

LASER

FIG. 2(b). Process of magnetic field generation. Electrons that are knocked out by the laser beam
E x B drift along the inner surface and produce the current that amplifies and maintains the initial
magnetic field. Because of their large Larmor radii, ions are simply accelerated radially inwards andproduce a high temperature.

410

NAKAI et al.

FIG. 3. X-ray pinhole image (a) and X-ray signal (~0.1 keV) (b), produced in an experiment with a
two-hole target of 2 mm diameter. The sweep time in (b) is 2 ns/div. The existence of a hot core witha large energy confinement time f = 10 ns) is clearly visible.

to 3 mm [5]. The average intensity of the laser at the focal point is 2 x 1014 W/cm2
and the hot electron temperature is estimated to be around 15 keV [6]. The selfgenerated magnetic field is expected to be more than 102 T at this laser intensity [3].
The energy lifetime of the plasma is measured by the duration of the soft X-ray
signals (0.1-1 keV) emitted from the plasma. The plasma size is measured with an
X-ray pinhole camera and the temperature is measured spectroscopically, using thesodium line intensity ratio for He and H.
Four types of target were studied. The first type has one hole through which
one laser beam is radiated (see Fig. 2). The second type has two holes through which
two 100 J laser beams are radiated at an oblique angle. The third type has a toroidal
coil, which is used to produce a poloidal magnetic field by irradiation with anotherCO2 laser. This target is designed to form a spheromak type magnetic field configuration. The fourth type of target is the same as the first type, except that it is enclosedwith a gold shell.
Figure 3 shows a typical pinhole camera image and the profile of the X-ray
signal from the target. The example shown here is a two-hole target with 2 mmdiameter. The existence of a hot core with approximately 1 mm diameter can beclearly seen in Fig. 3(a). The first peak in Fig. 3(b) corresponds to the radiation that

IAEA-CN-47/I-I-2

411

1
2
Shell Diameter (mm)FIG. 4. Observed plasma lifetime as a junction of shell diameter for irradiation of a one-hole target
with a 100 J C02 laser (solid triangles). The linear dependence of the confinement time on the shell
diameter is visible. The enhanced confinement time in the spheromak type magnetic field configuration(open triangle) is also shown.

is due to the direct impact of the 1 ns laser beam, while the second, broader peakcorresponds to the radiation from the hot core plasma produced by ablation. The confinement of the hot core plasma for a duration of about 10 ns is clearly visible. The
duration of the soft X-ray signal is much longer than that obtained with a directly
irradiated target, the typical X-ray duration of which is comparable to that of the laser
pulse [7]. The measured plasma temperature is typically 300—400 eV.
Figure 4 shows the observed plasma energy lifetime as a function of shell
diameter for irradiation of a one-hole target with a 100 J CO2 laser. The lines correspond to the entire width of the X-ray signals (solid circles), the width of the second
peak of the X-ray signals (triangles), and the separation between the two peaks of the
X-ray signals (squares). We note the linear dependence of the confinement time on
the shell diameter. This is due to the fact that both ablation density and plasma pressure are approximately proportional to the laser energy per plasma volume and thefact that the sound speed becomes approximately independent of the plasma diameter.
The solid squares correspond roughly to a/cs (shell radius divided by the soundspeed), giving a thermal electron temperature of approximately 600 eV.
Figure 5 shows the width of the soft X-ray signals as a function of tamper shell
diameter for irradiation of a two-hole target with a 200 J CO2 laser. Two laser
beams irradiate the different regions inside the target. Thus, the laser intensity isalmost the same as in the case with a one-hole target and the sound velocity in the

412

NAKAI et al.
20

/

15
X-ray Signal
/

.

I
/
J

'

/ A
//

<0

c
w
>%
re

1
2
Shell Diameter (mm)FIG. 5. Observed plasma lifetime as a junction of shell diameter for irradiation of a two-hole target
with a 200 J C02 laser. The total width and the width of the secondary peak are seen to be approximately 30% larger than those obtained with the one-hole target shown in Fig. 4.

area of direct irradiation is almost the same for both targets. The values of the solid
circles and the solid triangles in Fig. 5 are systematically higher ( = 30%) than thosein Fig. 4, although the area of the holes in the two-hole target is twice that in the
one-hole target. The reason for this phenomenon is probably the more favourable
geometry of the self-generated magnetic fields. It can be seen that the X-ray widthis proportional to the shell diameter (Figs 4 and 5). An X-ray detector with a different
filter is also used to measure the X-ray signals (0.5-1 keV) simultaneously; signalsof almost the same shape are obtained. This indicates that the hot core temperature
is several hundred electronvolts during the emission time of the soft X-rays. In thecase of the targets with two holes, the thermal electron temperature is obtained fromthe intensity ratio of the helium-like and hydrogen-like sodium resonance lines which
are emitted from the seed material of NaCl in the shell, based on coronal equilibrium.
The measured plasma temperature varies from 600 eV at a shell diameter of 1 mm
to 450 eV at 2 mm and to 350 eV at 3 mm.
In our investigation of the dependence of the plasma lifetime on the magnetic
field configuration, we tested the modified one-hole spheromak type target [8]. Thistarget has a one-turn coil, made of 50 (xm thick copper, which surrounds the sphereazimuthally. The target is irradiated with a CO2 laser which has a time advance ofseveral hundred picoseconds, as shown in Fig. 6(a). The one-turn coil is 3 mm wideand has a radius of 2 mm and a drilled observation window. The measured decay
time of the magnetic field of this coil is longer than 10 ns and the peak magnetic field

413

IAEA-CN-47/I-I-2

X-Ray Pinhole
CameraA:112J

B:89J(b)
FIG. 6. (a) Geometry of the spheromak-type target.
(b) and (c) X-ray images for different poloidal magnetic fields produced by the laser energy (B).
The irradiation energy of the spherical target is also indicated (A). Target (c), which has a laser energy
on the one-turn coil two times that of target (b), shows a much brighter image.

at the centre of the circle is 50 T at a laser energy of 100 J; a search coil has beenused for these measurements [9]. Strong poloidal and toroidal magnetic fields exist
in the shell target. Typical X-ray images of the spheromak type target are shown in
Figs 6(b, c). The irradiation energy of the spherical target is almost the same in bothcases, but the irradiation energy on the one-turn coil of target (c) is two times thatof target (b). The intensity of the X-ray image of the hot core in (c) is two times thatin (b). This result demonstrates the importance of the poloidal magnetic field
and of plasma confinement. A comparison of two-band soft X-ray signals(Ch.l: 0.1-1 keV, Ch.2: 0.5-1 keV) shows that the spheromak type target has an
about 50% larger X-ray width and a 50% longer amplitude than the one-hole target.
This indicates that a hotter plasma with a longer lifetime is generated in the spheromak type target. The open triangle in Fig. 4 indicates the hot core lifetime of thespheromak type target. This longer lifetime gives experimental evidence of the magnetic thermal insulation of the core plasma.

414

NAKAI et al.

A scaling of the observed lifetimes of the plasmas can be obtained on the basisof the mass ablation rate [10]. If the lifetime is determined predominantly by thepiston speed of the expanding shock, the confinement time is proportional toWa Vps/P, where ps is the density of the shell material and P is the plasma pressurein the cavity. In the present experiment, the plasma pressure and the density of theablated tamper shell scale as
P = 3rtabEL/(4m3) (g-cm-s2)
ps - (2.6/a272) (EL/47r)0-86 V°J6 (g-cm"3)where EL is the irradiation energy in kJ, a is the cavity radius in mm, and r]ab is theabsorption efficiency. These relations give a confinement time of 10 x a115 ns,
which is in reasonable agreement with the experimental results. We can explain theconfinement time simply by the plasma expansion time.
The dynamics of the cavity plasma have been analysed with the spherically symmetric one-dimensional fluid code HISHO [11] in which a uniform magnetic field of
103 T is assumed in the cavity and the electron thermal conduction is reduced by afactor of (o)ceTe)"2. A typical example of the simulation results is shown in Fig. 7.
Figure 7(a) shows the streamline contour (flow diagram) of the plasma. It
clearly illustrates the formation of the hot core with a radius of =0.23 mm, in agreement with experimental observations [5]. A similar simulation without a reductionof the heat conductivity showed no formation of a hot core and the confinement time
was lower by approximately a factor of five. The experimentally observed confinement time is in agreement with the simulation that assumes a uniform magnetic field.
Figure 7(b) shows the profiles of plasma pressure P, electron temperature Te
and ion density n; (electron density is 3n;) at 6 ns after irradiation. The figure
clearly shows that the thermal energy of the plasma is confined by the magnetic field
while the plasma pressure is sustained by the inertia of the tamper shell, as expectedfrom the first investigation [1].
With a plasma density of 6 x 1020 cm"3, as obtained in the simulation, andan observed confinement time of 8 ns in spheromak geometry, the experiment demonstrates the production of a plasma of 500 eV temperature, withnr = 5 X 1012 cm~3-s, which is amazing in view of the relative simplicity of theexperiment and the small amount of input energy.
The fourth type of target is made of parylene, surrounded by a gold shell. Thesetargets are used to test the enhanced confinement time that is due to the large gold
inertia. The experimental results show that the confinement time as deduced from thewidth of the 0.5-1.0 keV X-ray radiation has a clear parabolic dependence on theshell diameter, giving « 34 ns at 2a = 3 mm. We believe that the parabolic dependence is a consequence of the diffusive lifetime of the magnetic field and/or of the
lifetime of the electron heat, which is shorter than the gold inertia time in theexperimental temperature range; this phenomenon strongly indicates the existence ofa magnetic field.

415

200

400
600
Radius ((Jm)

800

10*

~10
CO

J3

E10"
= 1O21

10

10 0

200

400

600

800

Radius (pm)FIG. 7. (a) Flow diagram; (b) profiles of plasma pressure P, ion density nt, and electron temperature
Te, at 6 ns, based on a one-dimensional spherical simulation. A uniform magnetic field of 103 T is
assumed in the calculations of the radial heat conductivity.

To verify the excellent energy and particle confinement directly in the cavity,
C and D targets of 100 pm thickness, with 100 /xm gold coating on the outer surface,were irradiated by 1 /*m laser light of GEKKO XII. A D-D neutron yield of
107-109 was observed when a laser energy of 3-7 kJ was injected into a cavity of adiameter of 1-4 mm. The results indicate stable confinement of hot C and D plasmas
in cavity targets, as predicted theoretically and as inferred from the experiments with
LEKKO VIII.

3.

REACTOR TARGET CONSIDERATIONS

In target design, the most important parameters are the input energy E;
required to achieve ignition and the energy multiplication factor Q. Since E; isproportional to a3 and since the plasma radius is inversely proportional to the expansion speed v, the input energy is a strong function of v. However, the use of the416

NAKAI et al.

expansion speed in the evaluation is ambiguous because local heating of the interfaceof the cold fuel causes plasma ablation in the centre and reduces the expansion speed.
In particular, if the ablation rate is comparable to the piston speed, which is quite
feasible because the piston speed is only 10"2 of the sound speed of the hot core, theplasma density of the hot core remains unchanged during the expansion and theignition condition can become independent of the expansion speed.
To study these effects, we have performed 1-D spherical hydrodynamic simulations using the HISHO code for reactor size targets. The following assumptions havebeen made in the simulation: (1) The effect of magnetic field becomes manifest onlyin the reduction of the conductivity by a factor of (1 +a>cT2)~1 at the time when thelaser power drops to 1CT2 of the peak value. (2) The absorption rate of the laser is
50% and the absorbed energy is deposited in the electrons and ions in 5/7 of the fuel.
(3) The alpha particle energy is absorbed by electrons at the point of its generation.
The physical parameters used in the simulations are:— Laser pulse duration (with a Gaussian shape) of 10 ns;— Laser energy range of 13-26 MJ;— Inner radius a of the target of 3-4 mm (Fig. 1), with a D-T fuel thickness (b-a)
of 1250-1500 fun and a metal shield (gold) thickness (c-b) of 3000 fim;
— Magnetic field of 104 T.
Figure 8 shows the temporal variations of the fuel internal energy for different
values of absorbed energy Ea (one-half laser energy) and inner radius a of the target
as obtained in the simulation. When the absorbed energy is not sufficiently large sothat the plasma temperature is subcritical, the energy loss is due to expansion, radiation and heat conduction. However, beyond a critical level of the absorbed energy,alpha heating is stronger than the energy loss and ignition is possible. The condition
for ignition is an increase of the internal energy after the laser is shut off. In Fig. 8,
t = 0 corresponds to the time of the peak energy of the laser, hence at t = 5 ns thelaser is shut off. We note that ignition occurs at Ea = 16.7 MJ for a 3.0 mm radiustarget and at Ea = 20 MJ for a 4.0 mm radius target. The ion temperatures at ignition are approximately 10 keV. In the plot for a = 3.5 mm, results obtained withouta magnetic field (open circles) are also shown, indicating the importance of the magnetic field. Without a magnetic field, ignition does not occur.
Figure 9(a) shows the flow diagram for the case of 26 MJ/3.5 mm presented
in Fig. 8. The formation of the radially propagating shock wave, the rarefaction wave
behind the shock, the piston motion behind the rarefaction wave and various streamline contours of the internal plasma flow are shown. The shock speed us obtained
from this figure is approximately 6 x 107 mm/s; it agrees with the theoretical value
VP/ps = 4 x 106 VnT, where P is the fuel pressure and ps is the gold massdensity, when the observed values of T = 30 keV and n = 4 X 1021 cm"3 (seeFig. 9) are used. However, the piston speed, which is two-third of a,, is lower than
the theoretical value. This is a consequence of the rarefaction wave, which tends to417

IAEA-CN-47/I-I-2

R= 3.0mm
100

4.0mm

3.5mm
T

i

i

I

26.7MJ
- j

o
w

1 1°
<5

O no B-fleld

i
0

i
i I
20
40 0
Time (ns)

I
I I
20
40 0
Time (ns)

I
I
20
40
Time (ns)FIG. 8. Fuel internal energy as a Junction of time after the laser pulse, for various inner radii of thetarget and different values of absorbed energy (one-half laser energy), obtained in the one-dimensional
spherical simulation HISHO. Ignition is shown to be achieved at an input energy of around 1 MJ.

push the piston backwards. The rarefaction wave can be produced by the expansion
due to the bremsstrahlung photon deposition in the gold tamper.
The most remarkable feature in Fig. 9(a) is the significantly slower expansion
speed of the internal hot plasma. For example, if we take the streamline that crosses
at R = 2 mm at 50 ns, where the temperature begins to drop, the expansion speed
can be deduced to be 0.5 us. The reduction of the expansion speed v is a consequence of ablative inward flow of the solid fuel and of the rarefaction wave whichslows down the piston motion. Both the ablative flow and the rarefaction wave are
due to heating by radiated photons, heat conduction and alpha particles produced in
the central hot core region. Consequently, the entire system has a built-in energy
recycling mechanism, which is absent in most other fusion schemes. Figure 9(b)shows the electron temperature (T), the ion number density (n^ and the plasmapressure (P) of Fig. 9(a) at 50 ns.
The slow expansion speed v of the hot core indicates that the central plasma
density remains approximately constant rather than decaying as r"3 in accordancewith the expansion of the tamper.
Figure 10 shows the temporal behaviour of the ion density (ni0), and of theelectron and ion temperatures (Teo, Ti0) at the centre of the core shown in Fig. 9.

NAKAI et al.

418

104

1O 3
(0
3

102

10

10 2 00.0

2000.04000.06000.0R
FIG. 9. (a) Flow diagram for the case of a = 3.5 mm and Ea = 26 MJ. The laser peak is at IS ns.
(b) Spatial profiles of the electron temperature Tt (the ion temperature is approximately 'ATJ, theion density n, (= ne = n), and the plasma pressure P, at 50 ns.

The broken line indicates the temporal variation of the central density when it obeys
r"3 variation. This figure clearly shows that the core plasma density remains almost
constant, in spite of the expansion of the tamper. Naturally, the plasma density isprovided continuously by the ablation of the solid fuel. We also note that the central
temperature increases at ignition.

419

IAEA-CN-47/M-2
1

1

1

1

1

I ^

O,Ny
/

10

1oao

/

- 10"

10 -

;

n.io
«

-

j

•

\

:

/
1
1

1 -

1

10

• i i
100

i •

Time (ns)FIG. 10. Temporal variation of the central electron and ion temperatures, Te0 and Ti0, and the central
ion density, nl0. The dot-broken line shows the density variation ifnr3 = const. Q is the fuel gain andNy is the neutron number density.

4.

CONCLUSIONS

A new fusion reactor scheme which combines the benefits of magnetic fusion
and inertial fusion is introduced. A plasma with a density n of > 1021 cm"3 and amagnetic field of > 102 T are generated by means of hot electrons and by ablationfrom the solid fuel contained in a spherical metallic shell of a radius of a few
millimetres. The plasma pressure is confined by the inertia of the expanding heavy
metallic shell and the plasma heat is insulated by the magnetic field to form a hightemperature low density core. Spherical simulations, based on a hydrodynamic code
(HISHO) in which the radial heat conductivity is reduced by a factor of (<oce7)~2,
showed the occurrence of ignition at an absorbed energy of 15 MJ. The expansion
speed of the hot core is found to be significantly lower ( = 0.1) than the expansion
speed of the shock wave. This is a consequence of the ablative flow from the solid
fuel towards the centre, produced by photon heating due to bremsstrahlung radiation.
Preliminary experiments, using 100 J CO2 lasers, achieved a confinement timeof up to 20 ns with a temperature of =500 eV in a cavity target of parylene. The
estimated electron density of 6 x 1020 cm"3, based on a simulation for a target of
1 mm diameter, gave nr of 5 x 1012 cm"3-s.
The new scheme presented here has potential engineering advantages over the
traditional magnetic fusion schemes because of its structural simplicity, the elimination of the wall loadjng problem and the lower output power. Furthermore, with thepresent scheme there is no problem of a pressure driven instability. Compared withthe traditional inertial fusion approach, this scheme is also advantageous because it
does not require implosion. The scheme can utilize high efficiency CO2 lasers or

420

NAKAI et al,

particle beams. For the design of the present scheme to.be applied in a reactor, morestudies are still required, for instance detailed studies of the generation processes ofthe magnetic field and the physics of the expanding boundary of the hot core which
controls the behaviour of the fuel density.
',

REFERENCES
[1]

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

HASEGAWA, A., Rev. Laser Eng. (Japan) 13 (1985) 585.
HASEGAWA, A., DAIDO, H., FUJITA, M., MIMA, K., MURAKAMI, M., NAKAI, S.,NISHIHARA, K., TERAI, K., YAMANAKA, C , Phys. Rev. Lett. 56 (1986) 139.
STAMPER, J.A., McLEAN, E.A., RIPIN, B.H., Phys. Rev. Lett. 40 (1978) 1177.
YATES, M.A., VAN HULSTEYN, D.B., RUTKOWSKI, H., KYRALA, G.,
BRACKBILL, J.U., Phys. Rev. Lett. 49 (1982) 1702.
YAMANAKA, C , NAKAI, S., MATOBA, M., FUJITA, H., KAWAMURA, Y., DAIDO, H.,
INOUE, M., FUKUMARU, F., TERAI, K., IEEE J. Quantum Electron QE-17 (1981) 1678.
DAIDO, H., FUJITA, M., TERAI, K., MIKI, F., NISHIHARA, K., MURAKAMI, M.,
MIMA, K., NAKAI, S., YAMANAKA, C , HASEGAWA, A., User Part. Beams 4 (1986) 17.
TAN, T.H., McCALL, G.H., WILLIAMS, A.H., Phys. Fluids 27 (1984) 296.
TERAI, K., DAIDO, H., FUJITA, M., MIKI, F., NAKAI, S., YAMANAKA, C , Jpn. J. Appl.
Phys. 23 (1984) L445.
ROSENBLUTH, M., BUSSAC, M.N., Nucl. Fusion 19 (1979) 489.
DAIDO, H., MIKI, F., MIMA, K., FUJITA, M., SAWAI, K., FUJITA, H., KITAGAWA, Y.,
NAKAI, S., YAMANAKA, C , Phys. Rev. Lett. 56 (1986) 846.
DAIDO, H., TATEYAMA, R., OGURA, K., MIMA, K., NAKAI, S., YAMANAKA, C ,Jpn. J. Appl. Phys. 22 (1983) L248.
NISHIHARA, K., Quart. Prog. Rep. ILE-QPR-83-5, Institute of Laser Engineering, Osaka
University (1983) 20.

IAEA-CN-47/I-I-3

PLASMA STABILIZATIONIN TRAPS WITH SPATIAL AXIS
V.M. GLAGOLEV, A.A. SUBBOTIN,
B.A. TRUBNIKOV
I.V. Kurchatov Institute of Atomic Energy,
Moscow, Union of Soviet Socialist Republics
Abstract
PLASMA STABILIZATION IN TRAPS WITH SPATIAL AXIS.
Possible methods of plasma stabilization within traps with spatial axis for both circularand triangular-elliptic magnetic surfaces, applicable to Dracon, Spitzer's Eight and Heliac, arestudied theoretically.

1.

INTRODUCTION

The purpose of this paper is a theoretical study of plasma stabilization
techniques in closed traps with non-flat magnetic axes. Spitzer's Eight, Dracon,
helical tori and Heliac, to be considered below, are examples of such traps. All
these traps have complex geometries which are difficult to construct.
However, the possibility of stationary plasma confinement without longitudinalcurrent is their main attractive feature. The Dracon trap proposed in [ 1 ](Fig. 1), which has long rectilinear sections, is the most promising device. For
stability analysis, we shall use the well-known Mercier criterion written as an
expansion in terms of the small parameter 0 = 87rp/B2 (see [2]):M = S 2 +|3W 1 +/J2wY + 03W3,+ /34W4 + ...

(1)

where S = p9 lna/3p is the shear, W^ the vacuum magnetic well (or hill), andW2.W3.W4... are the consecutive expansion coefficients.

2.

STABILIZATION AT CIRCULAR MAGNETIC SURFACES(V.M. Glagolev, B.A. Trubnikov)

Circular cross-sections are suitable because their design is simple and themagnetic field is used effectively. They do not need any auxiliary windings;
the following methods of plasma stabilization can, in principle, be used in theirpresence:(1) finiteness of ion Larmor radius [3];(2) additional rings of hot electrons or ions;
421

422

GLAGOLEV et al.

FIG. 1. Two projections of Dracon with stabilizing S-like plugs (3).

(3) sloshing ions;(4) feedback stabilization;(5) external travelling fields smoothing the surface [4];(6) rotation in the radial electric field;(7) a set of conductors with current near the plasma surface [5];(8) stabilization by shear, i.e. by field line crossing;(9) non-uniform field at axis of curved pieces ('S-plugs');(10) self-stabilization effect due to finite pressure [6].
Methods 1 to 7 were, in part, considered previously and applied in practice.
For example, a finite ion Larmor radius, xv = vTi/c0j, stabilizes the flute-like
perturbations with sufficiently small wavelengths [3]:(2)where a is the plasma column radius, and K the curvature of its axis. This method
is expected to be particularly effective under ECRH, when, for T e > T,, as isshown in Ref. [7], complete stabilization of flute-like perturbations can bereached if the condition

IAEA-CN-47/I-I-3

rfff = V2T e /mj > 0.4 a VaK ~ - a

423

(3)

is satisfied and displacement of the plasma column as a whole is excluded.
Methods 8 to 10 will be discussed below in greater detail for Dracon andSpitzer's Eight.
2.1. Stabilization by shear, S = p3 In a/dp
If the magnetic axis of a trap has curvature K(s) and torsion K(S), the angleof rotational transform for magnetic field lines at the axis itself will be a 0 =outside the axis, we have [8]:a L ( p ) = a 0 ( l - — P2<K2>J;<K2> = Cp K 2 «ds/ j> «ds(4)Hence, the shear is found to be

,

<

-

-

-

(

•

5

)

16
It differs from zero even for circular magnetic surfaces.
The Mercier criterion (1) for low j3 provides the stability condition
»
9
~2LB°
n
S +/3W!>0; Wt = L(B 0 /a 0 ) V (0)p — l n p = — — - 0V"
dp?ra oc0
,

2

2

(6)

Here, L is the total length of the trap, 0 = np2B0 is the longitudinal magnetic flux,
and, for constant curvature of the bent pieces, K o , and uniform field Bo attheir axes, the quantity V"(0) provides a 'hill' equal to= J> (K/B) ds/7r = - L (K /B )
2

K

J

0

2

0

(7)

IT

where LK is the total length of the bent pieces. In these conditions, theMercier criterion gives(8)

which is too small at the plasma boundary. However, the hill can be reduced andeven transformed into a well if one makes the magnetic field at the axis nonuniform as will be shown below.

424

GLAGOLEV et al.

2.2. Stabilization by a non-uniform field (by 'S-plugs').
Such a method is considered in Refs [9, 10].* Therefore, we shall, below,
present a useful technique for its optimization in Dracon, where one can use adimensionless variable t = /Kds instead of length s, considering it as a certain
fictitious current 'time'. Let us write the field at the axis of the bent pieces
as B = B o b(t) and introduce a flat vector:t=/cosa; -sina}t~ 3/2 dt

(9)

-T/2

and its 'velocity' v = q with its magnitude |v| = b 3 / 2 . Then, the expression for a'well' can be written as [10]:M(M •'«•*£'•
The well V"(0) < 0 can be obtained from the last term when a number N of
additional, S-shaped, magnetic plugs is introduced at the centre of each bent
piece (CREL, i.e. a curvilinear element), where |q| is a maximum one. The
curvature in the plugs is chosen, for the sake of convenience, as K = VbK e n d ,providing constant aspect ratio along the plug. Then, from (10), we find(11)
-T s /2

where Ko is the curvature, L$ the length of CREL without taking account of theplugs. The function L turns out to be equal to E = v^f- (3 q2 + q2 + qq)/3 q in
this case; it plays the role of a Lagrangian. Hence, solving the Euler-Lagrange
equations,9L

d2 3 t

d 3L

Id

+

where X = (3/2) In b(t), we find the optimum field profile:
X(t)dt: = t + - Ts

VX-(e x -l)A

2

(12)

IAEA-CN-47/I-I-3

425

b(s)
,-3/2^IK

Vlnb-|A(b3/2-l)

c;

/

N3\

,

-LsKend2'(13)

where A = f (In Il)/(ri 3 / 2 - 1), and n = B m a x /B m i n is the mirror ratio in the plug.
The integrals (13) can be evaluated only numerically, but the expansione^ — 1 = A(l + X/2) is a good approximation, giving the following profile:which is used in [ 11 ], where a well V"($) < 0 is obtained for Dracon withcircular magnetic surfaces, under the condition:
K0/NKend<F(n)«(l/TCREL)[lqlmax-(4/n)]lnn(15)

The function F(I1) is weakly dependent on the CREL configuration, e.g. for
11 = 2; 3; 4F(2) = 0.10-0.15; F(3) = 0.25; F(4) = 0.35. Thus, for a mirror
ratio n = 3, it is sufficient to have two plugs only and K 0 /K end = 2 for obtaininga zero well.
Note that the well in Dracon can be produced by a traditional triangularelliptic deformation technique for the magnetic surfaces of a uniform field as
is shown in Ref. [11].
2.3. Stabilization of Spitzer's Eight
A non-uniform field with S-plugs can be used for plasma stabilization, notonly in Dracon but also in Spitzer's Eight, by arranging the plugs as is shown in
Fig. 2. Such a trap includes two tilted semitori with curvature Ko, four toroidal
pieces with an angle of turn \j/l adjacent to them and placed in two parallel planes
and, finally, two stabilizers with flat axis and S-like plugs. In contrast to the
Dracon, there is no equilibrium in the separate pieces, but equilibrium is provided
as a whole, as in any spatial trap. The curvature of each plug is chosen to be
K = VbK e n d ; the angle of turn, read off from the plug centre, is designated by <p(Fig. 2). These parameters should be matched geometrically, and then stability' isanalysed by an expression for the well, V"(0), from [8].
The calculated dependences of <p, i/>, K 0 /K end = g on the mirror ratio, n, inthe plug are given in Fig. 3. For II = 3, we obtain a magnetic well of about *-1% •at an equilibrium 0 = 2%, if there is only one plug in each stabilizer. The enhanced
transport coefficients, due to the drift of the particles captured in the S-plugs,
may turn out to be a disadvantage of such a trap, as well as in the Dracon •426

GLAGOLEV et al.

FIG. 2. Two projections of figure-of-eight-like trap with S-like stabilizers.

1 • 0.2
0.5

4 n
FIG. 3. Dependence oftp, \p,g on mirror ratio FI for stabilized figure-of-eight-like trap
as shown in Fig. 2.

discussed previously. A reduction of this effect is possible by reducing the mirror
ratio together with an appropriate rise in the number of S-plugs, which allows
higher values of j3 to be achieved.
The higher order terms in the Mercier expansion (1), which were consideredpreviously in Refs [2, 12], are taken into account in the next section. The terms
j32 and 0 3 have been shown to be eliminated for circular surfaces. They are,however, different from zero for triangular-elliptic surfaces and can assist in
stabilization as will be shown below for the Heliac trap.

IAEA-CN-47/I-I-3
3.

427

ANALYTICAL OPTIMIZATION OF A THREE-DIMENSIONALHELIAC-TYPE CONFIGURATION (A.A. Subbotin)(1) As is shown by numerical calculations [13, 14], in systems withhelical symmetry, the jS-limit can reach high values (~30%). When, however, aconfiguration is closed into a torus, plasma stability, for a small number of
periods in the system, may deteriorate considerably [15]. Numerical optimizationof 'flexible Heliac' systems has shown that )3 =* 8% can be reached [16]. In thegeneral case, numerical optimization of toroidal systems with helical axes, of theHeliac type, is difficult because of the great number of parameters on which the
MHD plasma stability depends (pitch, ellipticity, triangularity, etc.). One musthave analytical models to specify the directions in which the optimum should
be sought.
In this paper, an analytical model calculation of plasma stability of Heliac
type systems is carried out by the method of expansion into a power series interms of the distance from the magnetic axis. The estimates of j3 are done by
the Mercier criterion and by the sufficient stability condition. The method usedin this paper is similar to that developed in [ 12, 17].
(2) The plasma equilibrium equations in the flux co-ordinates (a, 8, s) withstraightened field lines have the form [12]:

812

9y?

§22

,

,

Sis

§23

J + ^ =T X ' + T *
dtp

§23

9? Vi

,

,

(19)

§33

\/g

where primes denote derivatives with respect to a, <... )gg denotes the averageover 6 and f, <j> and x are the toroidal and poloidal magnetic fluxes, J and I are thecurrents, v and v? are periodic functions of the co-ordinates, 6 and f, whichcharacterize the 'charge separation' current and the scalar potential of theirrotational part of the magnetic field; f = s/R, R = L/27T, and L is the system
length. The magnetic axis is characterized by curvature K(s) and by torsion K(S).

428

GLAGOLEV et al.

Let us consider the x, y co-ordinate system orthogonal in the s = const planeand rotating around the tangent to the magnetic axis with an angular velocity
of 5'(s).
Let x and y be given as series in powers of a:cj + a 2 £ 1 (f)

(21)
(22)

where co = co0 + af(0, f), co0 = 0 + h(f); h(f) and f(6, f) are periodic functions,and X and 7a are elongation and triangularity of the magnetic surface cross-section,
respectively. The calculations are done on the assumption that Xj = 75- = 0. The
field at the axis is assumed to be uniform, Bj>(f) = 0. Configurations with72a2) < 1 and a ^ / R 2 < 1, where a0 is the co-ordinate of the plasma boundaryand n the number of axial periods, are considered.
One can derive the expressions for the metrical coefficients g^ in the zero,
first and second approximations in terms of small parameters. Averaging Eq. (9)in the zero approximation with respect to 6 and f, we obtain for the number ofrotations:
V = x'/4>' = tto+t*j

(23)

where
2X JR

2X

Let us write <5l> = £, K0R = <«R>f + N, where £ and N are integers. In Heliac typeconfigurations, we have £ + N = n. Then, for 5 and ju0 we can write the following
relationships:

t
(24)

2X

X2+l

-

-

(25,

Then we assume that there is no longitudinal current, i.e. MJ = 0, and the pressure
is a parabolic function of a. The parameter (3= 2p/Bo =-2np'aal/<t>'2.

429

IAEA-CN-47/I-I-3(3) By using methods similar to those employed in [12, 17], the Mercier
criterion and the sufficient stability criterion can be written as(26)

5U/U + W - B M ; S > 0

Here,5U/U is the vacuum magnetic well, W the deepening of the magnetic welldependent on the magnetic field line displacement due to the plasma pressure,
B M ; S is the destabilizing 'ballooning' term, which depends on the pressure via asquare law. The subscript M corresponds to the Mercier criterion and the:subscript S to the sufficient stability condition.
For the quantities entering (26) the following relationships have been
derived:

(27)

(28)
87r2p'R2a3 X2 + 1
2X<K 2 >/7+X 2
4X \3X2 + 1

6U/U = -a<K2cos26> /7X2 + 1
3+X2
4X

X2-l

(29)

7X2
X2
3 + X22_

3 (X2 - 1)
3X2 + V

3X2 + 1

7 <K cos 5>

X2 3

2X

X |-^—1 (2X2 + X + 2) (Ke'' i6i6)[)[-

i6)[(2X2 - X + 2) (Ke i 6 )[(30)

The function ov is the solution to the equation
(31)

430

GLAGOLEV et al.

(4) The calculations have been done for a spatial axis given in the form of
a 'spirel' (spiral e/ement) [18], i.e.

exp

(32)

•I «(?)dr R

This representation holds for a wide range of closed curves, including curves close
to the magnetic axis of Heliac. The parameter K0 characterizes the amplitude of
the helical magnetic axis. The magnetic axis is an n-periodic helical line upon thetorus with minor radius_ R / 2n2 *oR~ n V n2 - 1 n
for low K0. The number of periods should tend to infinity for the transition
to a helical symmetry, considering R/n to be fixed. If we prescribe the function
h(f) as h = hi sin nf, we shall be able to obtain an analytical solution to Eq. (31)and to perform the necessary averaging procedures. The stabilizing terms turnout to be maximum for 6(0) = 0. Writing down the consecutive relationships,
we may consider this condition to be satisfied. Solving Eq. (31), one gets a
relationship

(33)
M=-°°
where
X+ 1
K

2X1/2
X- 1

' 2X 1 / 2° J M-l( h l(X+l) 2
2X

(X-l)
2X

2X

~ Mn)

Here, J M is a Bessel's function. Substituting this expression into (27) to (30), we
obtain the relationships for the j3-limit:
aon

aon

(b M ; S - v)

(34)

IAEA-CN-47/I-I-3

431

Here,

C.=

3(X2-1)

R

2

2X

3X +1

1 / 7 + X2n

7X2 + 1(
4X\3X 2 +1

K2Jo(^r^h1l-2K0K1J1

3+X 2

h,

•K?J 2

X2+l

2X

2X

2X Y
V 2X

(X2-1)Xn (3X 2 + 1)(3 +X 2 )M

16X(X2+v = (3X 2 +1)(3+X 2 ) 2 S . 2 ( X 2 + l )

R2

R2

2

X2+l
X

max

S

m =0

X2-l
2
m TTTTZm)cosmnf R
+

X2+l

432

GLAGOLEV et al.

0.1

FIG. 4. Limiting $ obtained from optimization with respect to aon/R, X and hlt shown iscasewitk: yao = O.3, l:n = 2, 2:n = 4, 3:n = 6, 4: n ->«>. Solid lines correspond to Mercier
criterion, dashed lines to sufficient stability condition.

X+l

J

4X

-i

(X+l)2-hi
2X(X+l)22X

M+1(X+l)2

X2+l
•h,
4X

(X + I) 22X

1J

M+i( h l

2X-

K0JM ( h tn2(X+ I)

l

2X

- n2

433

IAEA-CN-47/I-I-3

2

0.2

10

1

20

FIG. 5. Limiting /} (Mercier criterion) as a function of number of periods obtained from
optimization with respect to X, ht and K0R/nfora0/R = 0.05; l:yao = 0.3; 2:ya0 = 0.4.

c

0 =

Md-M
M

M

M

•-I

M

d», =
M'

Thus, the )3-limits are the functions of triangularity, ellipticity, aspect ratio,
number of periods for the magnetic axis, and of tc0R and h^ As is seen from (34),the /3-limits depend monotonically on triangularity. Optimization can be carriedout for the other parameters.
If triangularity and number of periods are given, the optimization for theother parameters results in the dependences shown in Fig. 4. The case with
7a 0 = 0.3 is illustrated. With an increase in triangularity, the graphs 0 M . S (/c0R/n)
are stretched along the axis of ordinates, proportional to y2a.l. In most cases,
the optimum cross-section elongations lie in the range from 1.6 to 3, themaximum angle amplitude is 6 < 0.3. Figure 4 shows the optimization of thesystem for a given number of periods and for given triangularity. In this case,the /3-limits rise with the number of periods and, for n -»• °°, can be very high,because of the increase in the period length. If the aspect ratio is finite andprescribed, the number of periods can be optimized for the system. As can beseen from Fig. 5, the optimum n o p t «* 6 to 8, for a o /R = 0.05.
(5) Thus, rather high /3-limits (according to the Mercier criterion) (> 10%)can be reached even for a low number of periods and for not too large triangularity,
in toroidal configurations with helical axis. The /3-limit found from the sufficientstability condition turns out to be lower by about an order of magnitude.

434

GLAGOLEV et al.

Although the results obtained in this paper are, strictly speaking, only
correct for small torus curvature, they may provide some limits as to the
optimization of real systems.

ACKNOWLEDGEMENT

The authors thank V.D. Shafranov for stating the problem and for useful
discussions.

REFERENCES[ 1 ] GLAGOLEV, V.M., KADOMTSEV, B.B., SHAFRANOV, V.D., TRUBNIKOV, B.A.,
in Controlled Fusion and Plasma Physics (Proc. 10th Europ. Conf. Moscow, 1981),
Vol. 1, Moscow (1981) Paper E-8.
[2] VOLKOV, T.F., DOBRYAKOV, A.V., TRUBNIKOV, B.A., Nucl. Fusion 25 (1985) 891.
[3] KADOMTSEV, B.B., in Voprosy teorii plazmy, Vol.4, Moscow (1964) 281; see also
ROSENBLUTH, M.N., ROSTOKER, N., KRALL, N.A., Nucl. Fusion Suppl., Part I(1962) 143.
[4] TRUBNIKOV, B.A., in Fizika plazmy i problema UTS, Vol. 4, Moscow (1958) 309.
[5] FIRSOV, O.B., ibid., Vol. 3, p. 327.
[6] MIKHAILOVSKU, A.B., SHAFRANOV, V.D., J. Ehksp. Teor. Fiz. 66 (1974) 190.
[7] MIKHAILOVSKIJ, A.B., Nucl. Fusion 4 (1964) 108.
[8] SOLOVEV, L.S., SHAFRANOV, V.D., Voprosy teorii plazmy, Vol. 5, Moscow (1967) 3.
[9 ] TRUBNIKOV, B.A., Vvedenie v teoriyu plazmy, Vol. 1, MIFI (1969).
[101 TRUBNIKOV, B.A., GLAGOLEV, V.M., Fiz. Plazmy 10 (1984) 288.
[11] TRUBNIKOV, B.A., GLAGOLEV, V.M., LAZAREV, S.L., DOBRYAKOV, A.B., Fiz.
Plazmy 11 (1985) 155; GLAGOLEV, V.M.,TRUBNIKOV, B.A., CHURIN, Yu.N.,
Nucl. Fusion 25(1985)881.
[12] SHAFRANOV, V.D., YURCHENKO, E.I., Nucl. Fusion 8 (1968) 329.
[13] DEGTYAREV, L.M., POSHEKHONOV, Y.Y., SHAFRANOV, V.D., in Controlled
Fusion and Plasma Physics (Proc. 1 lth Europ. Conf. Aachen, 1983), Vol. 2, European
Physical Society (1983) 103.
[14] MONTICELLO, D.A., et al., Phys. Fluids 27 (1984) 1248.
[15] GROBMANN, R., et al., in Controlled Fusion and Plasma Physics (Proc. 11th Europ.
Conf. Aachen, 1983), Vol. 2, European Physical Society (1983) 155.
[16] GUASP, J., et al., in Controlled Fusion and Plasma Physics (Proc. 12th Europ. Conf.
Budapest, 1985), Vol. 1, European Physical Society (1985) 441.
[17] SHAFRANOV, V.D., YURCHENKO, E.I., Nucl. Fusion 9 (1969) 285.
[18] ZAKHAROV, L.E., SHAFRANOV, V.D., in Voprosy teorii plazmy (LEONTOVICH, M.A.,
Ed.), Vol. 11, Moscow, Energoizdat (1982) 118.

IAEA-CN-47/M-4

RELAXATION OF
PLASMA TOROIDAL ROTATIONIN TOKAMAKS AND ANOMALOUS VISCOSITY
D.Kh. MOROZOV, O.P. POGUTSE
I.V. Kurchatov Institute of Atomic Energy,
Moscow, Union of Soviet Socialist Republics

Abstract
RELAXATION OF PLASMA TOROIDAL ROTATION IN TOKAMAKS AND ANOMALOUS
VISCOSITY.
A theory of anomalous viscosity in tokamaks is presented. Previous results are generalized
for the case MpVm <; I, where M and m are the ion and electron masses, respectively, and |3 isthe ratio of the plasma pressure to the magnetic pressure. A complete set of equations foranomalous transport with gradients of density, temperature and toroidal momentum is obtained.
This theory describes the damping of the toroidal rotation after neutral beam injection has beenswitched off. A comparison of theoretical and experimental results is presented and good
agreement is observed. In addition, the contribution of drift waves with wavelengths
comparable to the ion Larmor radius is analysed, and model equations for the toroidal andpoloidal velocities, including parallel neoclassical viscosity and two types of anomalous
viscosity, are obtained.

1.

INTRODUCTION.

A series of plasma experiments with non-perpendicular neutral beam injection
in tokamaks have shown anomalously strong damping of rapid toroidal rotation
after the neutral beam has been switched off [1-5]. This damping is not described bystandard neoclassical theory. There are two points of view regarding this problem:(a) Stacey and Sigmar [6] attempt to explain this effect in terms of inertia andgyro viscosity; (b) Kadomtsev, Morozov and Pogutse [7] propose an anomalous
viscosity mechanism based on small scale turbulence. This mechanism follows
naturally from a set of equations describing self-consistent transport in the strong
magnetic field of a tokamak [8]. The results of Ref. [8] are, however, correct for
Mj3/m < 1 only (see later), where M and m are the ion and electron masses,
respectively, and jS = 87rnT/B2. The theory of anomalous viscosity in Ref. [7] isgeneralized to the case of M0/m > 1, and a complete set of equations for anomaloustransport with crossfield effects (density, temperature and toroidal velocity
gradients) is obtained by averaging the set of equations [8] over small scale
fluctuations.
435

436

MOROZOV and POGUTSE

The purpose of this paper is to expound briefly the main results of anomalous
viscosity theory and to compare these results with various experiments and withgyroviscosity theory. We also take the contribution of drift instabilities toanomalous viscosity into account.

2.

AVERAGED TRANSPORT EQUATIONS AND ANOMALOUS
VISCOSITY
The set of equations proposed in Ref. [8] consists of four equations:(a) Continuity equation:
dn

c

-*
47reB0 ( B' V ) A ^

dt

(b)

(1)

Generalized vortex equation for V = -pj"2 (1 - exp (p2 A^)) Io (-p 2 A^)drdt

1
47rMn0(2)This is a model equation since it does not take into account the diamagnetic ion
drift.
(c)

Generalized Ohm's law:
d^
c -+
n
C T ?
-T^c&r-VVBVT eeln+— 1 A | *
B-VT
ln
—
+
dt
eB0
n0 47T x
4(3)(d) Heat transport equation:
1 d(B,Vq||)
T

Here, d/dt = 9/9t + (c/B0) [e T , V¥?], B p = [e T , V\p], n is the density, n 0 the nonperturbed density, ps the ion Larmor radius, f) = a"1 the collisionless resistivity
operator, andm2 f ,/ m v? a\=" T J v vHf(vll)dvll = ^ ~ ^ e ~ + 7 "•

IAEA-CN-477I-I-4

437

In a linear approximation, the operators fj and £ can be obtained from the electron
drift kinetic equation for a non-uniform plasma. To the set of Eqs (1) to (4), weshould add the following equation:j,, = ^ A ^(5)

Averaging this system over the small scale fluctuations, we can also describe therelaxation of toroidal momentum, since if we neglect diamagnetic drift, thetoroidal velocity will be proportional to the large scale electrostatic potential
gradient [10]: v T = - (c/Bp)/9^>/8x, which is determined by the averaged
equation (2).
We introduce the surface $ = const moving with the plasma when f? = 0
and calculate the flux of N (N means n, nT or i^) across this surface to determine
the transport coefficients. The equation for 3> iss-°
The surface area is S. Its velocity is v<j> = (c/B 0 )[e T ,Vy]- Thus, we havey /NdV= /"^dV+ /*S(7)Substituting 8N/3t = v^-VN + F N from Eqs (1) to (4) into (7), we obtain (F N is
the right hand side of the corresponding equation):

NdV

Let us analyse the case N = n:Using the expressions [8]dS"Tv£i d S

(8)

438

MOROZOV and POGUTSE

and
c2
=—

L

f
/ dt'lV(f?A 1 i/'),V*]-e T

we obtain:^f [Sx^Ax(10)

All variables in the square brackets are functions of t', according to the equation

We carry out a Fourier transformation of \j/ and an averaging of Eq. (10) over asmall time interval T and over the co-ordinate x. The x-axis is normal to thesurface 3> = const. In addition,

J n ='5 (kz"X exp [i.(kx + k x )x] Tj kw (x)dxdkdk'dwdco'(11)Here,J2 = co + CJ^, co^, = -(ck y /B 0 ) (9/9x)<v?>- The parameters w^ and TjkjW are
weak functions of x and may be expanded in Taylor's series. Expression (11)includes 5 = Re(7j/f2) = 6 0 + Sjx + 5 2 x 2 /2. Substituting S into (11), we obtain_(47r)2eB0SAxT

/dk

8
9k v

2

(12)

For X = |J2/k|| v e | < 1, the real part of f? is reduced exponentially. Thus, we canuse the next model for f?: Re f) = 0 for X > 1, and for X < 1 we have:

IAEA-CN-47/M-4
27r 3/2 |k||v e |
J

439

(13)

pe

2TT

Here
ck y d l n T e
COj

eB 0

dn
n o eB o dx<*>„=-•ndx 'Using strong turbulence methods, we can estimate the spectral density for plasma
displacement:c 2 k?
k:,a;12= —^l?l2

Hence, it is necessary to express i ^ ^ in terms of ? k c o , using longitudial Ohm's
law and Eq. (5):(14)
The coefficients TJ,O and k|| depend on x very weakly and may be expanded.
To zero order, we have
(15)

'SI
Corrections may be calculated from perturbation theory:(k«a)'i 8 (an)'
Sl+iv a 9kx Sl + iv
Ck|,

i

"

2
X

Sl+i V

2a dk(kja)"
k||

?k

(16)

We have introduced here the notation v = c2k^i7/(47r). The spectral density may
by derived from the following arguments [11]: If the velocity of matter
displacement is of the order of the phase velocity of the wave, the wave will become
unstable. Thus, we may estimate v ~ a>k/k^, %± ~ v/cok ~ 1/k^. To obtain the440

MOROZOV and POGUTSE

spectral density, we represent v2 as a function of v^, substitute it into | | a n d
average it over the volume SAx and over the time interval T:

•* ->

\

co')t]dkdk'dcodco')~|£ k |

kj

(17)

where k* = oj pe /c is the typical value of k^ for small scale fluctuations. The
typical value of k|| is k|| = (qR)" 1 . We must also replace CJ by S2 everywhere in(17) because this expression is correct for a system moving with the nonperturbed plasma. Finally,

SAxrc

-j

=——

(18)
9n
91nT e \r - + C11n—
)
ox
3x /

ay |

e

(19)

Here, 7 is a number between 1 and 2. At present, 7 is assumed to be 1. Xo isthe minimum possible value of the parameter A = |£2/k||ve|, i.e. the ratio of thelongitudinal phase velocity to the mean thermal electron velocity. In a linear
approximation, the phase velocity is the velocity of the Alfven waves. It may
be assumed that the spectrum hardly spreads into a region with lower phasevelocities. We take it that T e = T\ = T. In this case, we obtain:
Xo = c A /v e = \/m/(M/3 e ). This result is correct for Xo < 1. If Xo ^ 1, only anarrow part of the spectrum near X = 1 will contribute to the integral (12), andwe can assume that Xo = 1. Since this calculation is only an estimate, we
introduce some coefficients of the order of 1, Clt C H , C2, into (19). We cancalculate the fluxes J^ and Jn-p in a similar way. Such calculations are performed
in Ref. [9]. We see from (1) and (2) that thegeneral vortex flux may be obtained fromthe particle flux by multiplying it by o>Bi/n.
Complete averaged equations are obtained by adding the divergences of thecorresponding fluxes J N , arising from small scale averaging, to the right hand sidesof Eqs (1) to (4). Finally, we havednc(B,V)Anfr

dt

4?reB0dt

47rMn0

IAEA-CN-47/I-I-4

441

Aj(Vn + C u nVlnT e )

(20)

n

n + C n nVlnT e ) + A2
(21)

c

At

n

(22)

n0

~ n T e = E||j|| + v|A4(Vn + C41 nVlnTe)(23)

We have performed all the calculations in Ref. [9], in detail. The coefficients A ;are the transport coefficients:
c2

c4

ve

— #' y ;
qR

- C3
c4

A4 - C4n e ve

qR

nemv

4 w4

o pe T e qR

In particular, A2 determines the effective viscosity coefficient, and A4 theelectron energy confinement time in a non-rotating plasma, T^. Comparing A2
and A 4 , we can express the toroidal momentum confinement time r T as afunction of r E . For Xo = (kfp^y1 < 1, we have

m
c2

ve

(24)

(25)

This result can be obtained from qualitative arguments only. Indeed, for s>eff adimensional analysis yields vei{ s= (Ax)2/At. For k*p ; > I, a single collision withthe wave can displace an ion over a distance pv A typical wave-particle collision

442

MOROZOV and POGUTSE

50
tms)

FIG. 1. Comparison between theories and experiment. Horizontal axis corresponds totoroidal rotation relaxation calculated from anomalous viscosity the'ory, vertical axis
designates experimental time. Solid line represents anomalous viscosity theory results,
circles give experimental results, and squares show results of gyroviscosity theory.

frequency corresponds to the frequency of appearance of the potential hill at thispoint, i.e. the wave frequency. The greatest contribution to the anomaloustransport is given by the waves with co ~ v e /qR. In this way, we obtain
a2
v an
2
T T ~ a "eff ~ qR/P? e d> after some elementary transformation, we arriveat (24).

3.

COMPARISON OF THEORY WITH EXPERIMENTS

From the time Ref. [9] was sent to press, some new results on toroidal
momentum damping in tokamaks have been published. There are experimental
results on DITE [3], TFTR [4], ISX-B [5]; also, an alternative gyroviscosity
theory has appeared [6]. In Fig. 1, the horizontal axis corresponds to thetoroidal momentum confinement time calculated from expressions (24) and (25).
The vertical axis shows the experimental confinement time (circles) and thegyroviscous confinement time (squares) calculated from the expression

TT

2R 2 Ze•G(fl)
cT(26)

and we assume G (0) = 1. The quantity r T ( 0 ) is assumed to be an
experimental time. The first point r T = 19 ms relates to ISX-B [5]. The
following parameters are chosen for the calculation: deuterium plasma,<ne> = 2.25 X 1013 cm" 3 = n e (0)/2, <Te> = 400 eV, B0 = 1.4T, R = 93 cm,

IAEA-CN-47/I-I-4

443

a = 27 cm, q = 2, Zeff = 2.5. Very good agreement with the anomalous viscosity
theory is observed. Agreement with the gyroviscosity theory is not so good, but,
considering the degree of uncertainty in both the theoretical and experimental
data, the agreement should be regarded as good in both cases. In all original
papers where the mean temperature or density is not shown, we assumed them
to be equal to half the corresponding value at the1 centre of the plasma column.
The next point corresponds to PLT [ 1 ]. The following parameters arechosen: hydrogen plasma, <ne) = 3.2 X 1013 cm" 3 , <Te) = T e (0)/2 = 1 keV,
Bo = 2.5 T, R = 132 cm, a = 40 cm, q = 2. According to [12], Z eff «* 1 for<ne> > 2.5 X 1013 cm" 3 in PLT. There is good agreement with the anomalous
viscosity theory and a slightly worse one with the gyroviscosity theory. For low
density experiments, (ne> = 1.6 X 10I3crrf3, the damping time is very short,
Tj = 9 ms. (This point is not shown in Fig. 1.) However, for such low densities,
the density of impurities in the PLT plasma is very high, and another mechanism
may be responsible for toroidal momentum relaxation. We cannot compare thisexperiment with the gyroviscosity theory, because Zeff is not presented in theoriginal paper, and, according to [12], Zeff can vary from 1 to 10 for different
regimes.
DITE[3]: deuterium, <ne> = 3 X 1013 cm" 3 , <Te> = Te(O)/2 = 35OeV, Bo = 2.7T,
R = 117 cm, a = 27 cm, q = 2, Zeff = 1. In this case, both theories demonstrate
good agreement with the experiment.
Again PDX [2]: <ne> = 2.9 X 1013 cm" 3 , T e = (l/2)T e (0) = 0.6 keV,
Bo = 2.17 T, R = 143 cm, a = 42 cm, q = 2, Z eff = 1. The time from anomalous
viscosity theory is twice as short, and the time from gyroviscous theory is six
times as short.
TFTR: On the basis of published data only, a rough upper limit for theexperimental time can be estimated. This is the toroidal rotation decay time
which is equal to 0.3 ns [4]. To obtain the net viscous time, we must subtract
the fast ion relaxation time from this value. However, we do not have information
enough and provide only theoretical data for TFTR. For the calculation, we takethe parameters of a regime with a high central temperature of about 10 keV;<ne> = 3.2 X 1013 cm"3, <Te> = (l/2)T e (0) = 5 keV, B o = 4 T, R = 250 cm,
a = 82 cm, q = 2, Z eff = 1.4, r T = 40 ms for a hydrogen plasma, and T T = 20 ms
for a deuterium plasma. The gyroviscosity theory gives r T = 15 ms. A comparisonof these data with experiments should clarify the picture. Bearing in mind thatboth theories are only estimates, we may state that there is good agreement of
the experiments with both theories. Thus, at present, we cannot choose between
the two theories.
4.

DRIFT EFFECTS

We have examined anomalous viscosity associated with electron motion
along the field lines, stochasticized by small scale oscillations with kj/ij^ 1.

444

MOROZOV and POGUTSE

However, ion drift waves with k^p; ~ 1 exist in tokamaks [14] and should resultin momentum transport directly along the ion channel. Let us estimate thecorresponding transport coefficient:

(27)

ot

3 v ||— - + <v||)V||V|1 + ( v r V ) v | | = 0
ot

(28)

Here, we assume V-vj^ = 0. In a quasi-linear approximation, the momentum flux
along V<v||> takes the form:||= Mn<V||v x >=D

"

x

|x|
=Re(

dx

\

)

\Zi(w-kyvn)/

3<v||>
L

)

(29)

3x

We introduce the plasma displacement in the co-ordinate system moving with theunperturbed plasma |£ k | 2 = I vx !/(*;£ and assume that cok = cok + iw k . Thus, we
have

(30)

Here, cok is the inverse wave decorrelation time which determines the spectralline width. Usually, in experiments with drift waves, we have the frequency
k v cTdn0

eB o n 0 9x
and the displacement | § k | — kj 1 — pj. So, the estimate of D ~ (cT/eB0) P;/a andthat of the relaxation time r d ~ a2/D ~ a 3 eB 0 /(cTp i ). The ratio of this time tothe time obtained previously is large: 7"d/rT = (2<#2/q) VM/m for normal
tokamak parameters. Now, we can write down the model equations describing
the relaxation of the parallel and perpendicular velocities, including neoclassicallongitudinal and some anomalous viscosity mechanisms. It is well known [10]that the plasma velocity may be represented in the form v = wej + ub, in a first
approximation in p/a. The equation for vj is
9v ||
at

u
-rneo

IAEA-CN-47/H-4

445

Here, Tneo is the neoclassical poloidal velocity relaxation time. Expressing u interms of vy and v^, we obtain:
9v||

at

=

1 /

vii

r neo V

Vj.B0\

v||

BP/

rd

\
i \
V!(-+—)

(31)

(32)

Usually,
1

1
1
>— +—
T
r
neo
T
d

r

From (31), we may write Vj^ = v||B p /B 0 . Substituting this expression into (32),
we obtain the equation for the parallel velocity:
9v

ll
—U-

/I
1 \
(-+-)

(33)

which has been used for the estimates.
REFERENCES[1] SUCKEWER, S., EUBANK, H.P., GOLDSTON, R.J., McENERNEY, N.R.,
SAUTHOFF, N.R., Nucl. Fusion 21 (1981) 1301.
[2] BRAU, K., BITTER, M., GOLDSTON, R.J., MANOS, D., McGUIRE, K.,
SUCKEWER, S., Nucl. Fusion 23 (1983) 1643.
[3] HAWKES, N.S., PEACOCK, N.J., Nucl. Fusion 25 (1985) 971.
[4] BITTER, M., HILL, K.W., ZARNSTOFF, M., S. VON GOELER, at al., Sattelite Spectra
for Helium-like Titanium, Part II, Princeton Rep. PPPL-2239, Princeton, New Jersey(1985).
[5] ISLER, R.S., WOOTTON, A.J., MURRAY, L.E., LANGLEY, R.A., et al., Nucl. Fusion
26(1986)391.
[6] STACEY, W.M., Jr., SIGMAR, D.J., Phys. Fluids 28 (1985) 2800.
[7] KADOMTSEV, B.B., MOROZOV, D.Kh., POGUTSE, O.P., Pis'ma Zh. Ehksp. Teor. Fiz.
40 (1984) 80 (English translation: Sov. Phys. - JETP Lett. 40 (1984) 818).
[8] KADOMTSEV, B.B., POGUTSE, O.P., Pis'ma Zh. Ehksp. Teor. Fiz. 39 (1984) 225(English translation: Sov. Phys. - JETP Lett. 39 (1984) 269).
[9] MOROZOV, D.Kh., POGUTSE, O.P., Fiz. Plazmy 12 (1986) 666.
[10] HIRSHMAN, S.P., Nucl. Fusion 18(1978)917.
[11] VEDENOV, A.A., Theory of Turbulent Plasma, Results of Science, Moscow, USSR,
VIN1TI Press (1965) 7 (in Russian).
[12] HOSEA, J., GOLDSTON, R.J., COLESTOCK, P., Nucl. Fusion 25 (1985) 1155.
[13] GROVE, D.J., MEADE, D.M., Nucl. Fusion 25 (1985) 1167.
[14] LIEWER, P.C., Nucl. Fusion 25 (1985) 543.

IAEA-CN-47/I-I-5

THEORY OF TOKAMAK PROFILES
D. BISKAMP, D. PFIRSCH, F. POHL
Max-Planck-Institut fur Plasmaphysik,
Garching, Federal Republic of Germany

Abstract
THEORY OF TOKAMAK PROFILES.
An MHD transport model giving rise to enhanced convection is presented. It leads tonearly 'natural' current profiles over a wide range of transport parameters (Part 1). - It isshown that experimental temperature and density profiles are consistent with the assumption that they relax towards profiles related by an expression given in the text (Part 2).

NATURAL CURRENT PROFILES AND PLASMA CONVECTION1

1.

It has recently been conjectured that tokamak plasmas tend to set up
certain current density profiles, which are the solutions of the following variational principle [ 1 ]

-0

(1.1)

Here, Vw, X Vf = B^ is the poloidal magnetic field, jf(\I/,R) the toroidal current
density, I the total current, and a a Lagrange multiplier. This variational principle,
although being somewhat related to the more obvious principle of minimizing
the magnetic energy under the constraint of constant total current, the analog
of Taylor's [2] principle with the total helicity replaced by the current, isnevertheless different from that principle. Instead of providing just one solution, as the Taylor-like principle would do, to the surface current distribution,
which is not very relevant to tokamak plasmas, Eq. (1.1) determines the current
density as a function of 4i:
^a

(1.2)

which for given I yields a whole class of solutions depending on a free profile
1

D. BISKAMP.

447

448

BISKAMP et al.

parameter. For the case of large aspect ratio and circular cross-section, these
profiles, called the natural ones, are given byExperimentally, the profile parameter is usually uniquely determined by the
total current and the presence of sawtooth oscillations. Using the observational
result rj = a/c^ where r j is the sawtooth inversion radius, and assuming
q(rj) = 1, we obtain ( q a / q 0 ) - 1 = q a . It follows from the equilibrium equationthat the pressure profile is also determined, up to a constant factor @p , and is
identical with Eq. (1.3). We should like to note that these profiles havepreviously been derived as the most probable state, by using a statistical
approach [3].
Since tokamaks are open, driven systems, the basic question in the present
context is under what conditions actual tokamak profiles assume the shape given
by expression (1.3). Tokamak transport computations that are being performed
at various places, using sophisticated codes, do not seem to yield simple, selfsimilar profiles. On the other hand, dynamical computations of current sheet
formation and magnetic reconnection [4] reveal that the cross-sheet current
distribution very closely assumes the natural profile (satisfying Eq. (1.1)) inplane goemetry. Hence, it appears that convection plays an important role in
the relaxation process.
Since plasma flows across a strong magnetic field are essentially incompressible, the flow cannot be one-dimensional, as is obvious in the reconnection
problem treated in Ref. [4]. In an axisymmetric system, the flow ~v = B X V<I>/B2
satisfies toroidal incompressibility, V-iT/R2 = 0, the simplest example being thePfirsch-Schluter flow. Such a flow leaves the potential part of the toroidal fieldunchanged, while toroidal geometry, in general, leads to a net transport of the
non-potential part of the field as well as pressure and density. An exact treatment of this problem requires a two-dimensional transport computation, which
will be left for the future. Here, we suggest a 1-D mock-up model assuming
cylindrical symmetry, with vx representing the net flow, and B z replaced by
B z - B o in the magnetic convection term to account for the invariance of thepotential field. In addition, we assume, for the time being, that the density ishomogeneous, i.e. we assume sinks and sources such that p = 1, which, becauseof the loose coupling to the remaining MHD equations, does not seem to be avery stringent condition. Hence, we consider the equations:

dBe
d
d .
-^r= -^-vrB8+-T-r)j2
dt
d
o

(1.4)

IAEA-CN-47/I-I-5

449

dBz
Id
,_
1d
.
t,
— - = - - — rvT {BM-1- - — rt]3e
d£
r dr
r or

(1.5)

dp

Id

. . I d

(1.7)

The flow ur resulting from these equations is essentially resistivity driven,
but because of the effective elimination of the toroidal field we now havevT ~T?p'/Bg, which is larger by a factor BQ/B^ ~ q 2 A 2 than the usual classicaldiffusion. Transport models, on these lines, with plasma confinement producedby the poloidal field only, were also discussed previously and are called
pseudo-classical. One consequence is a more general condition for the resistive
equilibrium, 7}}z — urBg = const, where both terms are of the same magnitude,
so that the current profile can no longer be inferred from the Te-profile. In
addition, configurations are found to relax to /3p = 1, i.e. the poloidal currents
are not sustained, which though not unreasonable appears to be somewhat
restrictive.
Let us first point out a special solution to Eqs (1.4) to (1.7). Since weconsider the case of small transport coefficients, p and B satisfy the equilibrium
equation. We assume a Spitzer-type resistivity, 77 = T?0 (p(0)/p(r))*>; a is aphenomenological heating efficiency and P(r) a general auxiliary heat source.
A simple algebra yields that for v = 1.5, 7 = 2, a = 1, P = K = 0, the natural
profile (1.4) is an exact non-trivial solution of the stationary equations.
Prescribing the total current I and p at the boundary completely determines
the remaining parameters q a /qo and j(0).
This result would not, however, be of great interest if it depended strongly
on the special choice of v and 7, since real plasmas have 7 < 2; v as defined above
would effectively depend on the density and the impurity distributions bothassumed uniform here, and K is finite, K~t\. We have, therefore, solved Eqs (1.4)
to (1.7) numerically for different values of transport parameters with boundary
conditions B z = 1, B# = I(t), p = pl, and a free outflow condition v"- 0. In thecases discussed here, we start from a uniform pressure, p(r) = Pi and BQ (r) = 0,
and increase the current I up to a saturation level on a time-scale r 0 . Except
when T 0 is significantly shorter than the initial skin time, no skin current isformed. For the special parameter choice, we precisely recover the natural

450

BISKAMP et al.

FIG. 1.1.

Current and pressure profiles for v — 5J4, 7 = 5/5, a= 1, K — 0.

a = 0.66

FIG.

1.2.

Current

profilesfor a = 2/3 and 3/2, with v = 3/2, y = 2,K = 0.

profiles. Relaxation is rapid, so that the evolution of j(r,t) is just a sequenceof natural profiles.
Varying the transport parameters we find, in fact, that over a wide range
the current profile remains almost natural - though the profile parameterchanges considerably - while the pressure profile may exhibit stronger deviations from the corresponding natural shape.

IAEA-CN-47/I-I-5

FIG. 1.3.

451

Current profiles for K = T?0 and 70r? o , with v - 3/2, y = 2.

(a)

The exponent v and the coefficient 7 have been varied independently,
1 . 6 > I > > 1 , 2 . 5 > 7 > 1 . A typical example is shown in Fig. 1.1 withv — 1.25, 7 = 1-66. The continuous curve is the numerical profile, thedotted curve is the profile (1.3) with j(0) and q a /qo a j(0)/I from thenumerical result. There are special combinations, such as v - 1.25,
7 = 1.33, where j is extremely close to a natural profile, giving theimpression of an exact solution.
(b) The heating efficiency a has been varied between 1.5 (effective additional
heating) and 0.66 (effective radiation damping), see Fig. 1.2.
(c) Heat conduction has been switched on, in Fig. 1.3. For K < i70, the profiles
are still well approximated by the natural ones, while for K = lOr?o the deviation is substantial. The 'degradation' occurs both in j and p.
(d) For additional heating, the P(r) > t])2 profiles become, in general, strongly
non-natural. However, there seems to exist a (constant) K which minimizes
the deviation.
In conclusion, we have presented a transport model, which within a broad
range of transport parameters yields current distributions there are very similar
to the natural profiles obtained from a variational principle [1 ]. The relationshipbetween the latter and the present transport model is not apparent. Whether theobserved agreement is a special feature of this model or only due to enhanced
convection is an open question. The actual presence of strong convection in
tokamak plasmas has become evident by the distinct variations of p, T andeven p on magnetic surfaces observed recently in several experiments.

452

BISKAMP et al.
ACKNOWLEDGEMENTS

The author appreciates valuable discussions with Dr. D. Pfirsch as wellas the assistance of Mrs. M. Walter in carrying through the numerical computations.

REFERENCES
[1]

[2]
[3]
[4]

BISKAMP, D., to appear in Comments Plasma Phys. Controll. Fusion.
(A similar result has been obtained by KADOMTSEV, B.B. and by
TSU, J.Y.andCHU, M.S.)TAYLOR, J.B., Phys. Rev. Lett. 33 (1974) 1139.
MONTGOMERY, D., TURNER, L., VAHALA, G., J. Plasma Phys. 21 (1979) 239.
BISKAMP, D., Phys. Fluids 29 (1986) 1520.

2.

ENTROPY PRINCIPLE FOR TOKAMAK PROFILES2

It is shown below that experimental temperature and density profiles
are consistent with the assumption that they relax towards profiles related by

T(3T)

I

l

/ /
no\\
expla U-—=?-I 1\ \n(x)//(2.1)

V n0
Here, y = 5/3 is the adiabatic constant, and a > —(7 — O n ^ / n o is independent of
x ; in most cases, one has n ^ = 0. Cases of incomplete relaxation are exceptions,e.g. pellet injection in which the temperature profiles are too flat compared withthe corresponding density profiles.
Relation (2.1) follows from the entropy principle proposed here. According
to it, tokamak plasmas should relax towards states described by relations
T = T(n(x)), in which the total entropy S of the plasma does not change when
the plasma performs arbitrary internal motions slow enough so as not to alter
the relation between T and n. Incomplete relaxation occurs if such motions
are too slow to be able to compete with influences such as those of strong local
heating or large radiation losses. Accordingly, Eq. (2.1) is obtained from2

D. PFIRSCH, F. POHL.

453

IAEA-CN-47/I-I-5

6S= — ^
7-1

/ d3x 8n{x)~
J

=0

In

(2.2)

plasma(2.3)

5n(x) = -with the displacement £(>?) of the plasma vanishing at the boundary, but else
being arbitrary; S o is the entropy constant, a = 1 corresponds approximately
to the profiles obtained by Biskamp [1] and Kadomtsev [2] from the foregoingenergy principle on the assumption that the electrical conductivity is given by
Spitzer's law, a = const-T(x) 3 / 2 . To achieve agreement with the experimental
profiles considered, however, a-values ranging from 0 to 3 are required.
Figures 2.1 (a) and (b) present a comparison of experimental temperature
versus density plots from ASDEX and ALCATOR A, with T(n) as obtained
from Eq. (2.1). An appropriate value of a is given for each discharge.
Tokamak equilibria can usually be described in terms of two arbitrary
functions of the poloidal flux \p. For resistive plasmas where Spitzer's formula
holds one can choose the temperature and the density as these functions.
Equation (2.1) reduces this freedom to the free choice of one function, say T(\p),
and of a special value of the parameter a. This is a feature that might be related
to what is called 'profile consistency' [3],

1.0

T/T,

T/T.

Asdex [4]a = 0.23

FIG. 2.1. Comparison of experimental temperature versus density plots from ASDEX (a)and ALCATOR A (bj, with T(n) as obtained from Eq. (2.1).

454

(a)
a = 0.01

10

FIG. 2.2. Examples for cylindrical plasmas with circular cross-sections:(aJa = 0.01;(b)a = l.

V2Q/3-aa

FIG. 2.3. Density (a) and temperature (b) profiles.

Figures 2.2(a) and (b) show examples for cylindrical plasmas with circularcross-sections, jfip = 1, Bt > B p and j t a T 3 / 2 . The dotted curves are the temperature profiles obtained from Coppi's formula [3] with the safety factor ratios
replaced by corresponding exact constants, Q.
Taking Coppi's formula and our relation (2.1) as independent of eachother leads to density and temperature profiles given by Eq. (2.1), and
2

AV

-Q(-

/n\1/5/

=( —)

/no

\

(«{—n-

n
—
n0

(2.4)

3 \a/
\n 0 /
They are completely determined by the two parameters a and Q; examples areshown in Figs 2.2 and 2.3.

IAEA-CN-47/I-I-5

455

REFERENCES
[1]
[2]
[3]

BISKAMP, D., Rep. IPP-6/258, Max-Planck-Institut fur Plasmaphysik, Garching (1986).
KADOMTSEV, B., personal communication.
COPPI, B., Comments Plasma Phys. Cbntroll. Fusion S (1980) 261.

I

IAEA-CN-4T]i/l-I-6

PELLET FUELLING OF JET PLASMASDURING OHMIC, ICRF AND NBI HEATING'

A. GONDHALEKAR, M. BURES, D. CAMPBELL,
A. CHEETHAM, S.A. COHEN1, J.G. CORDEY, S. CORTI,
A.W. EDWARDS, J. EHRENBERG2, R.D. GILL,
R. GRANETZ, F. HENDRIKS, W. HOULBERG3,
H. JACKEL2, T.T.C. JONES, M. KEILHACKER,
G. MAGYAR, P. MANTICA4, V. MERTENS2,
P.D. MORGAN, P. NIELSEN, J. O'ROURKE, G. SADLER,
F.C. SCHULLER, M.F. STAMP, A. TANGA,
P.R. THOMAS, M.L. WATKINS, A. WELLER2
JET Joint Undertaking,
Abingdon, Oxfordshire,
United Kingdom
Abstract
PELLET FUELLING OF JET PLASMAS DURING OHMIC, ICRF AND NBI HEATING.
Pellet telling experiments have been performed on JET using a single-shot pneumatic injector
giving 4.6 mm (4.5 x 1021 D atoms) and 3.6 mm (2.2 x 1021 D atoms) diameter cylindrical
deuterium pellets with velocity 0.8 < V(km-s~') £ 1.2. Zeff < 1.5 with nD/n,, — 1 on axis has beenachieved. In configurations with a magnetic separatrix, peaked electron density profiles withn,. (0)/ne ~ 2 are obtained, with n, (0) = 1.2 X 1020 m~ 3 and T e (0) ~ 1 keV. Separately a highvalue of nD (0)rETi = 1.3 x 1020 m" 3 -s-keV at T, (0) = 6.5 keV has been obtained with pellet
fuelling followed by NBI heating.

1.

INTRODUCTION

The main objectives of pellet injection experiments on JET
are (i) to explore if pellet fueling gives access to regimes ofhigher plasma density, increased plasma purity, improved
confinement characteristics, and improved ICRF and NBI heating,(ii) to diagnose plasma behaviour, and to model the measured
response in order to develop predictive methods.
The experiments were performed in different configurations,with plasmas limited by (i) the carbon-tile-protected innerwall, (ii) the outer carbon limiters, and (iii) a magnetic

1
2
3
4

Permanent addresses:
PPPL, Princeton University, Princeton, NJ, USA.
Euratom-IPP Association, Garching, Federal Republic of Germany.
Oak Ridge National Laboratory, Oak Ridge, TN, USA.
Istituto di Fisica del Plasma, Consiglio Nazionale delle Ricerche, Milan, Italy.

457

GONDHALEKAR et al.

458
3-#3506n e 10 19 m~ 3

T e (0) keVPRADPELLET 5°
HtLLLIMW6°
T I H E (s)FIG. 1. B$ = 2.1 T, 1$ = 3 MA, Ohmic heated plasma in deuterium, with 3.6 mm pellet injected at
47 s. The small increase in 1$ at pellet injection is pre-programmed to provide additional power. Thisfeature is not essential.

separatrix formed during single-null x-point operation. Both
ICRF and NBI heating power were applied to the plasma, during
and after pellet injection. The plasma parameters were: major
radius R-3m, minor radius a~1.2m, elongation k~1.H and 2.15B.(T)with target plasma density 0.65^(10'9 m~3)S3
and peak electron temperature 2.3ST e (keV)^6. The ;otal electron
content of the target plasmas was 0.6SN e (10 2 l )£3. Immediately
e
j
y
21
e
after pellet injection
this increased
21 by ANe=3.2(±0.3)x10
pellet)andAN=1.8(±0.2)x10(3.6mmpellet),represente
(^S
)
()
6
ing 71 % and 82? of the pellet mass,respectively. Most of thepellet mass was deposited at 0.25^r/a^0.5, with some pellets
penetrating to within 0.1m of the axis. We present here asurvey of the major findings. Fig.1 shows the temporal evolution of a typical pellet fueled limiter discharge with OH alone.

459

IAEA-CN-47/I-I-6

0

0 . 5

1 . 0

1 . 5

2 . 0

FIG. 2. TP and (I-R) for limiter plasmas.
O and A refer to gas fueled discharges, •

2.

2 . 5

3 . 0

and A to pellet fueled ones.

PLASMA EDGE BEHAVIOUR AND DENSITY LIMIT

Flux of deuterium at the plasma edge, its recycling andconfinement properties have been studied before for gas fueled
JET plasmas [1,2], Outer limiter discharges fueled by a 3.6mm
pellet show a large reduction in the influx of deuterium compared to a gas fueled plasma of the same electron density. The
global particle confinement time T p is thus larger in pellet
fueled plasmas, as shown in fig.2. The particle recycling
behaviour is also markedly different. Fig.2 also shows (1-R)
plotted against n e , R is the recycling coefficient measured 1s
after pellet injection. Whereas for gas fueled cases (1-R)£0,with pellets (1-R)>0 for several seconds after injection. This
corresponds to a net removal of particles from the system, apellet induced pumping.
Reduction is also observed in impurity influxes at the plasmaedge following pellet injection. However, both the ratio of
impurity to deuterium fluxes, and the bolometrically measured
total radiated power normalized to the square of the electron
density do not show any marked deviation from the behaviour ofgas fueled plasmas. Thus pellet injection effects a reductionof I^ff b v reducing the influx of all species, and by diluting
the impurities already present. Zeff<1.5 have been obtained inlimiter plasmas (see fig.2 of [3]). The values are derived frombremsstrahlung measurements taken 0.5s after pellet injection.
Eventually, on a time-scale of =-4s, recycling establishes thehigher Zeff concomitant with edge fueling.

460

GONDHALEKAR et al.

The favourable edge behaviour also helps in accessing higher
plasma density (see fig.3 of [3])- Both 4.6mm and 3.6mm pellets
have been used to probe the limit. The largest increase, bynearly a factor of two above the OH high-density limit indeuterium, was obtained with the 4.6mm pellets at q CV 2=3-6. The
high-density limit in JET is discussed in [4] in terms of column
contraction when the power radiated outside the q,L=2 surface
exceeds the input power. With pellet fueling higher n Q can beachieved for a given n e (q>2).

3.

DENSITY DEPLETION AND WALL PUMPING OF THE PELLET FUELED
INNER WALL PLASMAS

Gas fueled plasmas placed against the large-area carbon tiles
on the inner wall are characterised by a decay of the electron
density with a 1s time constant. Increased wall pumping isevident in the post-pellet phase of the plasma. This is an
important feature for multiple pellet injection strategies.
Pellet fueled inner wall (PFIW) discharges have shown two
distinct modes of density behaviour, fig.3. Some show smoothdecay in which the volume averaged electron density <ne> does
not return to its pre-pellet value, eg.#9226. This behaviour issimilar to that for OH gas fueled inner-wall discharges. In
most PFIW plasmas, <ne> returns to its pre-pellet value in a'scallop-like1 fashion of -200ms period, eg.#9238.
The decrease of <n e >, whether slow or rapid, depends on transport of plasma ions to material surfaces and then absorption ofthese ions by the surfaces. The plasma transport process isclearly different in the two modes of behaviour for the PFIW
discharges. The scallop-like density decay is synchronous withminor disruptions at the q=2 surface, at R~3.85m, as determinedfrom the magnetics data. A simultaneous decrease in T e occurs
at R~3.8m(q<2), whereas spikes of ATe~100eV occur at R-4.1m(q>2)(eg.#9238 at t=50.2, 50.4 and 50.6sec, for pellet injection at50s). This same behaviour is shown also in soft x-ray emissions. Interferometry shows a concurrent density increase atlarge radii. The rise in both n § (a) and T e (a) increases theflux of deuterium to the carbon inner wall tiles. Thus minor
disruptions give rise to rapid transport of plasma fromr=(a-40cm) to r=a. At the edge the particles acquire higher
energy due to increased sheath potential, enabling deeper
penetration into the wall and thus stronger pumping. During thetemperature spikes the D a emission from the inner wall tiles
decreases by a factor of four, showing reduced recycling. Discharges with a smooth decay of <ne> (eg.#9226) do not exhibit
such behaviour.
The ability of the 20m2 of inner wall carbon tiles to absorb
all the deuterium which impinges on it, throughout as many as 20

IAEA-CN-47/M-6

461

n e > 10 1 9 nf 3

1,0
0,8

Second harmonic
ECE
Te(R=3.8m) keV,

0.6
#9238
0.H

TIME (s)FIG. 3. The two modes of density decay for inner-wall plasmas: (a) slow and smooth as in #9226(b) rapid as in #9238, with mini disruptions atq = 2 and corresponding temperature spikes at the edge.

successive discharges with pellet fueling, is not explicable interms of the Local Mixing Model [5] of deuterium trapping ingraphite. Studies to clarify this are in progress.
k. A PELLET INDUCED OSCILLATION AT THE q=1 SURFACE
Soft x-ray imaging arrays have revealed the presence of aunique phenomenon, the 'snaker oscillation, fig.1*. It
frequently appears when a pellet is injected, and manifests

462

GONDHALEKAR et al.

(a)

SAWTOOTH'SNAKE' STOPSROTATING
FIG. 4. (a) Plot of line integrated soft X-ray fluxes showing the 'snake' oscillation after pellet injection.
(b) Tomographic reconstruction showing contours of equal emissivity, and localization of the 'snake'.
(c) Plot as in (a), showing a 'snake' that has stopped rotating and survives a sawtooth crash.

itself as a small region of enhanced x-ray emissivity with am=1/n=1 helical symmetry at the q=1 radius. Analysis has shownthat this increased emission is due to a density increase in the'snake', by up to a factor of two, above the background density,
as well as a drop of -20% in T e compared to the surrounding
plasma. In contrast to the short-lived m=1/n=1 perturbation
rotating in the electron diamagnetic direction observed
previously [6], the 'snake' in JET lasts for many hundreds ofmilliseconds and occasionally up to -1.5s, it rotates at varying

IAEA-CN-47/U-6463

frequencies in the ion diamagnetic direction, it sometimes comesto a complete standstill and then resumes rotation. Sawtooth
disruptions occur during this activity, without dissipating the1
snake'.
One plausible hypothesis for the 'snake' is that, an m=1/n=1magnetic island is created by the high density ablation cloudfrom the pellet as it crosses the q=1 surface. This island thentraps the excess density, forming the helical 'snake'. If thishypothesis is correct, the radial location of the 'snake' is an
accurate indicator of the radius at which q=1. We thus inferthat the q=1 region expands by 35% during the sawtooth ramp.
5.

ELECTRON DENSITY PROFILES AND RELATED TRANSPORT

Immediately after pellet injection, the electron density
profile is hollow (dne/dr>0). ru(r) recovers to a centrally
peaked shape in ~200ms. The evolution towards a peaked profile
is discontinuous when sawteeth occur. An analysis of profile
evolution of pellet fueled plasmas using a model of electron
particle flux r = -DVne+Vne indicates that the diffusion
coefficient D(r; and convection velocity V(r) fall in with thosefor gas fueled ones [7,2], which showed a large scatter, but nodependence on n e , 1^, Bx or heating mode. Analyses of indi^vidual pellet injection cases show that in the outer limiter
configuration, V/D increases from ^0.2aTl before pellet
injection to -O.^nT1 immediately after, reverting to -0.2m"1
1.5s later, as evidenced by the more peaked density profile
after pellet injection, which subsequently returns to its
original form. In the singles-null x~point configurations, V/D
increases from '0.3m'1 before pellet injection to ^O.Srrr1 after,
giving a very peaked density profile that is maintained for along time. Eventually V/D decreases to -O.iJnT1.

6.

PELLET ABLATION MODELLING

Accurate determinations of penetration of the pellet into theplasma and ablation rate are made using a combination of a soft
x-ray camera and D a detectors viewing the pellet trajectory. As
shown in fig.5, the penetration of the pellet into the plasma
for all the data considered is successfully modelled with aneutral gas shielding model, modified to include attenuation of
the electron heat flux by a cold ionized plasma surrounding theneutral gas [8]. Also included are a multi-group treatment of
the incident electrons, plus self-limiting ablation in thecollisionless plasma limit. With present heating power levels,the pellets traverse regions with fast ion density 11^=3x1018nT3
for NBI heating, and n fi <10 l8 nT 3 for ICRF heating. No
additional ablation processes associated with such a fast ion
population during ICRF or NBI heating have yet been identified.

GONDHALEKAR et al.

464
120_
100.
80.

Neutral and Plasma Shielding<^>

1mm Plasma Shield Thickness

X

60 _4.6mm OH
40.

o

3.6mm OH+

4.6mm ICRH

20.

D

3.6mm ICRH

X

3.6mm NBIo0
(Limiter)

i

i

20

40

i
60

i

80

100

120

d(cm) — expt

FIG. 5. Comparison between measured penetration, A (cm) - expt, and calculated one,A(cm) — calc. The data set includes 4.6 mm and 3.6 mm pellets, with OH, OH + ICRF heating as
pKRF <, 4.6 MW, and OH + NBI heating at PNB, < 4 MW.

7.

ICRF HEATING OF PELLET FUELED PLASMAS

An important limitation of ICRF heating in JET is theuncontrolled density increase during the RF pulse. Influx of allneutral species from the limiters and wall increases roughly inproportion to the applied power, and Z e f f and the ratio Prod/Pinremain nearly constant. As a result the density limit during
ICRF heating is at most 10$ higher than the corresponding value
for ohmic heating alone. We have probed this limitation by
applying ICRF heating immediately after pellet injection. We
have observed that the initial density increase effected bypellet injection can be maintained with ICRF heating and thehigh-density limit for OH+ICRF heating can be extended. Both
4.6mm and 3.6mm pellets have been used. The resulting increases
in temperature, for both electrons and ions, in the 3He minority
heating mode are consistently slightly larger than values
obtained with gas fueling under similar plasma conditions andfor the same normalized ICRF power. Our future work withmultiple pellet injection will further explore and optimize ICRF
heating of a densified plasma core.
8.

SUMMARY

Deuterium pellets have been injected into plasmas with OH,
OH+ICRF and OH+NBI heating. Plasmas with different magnetic
configurations were probed. It is shown that pellet fuelingIAEA-CN-47/I-I-6

465

gives access to regimes of higher plasma density and increased
plasma purity by modification of edge behaviour. Also, pellet
injection modifies the electron density transport in the bulk
plasma, most markedly in the single-null x-point configuration,
giving very peaked density profiles. These effects persist forseveral seconds. Careful modelling of pellet ablation andpenetration measurements has been performed. At present levelsof auxiliary heating power, no additional ablation processes
associated with application of ICRF or NBI power have been
identified. A 'snake' oscillation characterized by an island ofdense cooler plasma at the q=1 radius is observed. Its small
dimensions (Alr=15cm, AlQ=25cm) and long lifetime (T=0.2-»1 ,5S)
imply greatly reduced energy and particle diffusion compared tothe surrounding plasma.
A critical issue is whether profiles obtained with pellet
injection have better confinement properties than those produced
by gas fueling [9,10,11,12], With single pellet injection in
JET the behaviour of energy confinement is unclear. In thetransient period immediately after injection of a pellet, theconfinement time is much larger than the corresponding value
with gas fueling, as observed in other experiments employing
single pellet injection. However, the duration of the period ofapparent improved confinement is shorter than xE. The
significance of this observation is being investigated. An aim
of our future work with multiple pellet injection will be to
explore the feasibility of effecting a sustained improvement in
confinement properties.
ACKNOWLEDGEMENTS
Dedication and enthusiasm of many people has benefited theseexperiments. The authors are grateful to L Lamb, J-L Bonnerue,
J Ryan and K Scheidt for superb technical support, to C Wilson,
J Reid, P Roberts and A Tiscornia for excellent engineering
support and to C Andelfinger, K Buchl and H-B Schilling of IPP,
Garching, for manufacture of the pellet injector, and for strong
support during installation.

REFERENCES[1]

MORGAN, P.D., et al., in Controlled Fusion and Plasma Physics (Proc. 12th Europ. Conf.
Budapest, 1985), Vol. 9F, Part n , European Physical Society (1985) 535.
[2] CHEETHAM, A., et al., in Controlled Fusion and Plasma Heating (Proc. 13th Europ. Conf.
Schliersee, 1986), Vol. IOC, Part I, European Physical Society (1986) 240.
[3] JET TEAM, in Plasma Physics and Controlled Nuclear Fusion Research 1986 (Proc. 1 lth Int.
Conf. Kyoto, 1986), Vol. 1, IAEA, Vienna (1987) 31.
[4] SCHULLER, F.C., et al., in Controlled Fusion and Plasma Physics (Proc. 12th Europ. Conf.
Budapest, 1985), Vol. 9F, Part I, European Physical Society (1985) 151.

' 4 6 6 •;[5]
[6]

GONDHALEKAR et al.

&WCE, D.K., et al., J. Nucl. Mater. 111&112 (1982) 598.
I^ARKER, J., et al., Rep. P F C / R R - 8 5 - 8 , Plasma Fusion Center, Massachusetts Institute ofTechnology, Cambridge, MA (1985).
[7] GONDHALEKAR, A., et al., Bull. Am. Phys. Soc. 30 (1985) 1525.
[8] WATKINS, M.L., et al., in Controlled Fusion and Plasma Heating (Proc. 13th Europ. Conf.
Schliersee, 1986), Vol. IOC, Part I, European Physical Society (1986) 156.
[9] GREENWALD, M., et al., Phys. Rev. Lett. 53 (1984) 352.
[10] SCHMIDT, G.L., et al., in Controlled Fusion and Plasma Physics (Proc. 12th Europ. Conf.
Budapest, 1985), Vol. 9F, Part II, European Physical Society (1985) 674.
[11] ODAJIMA, K., et al., in Plasma Physics and Controlled Nuclear Fusion Research 1986 (Proc.
11th Int. Conf. Kyoto, 1986), Vol. 1, IAEA, Vienna (1987) 151.
[12] NIEDERMEYER, H., et al., in Controlled Fusion and Plasma Heating (Proc. 13th Europ. Conf.
Schliersee, 1986), Vol. IOC, Part I, European Physical Society (1986) 168.

IAEA-CN-47/M-7

FIRST EXPERIMENTS ON THE
GASDYNAMIC TRAPJ

P.A. BAGRYANSKIJ, A.A. IVANOV, V.V. KLESOV,Yu.L. KOZ'MINYKH, LA. KOTEL'NIKOV,
Yu.I. KRASNIKOV, A.A. PODYMINOGIN,
A.I. ROGOZIN, G.V. ROSLYAKOV, D.D. RYUTOV
Nuclear Physics Institute,'
Sibirian Branch of the
USSR Academy of Sciences,
Novosibirsk, Union of Soviet Socialist Republics

Abstract
FIRST EXPERIMENTS ON THE GASDYNAMIC TRAP.
The paper describes an experimental device called "gasdynamic trap" and reports the results ofthe first experiments on it. The gasdynamic trap is a mirror machine with a spacing of 7 m between
the mirrors, minimum magnetic field of 0.2 T and a mirror ratio varying between 25 and 80. The
preliminary plasma generated in the gas discharge source is injected into the trap through one of themagnetic mirrors. The plasma is then heated by the injection of six atomic hydrogen beams with a total
power of 5 MW, an atom energy of 20 keV and a pulse duration of 0.25 ms. In the different operating
regimes of the plasma source, the trap could be filled with plasma at densities from 5 X 1012 cm"3 to10' 4 cm"3 and temperatures in the range from 3 eV to 30 eV. The plasma filling dynamics of the trap
are in good agreement with gasdynamic evaluations. During filling the plasma in the trap is MHD stable,
regardless of the shape of the field lines in the expanders. This indicates the presence of stabilizationof the plasma at the ends, which is ensured by the high electrical conductivity of the plasma in thesource. After the source is switched off, the plasma MHD stability becomes sensitive to the shape ofthe field lines in the expanders. In particular, the plasma decay time in the trap is greater when the fieldlines in the expander have a favourable curvature than when the curvature is unfavourable.

1.

INTRODUCTION

The gasdynamic or GD trap [1,2] is an axisymmetric mirror machine with avery large mirror ratio and a length exceeding the effective free path of the ions inrelation to scattering into the loss cone.
An important advantage of the GD trap is that the rate of longitudinal losses
from it is totally insensitive to microfluctuations which may occur in the plasma,i.e. to an effect which represents a serious danger for many other types of open trap.
Another advantage is that MHD stability of the plasma in the GD trap can beensured within an axisymmetric magnetic configuration [1,2]. This is due to the factthat in the GD trap the plasma beyond the mirror, which is situated in a region offavourable curvature of the lines of force and exerts a stabilizing influence on thesystem as a whole, has a relatively high density (for more details see Ref. [3]).
467

468

BAGRYANSKU et al.
7000

6

5 4

Fig. 1. Schematic diagram of device:
1 - Central vacuum chamber; 2 - Coils of solenoid part of magnetic system; 3 - Internal mirror coils
for 110 kG; 4 - Coils determining magnetic field geometry in the expanders; 5 - Plasma absorbers;
6 - Plasma source; 7 - Atomic beam injectors; 8 - Expander vacuum chambers.

The prospects of the GD trap as a fusion reactor are discussed in Refs [1,2,4-6]. The key element and the most complex one in such a reactor is the mirror coil,which has to generate a magnetic field of several hundred kilogauss. What makes it
somewhat simpler to solve this problem is that such a strong field is required in a
volume with a diameter of only 10 cm and a length of 10-20 cm. The length of areactor based on the GD trap would be fairly large (1 km), but it should be bornein mind that all of it would be occupied by a simple axisymmetric solenoid with afield of only 15 kG.
The GD trap also offers interesting possibilities as a neutron source formaterials research [6]: oblique injection of 10-20 MW of fast 240 keV tritons intoa relatively cold deuterium plasma can give a neutron flux of the order of
3 MW-m~ 2 (at the plasma surface).
A prototype device of this kind has been built at the Novosibirsk Nuclear
Physics Institute in order to study the physics of GD traps. Section 2 gives a generaldescription of the machine and of experiments that can be performed on it. Section 3
discusses the problem of plasma MHD stability in the device. Lastly, Section 4
describes the results of the first experiments with the preliminary plasma, which confirm the stabilizing effect of the expanders.

2.

EXPERIMENTAL DEVICEIn choosing a design for the device, our aim was, first, that it should be suitable
for the widest possible spectrum of physical experiments and, second, that it should
be as simple as possible to make. In view of this approach, the machine was designed

469

IAEA-CN-47/I-I-7

-400

-200

200

400

Z (cm)

Fig. 2. Magnetic field on trap axis:
1 - Magnetic field distribution when the trap is being filled by one plasma source; 2 - Magnetic field
distribution at the plasma confinement stage.

not to obtain a quasi steady state plasma but to perform experiments with a decaying
plasma. As a result we can substantially reduce the requirements for the vacuum
system (the quantity of gas entering the chamber per pulse is small), the magnetic
field generation system (condenser batteries can be used as energy sources; forced
coil cooling is not necessary) and the injection system. A schematic diagram of thedevice is shown in Fig. 1.
The experiment scenario consists of the following sequence of plasma preparation, heating and decay processes. A relatively cold (~ 10 eV) preliminary plasma
is injected into the device from plasma guns located in the expanders (Fig. 1). The
magnetic field in the mirrors is somewhat weakened during the operation of the guns.
After the plasma density in the traps has attained the necessary magnitude of~10 1 4 cm' 3 , the guns are switched off and the field in the mirrors increases(i.e. the mirrors 'close'). Figure 2 shows the magnetic field on the trap axis at thestage when the trap is being filled with plasma and at the stage of plasma decay. Afterthe guns are switched off, the plasma located in the expander region flows rapidly
to the ends in a time Ttexpailder ~ Lexpander VM/f e . The time of outflow of the cold
plasma from the trap is much greater: 7piasma = (LR/a) VM/(2TTT) ^ Tafmia.
After the plasma has flowed from the expanders to the ends, pulsed atomic hydrogenbeams are injected into the trap from six START injectors [7]. The beams are
injected at an angle of 45° to the axis of the device in order to increase the power
contained in the plasma. It is thus possible during an injection pulse of 0.25 ms to
inject an energy of 1.5 kJ into the plasma in the form of 20 keV fast protons. The
fast protons are slowed down in the plasma, transmitting their energy mainly to electrons, in a time [8]:3MT 3/2
8 V2TT e4XVm n

470

BAGRYANSKU et al.

TABLE I. PARAMETERS OF THE DEVICES
Spacing between mirrors (m)7Expander length (m)1.8

Maximum magnetic field in the mirrors (kG)160

Maximum magnetic field in the solenoid (kG)

2.2

Magnetic field pulse duration in the mirrors (ms)10Magnetic field pulse duration in the solenoids (ms)100Energy of injected atoms (keV)

20

Equivalent current of six atomic beams (A)350

Duration of injection (ms)0.25

The electrons, in their turn, heat the plasma ions through collisions. The time ofequalization, T% of the electron and ion temperatures approximately coincides with
Tb- In the case of a plasma having a density of n = 1014 cm"3 and a temperature of
Tj = Te = 100 eV, 7f = if; = 2 x 10"4 s. The time of outflow of the plasma
heated to 100 eV from the trap is much greater, amounting to several milliseconds.
Table I gives the basic parameters of the device.
Temperature T, density n and plasma radius a can be varied over a wide range.
The magnetic field in the homogeneous sector can be reduced by half. Since all thesechanges can be made independently of each other, the basic dimensionless parametersof the system,

can also be varied over a wide range. The possible variants of the plasma parameters
are given in Table II.
Apart from the notations introduced earlier, we give in Table II the lifetime ofthe plasma in the case of anomalous diffusion across the magnetic field over a
distance of the same order as the plasma radius a with a diffusion coefficient equal
to the Bohm coefficient. This time is governed by the relation
16 eHa2cT
The mirror ratio can be changed by varying the magnetic field in the main part of
the trap and the plasma density by varying the energy injected into the gas discharge

471

IAEA-CN-47/I-I-7

TABLE II. EXPECTED PLASMA PARAMETERS
Variant
Plasma parameters
2

3

4

100

100

100

100

4

4

10

10

1
T(eV)
n ( x 10 13 cm~ 3 )

1.1

2.2

1.1

Ho*, (kOe)

2.2

Ryac

72

120

72

120

a (cm)

11

15

11

10

N

14

9

13

4

0.07

0.26

0.15

0.5

Vpiasmaflitres)r* (ms)

140
0.8

280
0.8

140

140

0.3

0.3

Tplasma ( m s )

3.2

6.2

3.5

7.5

TB (ms)

0.425

0.4

0.425

0.176

in the plasma guns. The plasma radius can be controlled by changing the current in
the pulse solenoids which generate the magnetic field in the guns at the stage when
the trap is being filled with plasma.
Open traps (including the gasdynamic trap) differ from other magnetic confinement systems in having a relatively small value of the parameter N (in an open trap
reactor N = 10 to 30 and in a tokamak reactor N = 150 to 200); moreover, mostopen trap designs include a long (L = (100 to 500)a) homogeneous magnetic fieldsector. Both these features can strongly affect the excitation of drift instabilities
which cause transverse plasma transport. The gasdynamic trap model offers favourable opportunities for the study of anomalous transport in a system with N ~ 15 andL/a — 50 under conditions not masked by the lack of axial symmetry of the system.
Firstly, since the model allows N to be varied easily in the range from 10 to 30, it
is possible to cross the threshold for the excitation of high frequency drift instabilities(N « 20, see Ref. [9]). Secondly, the value of L/a S 50 is quite sufficient for low
frequency drift instabilities to appear fully. Thirdly, the model enables us toinvestigate a range of 0 values from 0 to 0.4, which is quite sufficient to reveal thepossible role of the finite plasma pressure in stabilizing drift instabilities. Fourthly,
reducing the temperature to 40-50 eV and increasing N to 30-40 will make it possible to study the influence of collisions on low frequency drift instabilities.
The microinstabilities caused by sloshing ions can also be studied at the neutral
injection stage.

472

3.

BAGRYANSKU et al.

PLASMA MHD STABILITY

The plasma fills only the near axial region in the trap. In this case, the frequency(growth rate) of flute perturbations can be determined by the formula [3]where pi and p± are the longitudinal pressure and the transverse pressure of theplasma, p is plasma density, v the flow velocity along the magnetic field, r = r0
VB~/B(z) the distance from the axis to the field line which in the trap equatorial
plane passes at a distance r0 from the axis and B ^ the magnetic field in the trap
equatorial plane, the values of p, pt, p± and v being taken on the field line passing
through point r0.
After the fast ions have been slowed down and have transmitted their energy
to the plasma, the plasma pressure becomes isotropic, pi + p± = 2p. The plasma
flow velocity in the inner part of the trap is small (apart from the regions near the
mirrors, which make only a small contribution to Eq. (1)) so that the integral, overthe inner part of the trap in the numerator of expression (1), is equal to-Hi)'Mi)-(the integration is performed between the mirrors, and the pressure in this region isconsidered to be homogeneous). For radially decreasing plasma pressure the integral
sign corresponds to instability.
To reduce the destabilizing contribution of the central part of the mirror
machine, we have to minimize the integral in expression (2). It can be shown thatthe optimum profile of the field lines between the mirrors is defined by the relation(3)

where R > 1 is the mirror ratio, L the spacing between the mirrors and the coordinate z starts from the trap equatorial plane. The necessary calculations can be

IAEA-CN-47/I-I-7

473

found in Ref. [10]. In the design of the device, the position of the coils and thecurrents in them were chosen to bring the field lines as close as possible to the profile
in Eq. (3).
The system is stabilized by creating a favourable curvature for the field lines
in the expanders. With a view to enhancing this effect, the expanders are provided
with special coils (4 in Fig. 1), in which the current direction is opposite to that in
the other coils. This has resulted in a situation where the stabilizing contribution of
the expanders is three times as great as the destabilizing contribution of the central
part of the device.
At the neutral injection stage, until the fast ions are slowed down in the plasma,
they can contribute substantially to the pressure, which will consequently beanisotropic. Therefore, the contribution of the central part of the trap to the stability
integral is given by an integral of the form

(4)

where h refers to fast ions. As for the stabilizing contribution of the expander, thisis associated with the cold plasma stream flowing out from the expander and isroughly proportional to the pressure of the main plasma inside the trap. At the initialinstants of injection, when the electron temperature is low, the slowing-down timeof the fast particles is also small — and consequently their pressure is small. For thisreason, only the pressure of the main plasma p contributes to the integral in (4), andthe destabilizing contribution of the central part of the trap continues to be smaller
by a factor of three than the stabilizing contribution of the expanders. However, by
the end of the injection pulse the relative magnitude of the pressure of the hot component increases, so that the integral in (4) also increases relative to the stability integral
for the expanders. Estimates show that the stabilization is maintained also in thiscase, although the safety factor is now small. The actual situation will depend on thedetailed fast neutral capture conditions.
If it is found that at the final instants of the injection pulse the system is going
to become unstable, there are two ways of remedying the situation. The first is toincrease the injection pulse duration by non-simultaneous switching on of the six
injectors, which reduces the relative fast particle contribution to pressure. The second
method is to make certain readjustments of the magnetic configuration of the systemnear the fast particle stopping points in order to create a favourable curvature of the
field lines in this region [11, 12].
In concluding this section, let us note that the angular velocity of the fast ion
magnetic drift in our case is smaller than the characteristic flute oscillation frequency
estimated by formula (1). This demonstrates that a hydrodynamic description of theplasma is possible. Furthermore, the effects of the finite Larmor radius of the hotcomponent are already substantial, and this should increase the stability of the higher
modes.

474

BAGRYANSKU et al.

t (ms)

Fig. 3. Plasma density at the trap centre as a Junction of time for different field line geometries in the
expanders (see Fig. 4). The time is reckoned from the instant when plasma filling of the trap iscompleted.

4.

RESULTS OF THE FIRST EXPERIMENTS

The filling of the trap with hot plasma was carried out in different configurations of the expander magnetic field for a constant magnetic field in the trap with amirror ratio of R = 25 and a minimum magnetic field of 1.5 kG. In the first filling
regime the magnetic flux, corresponding to a diameter of 20 cm at the trap centre,
was limited at the plasma source anode outlet (4 cm in diameter) by switching on thepulse solenoid of the source. In this regime, the field lines in the expander where thesource was located had the shape shown schematically by curve 1 in Fig. 4, forming
an additional mirror machine with a spacing of 0.7 m between the mirrors and amirror ratio of 20-40. In this magnetic field configuration, by the end of thedischarge current pulse in the plasma source — which had a duration of 4 ms —, aplasma accumulated in the trap having the density profile shown in Fig. 5 (curve 1)
with a maximum density of 1014 cm"3 and a temperature of T ~ 10 eV. In thesecond regime, the pulse solenoid of the source was not switched on. Here the shapeof the field lines in the expanders corresponded qualitatively to curves 2 and 3 inFig. 4, depending on the current direction in the expander coils, while the magnitudeof the magnetic field at the source was about a tenth of that at the trap centre.
Nevertheless, in this case too, towards the end of the discharge current pulse, a wide
radial profile was seen for the plasma in the trap (curve 2 in Fig. 5); however, theplasma density and temperature were lower than in the regime with the solenoid
switched on — n = 1 x 1013 cm" 3 at maximum for a temperature of T = 5 eV.

IAEA-CN-47/I-I-7

475

Fig. 4. Field line geometry in the expanders:
1 - With plasma source solenoid switched on; 2 - with plasma source solenoid switched off and coilson the expander volumes 'co-switched on' (see Fig. 1) with the coils of the solenoid part of the magnetic
system; 3 - with the expander coils 'counter-switched on'.

By changing the current direction in the expander coils, it was possible tochange the curvature of the field lines in the expander regions remote from themirrors. This involved virtually no change of the magnetic field in the central partof the trap; the only thing that changed was the stabilizing contribution of the
expanders to the MHD stability of the plasma, from near zero (curve 2, Fig. 4) tothe maximum favourable contribution (curve 3, Fig. 4).
In both regimes of trap filling, the plasma density in the trap increases monotonically over the entire period of operation of the source.
The absence of fast transverse losses, in spite of the presence of an MHD
unstable configuration in the first regime (a 'steep' mirror machine between the gun
and the mirror), indicates MHD stabilization at the ends brought about by the hightransverse electrical conductivity of the plasma in the source. After the source has
been switched off, the MHD stability of the plasma becomes sensitive to the shapeof the field lines in the expanders. Figure 2 shows experimental dependences ofplasma density in the trap on time, determined from the attenuation of the probing
atomic beam. Similar dependences are detected by the Langmuir probes and by thesensors recording the end beams of ions passing through the mirrors to the expander.
In the case where the curvature of the field lines in the expanders is not favourable,
regardless of how the unfavourable curvature is created (by switching on the source
solenoid or by switching over the current in the expander coils), the plasma decay
time in the trap is substantially smaller than in the case where the curvature of the
field lines is favourable (with the source solenoid not switched on). In the MHD
stable magnetic field configuration, the plasma decay time is close to that estimatedfrom longitudinal plasma losses in the regime of gasdynamic plasma outflow throughthe mirrors. In the MHD unstable configuration, the plasma decay time is several
multiples of the longitudinal time-of-flight of ions through the trap, so that the probes
located in the trap at different distances from the mirrors and the sensors recording

476

BAGRYANSKU et al.

13 ™ 32-n m a x =1.1X10 1 3 cnV.-3

20

Fig. 5. Plasma density distribution over the radius in the central part of the trap:
1 - When the trap is filled with the plasma source solenoid switched on; 2 - When the trap is filled
without the source solenoid; the geometry of the magnetic field in the expanders corresponds to curve 3
in Fig. 4.

end losses of ions show flute-type plasma density perturbations. The above
experimental results as a whole confirm our ideas regarding the mechanism of MHD
stabilization of plasma in a gasdynamic trap.

REFERENCES[1] MIRNOV, V.V., RYUTOV, D.D., Pis'ma Zh. Tekh. Fiz. 5 (1978) 678.
[2] MIRNOV, V.V., RYUTOV, D.D., in Voprosy atomnoj nauki i tekhniki (Seriya Termoyadernyj
Sintez) (Problems of Atomic Science and Technology, Series Nuclear Fusion) No. 1(5)(1980) 57.
[3] NAGORNYJ, V.R., RYUTOV, D.D., STUPAKOV, G.V., Nucl. Fusion 24 (1984) 1421.
[4] RYUTOV, D.D., in Mirror Based and Field Reversed Approaches to Magnetic Fusion (Proc.
Course Varenna, 1983), Int. School of Plasma Physics (1984) 72.
[5] MIRNOV, V.V., NAGORNYJ, V.P., in Voprosy atomnoj nauki i tekhniki (SeriyaTermoyadernyj Sintez) (Problems of Atomic Science and Technology, Series Nuclear Fusion)
No. 3(16) (1984) 40.
[6] KOTEL'NIKOV, I.A., MIRNOV, V.V., NAGORNYJ, V.P., RYUTOV, D.D., in Plasma
Physics and Controlled Nuclear Fusion Research 1984 (Proc. 10th Int. Conf. London, 1984),
Vol.2, IAEA, Vienna (1985) 309.
[7] DAVYDENKO, V.I., ROSLYAKOV, G.V., SAVKIN, V.Ya., in Voprosy atomnoj nauki itekhniki (Seriya Termoyadernyj Sintez) (Problems of Atomic Science and Technology, Series
Nuclear Fusion) No. 2(12) (1983) 67.
[8] TRUBNJJKOV, B.A., Voprosy teorii plazmy (Problems of Plasma Theory) No. 1, Atomizdit,
Moscow (1963).
[9] MKHAJLOVSKIJ, A.B., Teoriya plazmennykh neustojchivostej (Theory of Plasma
Instabilities), Vol.2, Atomizdat, Moscow (1977).
[10] BUSHKOVA, O.A., MIRNOV. V.V., Rep. 85-103, Nuclear Physics Institute, Siberian Branchof USSR Acad. Sci., Novosibirsk (1985) (in Russian).
[11] HINTON, F.L., ROSENBLUTH, M.N., Nucl. Fusion 22 (1982) 1547.
[12] KOTEL'NIKOV, I.A., ROSLYAKOV, G.V., RYUTOV, D.D., Rep. 86-31, Nuclear Physics
Institute, Siberian Branch of USSR Acad. Sci., Novosibirsk (1986) (in Russian).

Session L
SUMMARIES
Chairman
R.S. PEASE
United Kingdom

IAEA-CN-47/L-M

SUMMARY ON
TOKAMAK EXPERIMENTS*
A. GIBSONJET Joint Undertaking,
Abingdon, Oxfordshire,
United Kingdom

1.

INTRODUCTION

There have been many important advances in tokamak research since the1984 Conference. A number of new devices have been brought into full operation, ranging from JT-60 (an advanced large tokamak from Japan — a country
with a well established and mature fusion programme) to HL-1 (a medium sized
tokamak from China - a country whose fusion programme has recently embarked
on a considerable expansion). Our Chinese colleagues are to be congratulated on
having achieved full Ohmic tokamak operation with all the usual features and on
providing a useful plasma for future research.
The tokamak has always been an enchanted system. Through the years it
has proceeded from success to success despite the ineptness of experimentalists
and in spite of the cleverness of theoreticians. The secret of this success may bethat, so far, fate alone has had control of the current density distribution; in thetokamak, it is this current density distribution which controls the magnetic shear,
the stability properties and indeed the form of the magnetic configuration itself.
Many physicists now believe that further advances depend upon securingexternal control of this vital feature, in being able to impose a current density
profile rather than having to accept the one given by inductive current drive andill defined wall boundary conditions. The obvious approach is to replace atleast some of the plasma current by non-inductively driven current. The difficulty
here is that the available methods are rather inefficient.
One of the most important results achieved on JT-60 is the demonstration oflower hybrid current drive on a large tokamak system. This is the first demonstration on a large tokamak and our Japanese colleagues are to be congratulated on
having achieved significant results with 1.2 MW of lower hybrid power at 2 GHz,driving up to 1.5 MA of current, with substantial current drive observed at meandensities of up to 2 X 1019 m~3. An important, hoped for, benefit of operation
in a large tokamak configuration is improved current drive efficiency. On JT-60,
a synergetic effect of neutral beam heating on lower hybrid current drive has
been observed and in this configuration 0.7 MA of non-Ohmic current has been
* This summary talk was also published in Nucl. Fusion 27 (1987) 481.

479

480

GIBSONTABLE I. JT-60 LOWER HYBRID CURRENT DRIVE
Frequency

2 GHz

Lower hybrid power

Up to 1.2 MW

Driven current

Up to 1.5 MA

Plasma density

Up to 2 X 1019 rrf3

Efficiency,
driven-"VR

Up to 1.7 A-m"2-W"' (no beams)
Up to 2.8 A-m~2-W~' (with beams)

Pobtained, driven by 0.7 MW of lower hybrid power in the presence of 7 MW ofbeam heating (see Table I). This gives a figure of merit I- ne • R / P L H °f
2.8 X 1019 A-m" 2 -W" J , which is about three times the value obtained in smaller
experiments.
These results will give increased impetus to Ohmic current replacement
experiments and we can expect new results in time for the 1988 Conference.

2.

SAWTOOTH BEHAVIOUR

Let us now look at some of the benefits which might come fromestablishing control over the current density profile. The most impressive
results to date concern the internal disruption relaxation - the 'sawtoothoscillation'. This form of instability has been a feature of tokamak operation
since the very first experiments were carried out in the USSR in the late 1950s
and early 1960s. Only very recently have cases been reported where thisoscillation is stabilized for long periods.
Figure 1 shows an example of spontaneous sawtooth stabilization in JET.
It is spontaneous in the sense that, while there is a well defined recipe for its
production, there is no control of a directly relevant physical quantity (forinstance, current density distribution). The diagram shows the electron temperature profile at a number of times, indicated in the insert, during a pulse exhibiting
sawtooth behaviour. Also shown is a profile obtained during the Ohmic phase of
the same discharge, when the sawtooth amplitude is very small. During thenormal sawtooth behaviour with neutral beam heating only, the flat profile atthe bottom of the sawtooth is about 0.5 keV higher than the Ohmic profile.
At the top of the sawtooth the profile becomes much more peaked and thecentral temperature exceeds the Ohmic value by 1.5 keV. The application ofion cyclotron heating (ICH) induces a stabilization of the sawtooth behaviour

481

IAEA-CN-47/L-I-1

2.8

3.0

32

3.4

36

38

4.0

4.2
R[m]

FIG. 1. Spontaneous stabilization of sawtooth oscillation in JET.

which lasts for 0.9 s (even longer in some cases). For the first 0.3 s of thisperiod the central electron temperature continues to increase, reaching 7.5 keV(5 keV more than the Ohmic value).
The JET authors believe that the mechanism for the sawtooth behaviour isassociated with the existence of a very flat current density profile inside theinversion radius. The application of non-Ohmic heating is supposed to broaden
the current profile very slightly, taking q 0 just above one to stabilize the mode.
There is evidence for this behaviour in that during the flat top of the 'monster
sawtooth' a very low level of MHD activity is observed, indicating that them = 1 mode is stabilized. Furthermore, at the sawtooth crash which terminates
the 'monster', the inversion radius has moved outwards, suggesting again thatj (r) has broadened.
However, Faraday rotation measurements on Textor appear to demonstrateconclusively that there are at least some instances of sawtooth behaviour where
q on axis is and remains significantly below one. The sawtooth behaviour inthese discharges is clearly not amenable to an explanation based on small changesof the axis q-value from one. This Jiilich group suggests that the sawtooth
behaviour is determined by the current density profile near the q = 1 surface(which is typically in the steep part of the temperature profile, between centre
and edge). In extreme cases the flattening of the profile near q = 1 to stabilize
the m = 1 mode can actually be measured and the sawtooth relaxation is
supposed to hold the profile near this marginally stable condition. The Jiilich
group expects this process to lead to a universal current density j (r) profile in482

GIBSON
TFTR
1.0
1
(V)

kA

Plasma850

Beam Drive260

Bootstrap

350

J'l

Calculated

Measured-05
4.0
Timo (s)FIG. 2. Evidence of the bootstrap current in TFTR.

terms of a suitable combination of external quantities (i.e. j in units of BT//a0R>
and r in units of (<x0RI/B)^2). It has been shown by direct measurements that
this is true for a range of Ohmic profiles and for the few profiles which havebeen measured in ICH discharges. These measurements would seem to provide,at last, some of the direct evidence that is needed to guide the debate on profile
establishment and profile consistency in tokamak discharges.
Stabilization of the sawtooth mode by replacing some of the Ohmic
current by lower hybrid current drive has been reported in many experiments(e.g. PETULA, ASDEX, PLT and ALCATOR). In PETULA and ALCATOR
the experiments extend to densities in excess of 1020 m~3. Also, it seems thatin these experiments a simple, small modification of the central current density
profile is not sufficient to explain the stabilization. Thus there are often regions
where an m = 1 oscillation remains (implying q < 1) and the inversion radiusdoes not change as the lower hybrid power is increased to the threshold forstabilization.

3.

BOOTSTRAP CURRENT

A further twist to the argument as to who shall control the current density
profile is introduced by the firstclear evidence for neoclassical pressure driven
current, the 'bootstrap current', which was predicted 15 years ago by
R.J. Bickerton, J.W. Connor and J.B. Taylor. Now, a careful comparison of

IAEA-CN-47/L-I-1

483

TABLE II. OPERATIONAL MEGAWATT LEVELHEATING SYSTEMS ON TOKAMAKS
Neutral beam injection(NBI)
JT-60(20MW), TFTR(20MW),
JET(10MW),DIII-D(6MW),
ASDEX(4.4MW)
Ion cyclotron heating(ICH)
JET (7 MW), PLT (4.5 MW),ASDEX (2.6 MW), TEXTOR (2.5 M W )Lower hybrid resonance heating(LHRH)
ALCAT0RCU.5MW), JT-6O (1.2MW),
PLT (1 MW), ASDEX (1 MW),PETULA (0.5 MW)Electron cyclotron resonance heating(ECRH)
T-10(2MW),TFR(0.6MW),
JFT-2M (0.2 MW), T-7 (0.4 MW),CLEO(0.2MW)
Alfven wave heating
TCA(0.6MW)

the experimentally measured and the theoretically predicted loop voltages incertain TFTR discharges gives a strong indication of the reality of this effect.
An example of such a discharge is shown in Fig. 2, where it should be notedthat about 40% of the current is driven by this neoclassical dynamo effectwhile about 30% is driven ohmically and about 30% is driven by injected beams,
the latter contribution having been minimized by using balanced injection.

4.

PLASMA HEATING

Just as we may now be reaching the point at which the inherent tokamak
current profile is no longer adequate for our purpose, so we have already passed
the point at which the inherent tokamak heating mechanism can no longer
produce the required temperatures. For a number of years, many of theresources of fusion research have been devoted to establishing and understanding
additional heating methods for the plasma. This work has come to full fruition

484

GIBSON

DIVERTOR

LIMITERf

FIG. 3. Electron temperature heating in JT-60.

485

IAEA-CN-47/L-I-1
25v £*-%:-- —

20

20keV

=0.6-1.1 (• Fe XXV 1.85 A (Vert.)
MA
U Fe XXIV 192 A

15

10
DO O

5-

1

"aT|
Pin,

OA

8

off,

n

*

2

3
4
P1NJ/n. (10-13W-rri>)5

s3.8xiO"keV-W-'-nr 3 in supershots(s1.5x10" in other shots)

FIG, 4. Production of high ion temperatures in TFTR.

at this Conference; a number of heating methods are routinely applied at themultimegawatt level to many tokamaks (see Table II). These heating methods
have now reached such a level of success that the emphasis has changed fromunderstanding the heating to understanding the plasma behaviour. Examples
are the study of energy confinement using modulated ICH by the Belgian groupon Textor, and the use of ECH on T-10 and of ICH on JET as spatially localizedheat sources to probe the energy transport.
The heating methods have been very successful in increasing the plasma
temperature. Two examples of successful heating from the very many available
are shown in Figs 3 and 4. First, Fig. 3 shows neutral beam heating on JT-60
with and without divertor; note the successful heating in each case (ATe > 1 keV),
but note also the greatly reduced radiation loss and the reduced particle influx(from H a ) in the divertor case. Second, Fig. 4 shows ion heating in TFTR by
neutral beams. Temperatures of 20 keV are reached, exceeding by a wide margin
the value necessary for fusion energy production.

5.

PLASMA-WALL INTERACTION AND IMPURITY CONTROLOne of the keys to obtaining high temperature, and indeed generally good
performance, in tokamaks is the production of pure plasmas, especially plasmas

486

GIBSONICRH IMPURITY BEHAVIOUR

0.0

0.5

1.0
Prt(MW)

FIG. J. Tfte wse o/ wall carbidization for impurity control in JIPP THUand ASDEX.

in which the heavy ion impurity content is sufficiently low. This has proved
to be especially important in discharges with ICH where antenna interactions(as demonstrated in JET) or power coupled to the wall (as in ASDEX) canrelease metal impurities. One solution is to surround the plasma with a low-Z
environment. One technique, first developed at Jiilich and demonstrated to beeffective in JET, is carbidization where glow discharge is used to deposit a hardcarbon layer on walls and antenna screens. Two examples of the effectivenessof the method are shown in Fig. 5. In JIPP TIIU this method is an essential
prelude to high power ICH operation, while the ASDEX results show that thetechnique leads to substantial gains even in a divertor tokamak.
An alternative technique of modifying the edge plasma behaviour has
been known theoretically for some years, but is now demonstrated for the firsttime by the group from the University of Texas. A helical perturbation is used
to produce an edge ergodic field layer, as shown in Fig. 6 (a). Measurements
show this to depress the edge temperature over a few centimetres and to steepen

487

IAEA-CN-47/L-I-1(a)
MAGNETIC FIELD STRUCTURE, m/n = 7/3: q(a) = 2.7-COIL
RADIUS

32

LIMITERRADIUSCENTRE
0

2it

7t

POLOIDAL ANGLEl h (POWER SUPPLY) = 4 kA: <lb r l>/B t « 0.1% AT 24 cm(b)Edge Eleclron Temperolure Profile
with 7 / 3 mode EMI•

200 --i

8o 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

•

•
•

150 -

o
o

100 -

'.
;
'.

O

50 i A A
•

21

.

i

22

.

23

24

1

i

•

25

26

27

•

A ;

1

28

29

rodius (cm)FIG. 6. Effect of an ergodic limiter on edge temperature profiles in TEXT.
(a) Computed magnetic configuration in the edge region, using the actual coil configuration.
(b) Measured edge temperature profiles for various currents in the helical coil.

the edge temperature gradient (see Fig. 6(b)). This may turn out to have implications both for impurity control and for production of the H-mode confinement
configuration.
Applied helical perturbations may also be used for the stabilization of MHD
modes, as was first reported by the PULSATOR group some years ago. Work
reported at the Conference by the groups from TOSCA in the United Kingdom,
HT6M in China and JFT-2 in Japan shows that the method is capable of further
development.

488

GIBSONrE versus power for various values of l p (JET)
0.8

0.6

0.4

0.2

0.0

5
P (MW)A 1 MA» 4MA
3 3MAX-point

10
Toial Power Input

< 2MA
0 5MA

o 3MA
2 2MAX-point® » H-Mode

FIG. 7. Degradation of energy replacement time with additional heating power in JET.

6.

CONFINEMENT

If the good news is that more heating does indeed give hotter plasmas, thebad news, until recently, has been that more heating has given worse confinement.
This has been so to such an extent that the fusion figure of merit, n D T E Ti
(product of deuteron density, energy replacement time and ion temperature),
had until the beginning of this Conference not been increased by additional
heating. I could choose data from almost any experiment to demonstrate thisdegradation, but it would be invidious of me not to use JET data for this purpose.
In Fig. 7 it can be seen that at each plasma current the energy replacement
time T E degrades with increasing power according to an offset linear law for theplasma energy. The incremental energy replacement time (corresponding to thevalue at large power) can reach only 0.2-0.3 s in JET compared to the best Ohmic
confinement times of about 0.8 s.
A way out of this dilemma was first found in ASDEX; it was shown thatfor divertor plasmas a second confinement mode - the 'H-mode' — could exist
in which the confinement time recovers to essentially the Ohmic value and does
not degrade with applied power. Experiments on Dili and later on PDX showedthat a similar behaviour could be produced in tokamaks without a divertor, but
including an internal separatrix.

489

IAEA-CN-47/L-M
200

O H-Mode§I"

I• L-Mode

~O

&

o

cPo

A H-Modewithstrong
elm's

: * \
A

£100

50 ;- i P =i.oKaye-Goldston L-Mode• Bt=2.10'_ Kappa=1.75

t • i

0

2

•

1

i

4
6
Total Power (MW)B

10

FIG. 8. H-mode behaviour in DIII-D.

Following in these footsteps, both separatrix and H-mode discharges havenow been produced in JET. For the first time in JET the np^gT; product has
exceeded the Ohmic value, reaching a value of 20 X 1019 m" 3 • s-keV. Someof the separatrix and H-mode points at 2 MA and 3 MA are indicated in Fig. 7.
The H-mode r g values for these discharges are better than those for similar
limiter discharges by more than a factor of two; this means that the H-mode in
JET appears to be worth about 2.5 MA in plasma current. Oddly enough, itturns out that in JET mechanical forces will limit the ultimate plasma current
capability for X-point configurations to about 2.5 MA below that for limiter
discharges!
In JET, it is not yet clear whether or not the H-mode points will be freefrom power degradation as was found to be the case in ASDEX. However, in
DIII-D, a new magnetic limiter plasma which has been in operation only for ashort time at GA Technologies, H-mode discharges have already been achieved
and in this case T£ does not appear to degrade with power (see Fig. 8).
At this Conference, the ASDEX group has given an account of the various
techniques which can allow a tokamak discharge to make the transition to
H-mode confinement. An increase in edge electron temperature, requiring athreshold in heating power, appears to be necessary. One of the techniques
which the ASDEX group advocates to achieve this is a reduction in edge recycling.
This may provide a link with the 'S-mode' high confinement discharges whichhave recently been produced in TFTR. The method used is to reduce the edgehydrogen recycling by preconditioning with helium discharges. So far, thisregime can only be produced over a limited range of plasma current with roughlybalanced beam injection. However, some extension of the range has beenobtained by ramping the plasma current and further improvements are planned.

490

GIBSON
0.20

0.15 -

0.10 -

0.05 -

10
P,NJ <MW)

FIG. 9. Improved confinement regime in TFTR.

TABLE III.

COMPARISON OF HIGH CONFINEMENT REGIMES

Experiment

I (MA)

R(m)(no radiation
correction)

Tu/Ip-R(normalized)(normalized)

ASDEX

0.08

0.3

1.7

2

2

JfT-2M

0.034
0.033

0.22
0.28

1.3
1.3

1.5
1.2

1.2
0.9

BIGD

0.17

1.0

1.7

1.3

1.3

JFTR

0.15

0.9

2.5

0.8

1.9

JET

0.5
0.6-0.9

2
3

3
3

1.1
0.8-1.3

1.5-2.2

IAEA-CN-47/L-I-1

491

In this supershot regime in TFTR the energy confinement time is restored
to the Ohmic level and again shows no obvious degradation with additional
heating power (see Fig. 9).
A further bridge between H-mode and S-mode discharges may be provided
by JFT-2M, which has obtained H-mode discharges in X-point configurations,but which also, remarkably, has produced H-mode discharges in a limiter configuration. As for the S-mode, the technique used by JFT-M for the H-mode isto restrict the edge recycling.
An important question to ask is whether, as has been speculated, theH-mode confinement time scales as the product of plasma current and dimension or whether it scales only as the current. Table III shows a comparison of
the five devices for which high confinement modes have been reported at thisConference. It will be seen that there is little to choose between the I p scaling
and the I p * R scaling, and we must await further results to resolve this tantalizing question.
An important difficulty with the best H-mode discharges in ASDEX is thatthe plasma suffers from impurity accumulation effects which finally destroy it.
At present, this effect does not seem to occur in JET, TFTR or DIII-D, but
further experiments are required.

7.

DENSITY AND BETA LIMITS

We have discussed temperature and confinement time - two of theparameters which determine the approach to fusion conditions. The thirdparameter is plasma density. The first constraint is that on the plasma pressure.
In order for the present generation of experiments (such as JET and TFTR) toreach fusion conditions and in order for future tokamak reactors to have achance to be economic, it is necessary that the plasma beta reaches the theoretical limit, established computationally by F. Troyon. The results from TFTR
in Fig. 10 show that this limit can indeed be approached in the low-q regime,
which is the one of interest for fusion relevant tokamaks.
In addition to the beta limit, realistic parameters for an approach to fusion
conditions require central densities in excess of 10 20 m" 3 , or mean densities inexcess of 7 X 1019 m~ 3 . The operational range of tokamaks is restricted by adensity limit above which edge impurity radiation initiates a shrinking of the profile
leading to instability and ultimately to disruption. For high field devices (suchas ALCATOR and the Frascati tokamak) this limit is high enough to permit
reactor relevant densities, but for lower field devices there has been concern
that, with the achievable plasma purity, the density limit would be too low.
At this Conference we have seen that a number of low and intermediate fieldtokamaks (ASDEX, JT-60, TFTR, JET) can routinely produce values of the
figure of merit of nR/B> 6 X l C n T ^ T " 1 . At this level the necessary densities

492

GIBSONOS

10FIG. 10. Approach to the theoretically predicted beta limit in TFTR.

Maximum Attained Densities

0.6

OH
Pellets• Carbldized
2.2T(ASDEX)

0.4

0.2

2

4

6

10

nR/B T

FIG. 11. Density limits in a number of tokamaks.

can be obtained. Some operating diagrams are shown in Fig. 11. The higher
values of nR/B are obtained with the help of pellet injection, or neutral beam
heating, or both.
Once densities above those possible with purely Ohmic discharges havebeen reached, however, some method of controlled density decrease is requiredin order to terminate the discharge. Success has so far been obtained by profilingthe power input reduction and by introducing some in-vessel pumping means,such as, for instance, the suitably conditioned carbon tiles on the wall of JET.

IAEA-CN-47/L-M
8.

493

PELLETSThe use of injected solid pellets of deuterium ice as a fuelling method fortokamaks has facilitated high density operation and has led to considerable
performance improvements in many experiments. Pellet injection can also
apparently lead to changes in the basic transport processes, as was reported ina paper by the ALCATOR group. Comparisons of the observed ion thermal
conduction loss with neoclassical theory are always imprecise because of theexperimental difficulty of establishing the various components of the ion
energy balance. The ALCATOR C experiment, however, appears to show aclear transition from a regime with an anomalous ion thermal loss of about
2.5 to 6 times the neoclassical value to a regime with a level that appears not
to exceed the neoclassical value (see Fig. 12). The transition is induced byinjection of a pellet which doubles the plasma density and steepens the density
gradient. Improved particle confinement and neoclassical impurity concentration are said to occur concurrently. This appears to be a regime worthy offurther elucidation.

9.

PROSPECTS

Following the discussion of each of the three components of the fusion
figure of merit, these components are presented on a diagram so as to assess the
progress towards achievement of ignition conditions with tokamak systems andthe prospects for the future.
Figure 13 is a plot of the product nrjTjrg versus Tj, the density and temperature being central values while the energy replacement time is an overall value.
Two computed curves are shown. The curve marked 'ignition' corresponds todominant alpha particle heating (taken as P a /Pi oss = 74%, or fusion power divided
by power input, Q = 14; with this condition the plasma temperature would
continue to escalate owing to the fusion power, even if the externally applied
heating power were to be progressively reduced). The other curve corresponds
to substantial alpha particle heating (taken as P a /Pi o s s = 14% or Q = 0.8) 1 .
The value of npT-Tg on these curves is approximately constant in the importantregion of its minimum near Tj = 20 keV and so forms a figure of merit. Also
shown on the diagram (in the box) are the best values so far obtained and1

The computed curves are for a plasma with 50% of its ions as tritium and 50% as
deuterium. The curves are for specific profiles, corresponding to n/n 0 and T/To = (1 —p2)a,where p is the normalized radius and a n + ar = 2, the subscripts denoting the indices fordensity and temperature. The values are insensitive to the profiles as long as broadly similar
shapes are considered; thus, taking a n + aT =0.5 reduces the 'ignition' curve minimum from
4X 1021 m"3-s-keV to 3 X 10 2 1 nr 3 .s-keV.

494

GIBSONIon Transport Anomaly

0.45

FIG. 12. Neoclassical transport behaviour in ALCA TOR C.

Progress Towards Ignition

10=
3MA / . JET
rE=0.6s
10"

TFTRt,=6keVn o =5.5x10 l9 m- :10

ALC. A . 1 9 8 3 »TFTR
1 9 ALC.A*.TFR'ORMAK"
10"

c

7.

TFR
•ORMAK

•T-3
10

100

t, (keV)• Ohmic heating only
• Ohmic+additional
heating
A 'hot ion mode'a Ohmic heating + pellet
A H-mode X-point (JET)« X-point (JET)

FIG. 13. The fusion figure of merit, n^T^^, versus central ion temperature Tv showing
progress towards ignition conditions with tokamak systems.

IAEA-CN-47/L-I-1

495

reported at this Conference. In addition, the best values of nDTjTE reported atearlier times are indicated for some representative experiments. It will be seen
that, in round terms, the figure of merit has increased by a factor of 10 over the
past three years, by a factor of 100 over the past ten years, and by a factor of1000 over the past twenty years. The currently obtained best value of2 X 1020 m~3-s-keV needs to be increased by about twenty times to reach thelevel of'ignition', but only by about 3.5 times to reach the level of substantialalpha particle heating. In a particular case the substantial alpha particle heating
curve would be reached by a modest increase of the parameters currently observed
in JET, i.e. by increasing n D from 0.55 X 1O20 m~3 to 1 X 1020 m~3, T; from
6 keV to 12 keV and r E from 0.6 s to 0.8 s.
In conclusion, it can be said that now there seems to be every reason to
expect that today's tritium compatible tokamaks (JET and TFTR) will reachthe regime of substantial alpha particle heating within the next few years.

IAEA-CN-47/L-I-2

SUMMARY ON
ALTERNATIVE MAGNETIC SYSTEMS —EXPERIMENTAL RESULTS*
T. SEKIGUCHI
University of Tokyo,
Hongo, Bunkyo, Tokyo,
Japan

1.

INTRODUCTION

About sixty papers presented at the Conference deal with alternative
magnetic systems (Sessions C, D and I). This category includes a wide varietyof plasma systems (see Fig. 1). The three areas: helical systems (17 papers),
open systems (14 papers) and reversed field pinches (RFPs, 13 papers) are
discussed most widely, in more than 70% of the papers dealing with this topic.
In particular, the RFP has received increasing attention in the last few years.

2.

HELICAL MAGNETIC SYSTEMS

Not many new data have been presented at the Conference, which reflects
the present situation. Table I summarizes the simultaneously obtained highest
values of plasma parameters for various machines.
2.1. Extension of plasma parameters(a)Powerful electron cyclotron heating (ECH) in the millimetre wavelength
range (50-75 GHz, 400-700 kW) has been introduced in Heliotron-E (D-I-l)
and in W VII-A (D-I-2, D-V-l); this permits the generation of target plasmas
with 1-2 keV electron temperature.
(b) The technique of carbon coating of the vacuum inner wall (carbonization)is very effective in reducing radiative losses due to high-Z impurities (by afactor of two to three); this facilitates both the production and the longtime sustainment of plasmas with an average density in excess of 1014 cm" 3 .

This summary talk was also published in Nucl. Fusion 27 (1987) 490.

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(c)

An ion temperature in excess of 1 keV was obtained in both Heliotron-E(D-III-2) and Uragan-III (D-III-5) by radiofrequency (RF) heating.
Uragan-III employs ion cyclotron heating (ICH) plus slow Alfven wave
heating. The first systematic ICH at a power level of 1 MW in both thefast ion Bernstein mode and the slow wave mode was applied in Heliotron-E.
The experimental results as well as those from simulation studies (D-III-1)
support the belief that fast wave heating is an attractive heating method forhelical systems.

2.2. Energy confinement time (r E ) scaling
The Heliotron-E group (D-I-l) proposes the following global empiricalscaling for the energy confinement time:
TE oc [n/P(NBI)] O.S-0.6
where n denotes the plasma density and P(NBI) denotes the injected neutralbeam power. This scaling seems to be similar to that of tokamaks (this should
be confirmed further).

500

SEKIGUCHI

2.3. Beta limit

The average beta value of about 2% previously attained in Heliotron-E
remains nearly the same in the 'stable Q-mode' which is characterized by a flatradial pressure profile. However, the present maximum beta value appears to be
restricted by 'resistive' interchange modes (occurring in the outer plasma region)
rather than by 'ideal' MHD modes (D-I-l).
Theoretical studies of possible improvement of the beta limit, hopefully
to more than at least 5%, are in progress (D-I-3, D-III-6, D-V-2, D-V-3). Thesestudies include: (a) 3-D magnetic axis systems, represented by Asperator andHeliac, where the magnetic axis itself forms a helix, (b) the Advanced Toroidal
Facility (ATF) at Oak Ridge which aims at an optimum combination of magnetic
well and shear, anticipating an average beta value of 8%, and (c) W VII-AS at
Garching, where a module coil system is employed and the Pfirsch-Schluter
current is minimized.
In Canberra, Australia, a small scale Heliac has started operation and the
H-l Heliac is under construction (D-III-6). Also, the ATF and W VII-AS, the
TJ-II Heliac (D-V-4) and Helias (D-I-3) are all in the construction or planningstages, and it will be at least another two to three years before the results ofexperiments in high beta helical systems are made public.

2.4. Radial transport

The origin of an induced radial electric field and its effect on radial transport
have been investigated by the W VII-A group (D-I-2). It is found that in neutralbeam injected (NBI) plasmas the loss of perpendicularly injected fast ions inducesa radial E-field, which in turn reduces the neoclassical ion thermal loss. On theother hand, in ECH plasmas the neoclassical ambipolar particle diffusion induces
an electric field, and in the outer plasma region electron thermal conduction is
anomalous whereas in the plasma centre it is neoclassical. The 'ripple diffusion',
which includes the effects of a radial electric field, has been studied in Heliotron,
but so far no clear results have been obtained. Recently, an experiment on'pump limiters' has been performed in Heliotron-E (one of the US-Japan
co-operative efforts, D-I-l); the findings indicate the possibility of active controlof high heat flux and particles with a combination of an ergodic magnetic field
and a pump limiter.
Transport in helical systems in connection with the induced radial electricfield and with impurity control is a very important subject of investigation. The
experimental verification of the numerous theoretical proposals is a central
problem.

IAEA-CN-47/L-1-2

501

2.5. Impurity control

The usefulness of carbon coating of the vacuum inner wall has already been
mentioned. Both the Wendelstein group and the Heliotron group found thatduring operation with ECH the impurity accumulation was reduced and overall
confinement somewhat deteriorated (D-I-l, D-V-l). In future machines it may
be necessary to provide a poloidal divertor.
2.6. Self-induced plasma current
In connection with helical systems, reference is often made to 'currentless'plasma. Actually, in present-day machines, self-generated low level toroidal
current is observed. It is necessary to investigate the physical origin of thisphenomenon (bootstrap, Pfirsch-Schliiter current, diamagnetic current, etc.) andwhether or not it is harmful for future upgraded machines.

3.

OPEN-ENDED SYSTEMS

3.1. Tandem mirror
In the category of open systems, the most intensive efforts are being devoted
to the tandem mirror. However, no new major machine has been put into operation in 1985-1986, except for the modified Phaedrus-B at Wisconsin (C-I-4), anddetailed studies have continued on existing machines. The main plasma parameters
achieved in the central cell of tandem mirrors are summarized in Table II.
(a) Thermal barrier. The axial electrical potentials achieved in mirror
machines are listed in Table HI. The potential depth of the thermal barrier (0 b )is found to be strongly governed by the level of selective ECH; both TMX-U (C-I-l)
and Gamma-10 (C-I-2) achieved 0 b values of nearly 1 kV. The recent Gamma-10
results also showed that the axial ion plug potential (0C) increases with increasing
barrier depth (0 b ) and attains 1.2 kV, and that for long-time sustainment of thethermal barrier it is necessary to conserve an appropriate spatial ion density
profile, the energy filling rate for which is found to be classical.
(b) Axial confinement scaling in the central cell. An axial ion confinement
time in excess of 0.1 s has been attained in TMX-U and Gamma-10 (0.4 s in thelatter) in the ion temperature range of 1-2 keV. TMX-U also attained an average
central-cell beta value of 6%. A further result of even greater importance is theconfirmation by experiments in Gamma-10 (C-I-2) that the axial ion confinement
scaling is consistent with the Pastukhov formula which is related to the ion plugpotential (0 C ).

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1AEA-CN-47/L-I-2

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SEKIGUCHI(c) Non-ambipolar and ambipolar radial ion loss. Non-ambipolar loss is
attributed to plasma rotation due to the plasma potential with respect to thecontainer walls. Both TMX-U and Gamma-10 demonstrated the controllabilityof the loss time by end-wall potential adjustment (consistent with resonant
neoclassical theory), in the range of 10 ms to more than 1 s. In addition, theambipolar radial ion confinement time was found to be nearly equal to or
longer than the axial confinement time. Thus, an overall central-cell confinement
time of more than 0.1 s was estimated for Gamma-10 under the assumption thatthe currently existing relatively large charge exchange losses are absent.
3.2. RF-plugged cusp anchored mirror
For the RFC-XX-M (IPP, Nagoya) very interesting results have been reported(C-II-6): (a) A plasma pressure in the small-volume end-cusps about an order ofmagnitude lower than that in the central mirror cell is capable of sufficiently
stabilizing the MHD modes in the central cell, (b) An empirical scaling of the
RF plug (ponderomotive) potential has been established and a maximum plugpotential of nearly 1 kV obtained, (c) A new controllable operating mode(the 'negative potential mode') has been discovered; this is essentially a tandem
mirror sustained only by ICH for at least 150 ms, with a thermal barrier likenegative potential (as high as 500 V) between central cell and cusp end-cell.
The possible application of this unique cusp anchor concept to large axisymmetric tandem mirrors should be investigated.
3.3. RF utilization in open systems
In recent years, the utilization of high power RF techniques has been
extended to open systems, particularly to axisymmetric mirrors. Besides
conventional plasma heating by slow waves and ponderomotive end-cellplugging (RFC, Nagoya: C-H-6; TARA, MIT: C-I-3; Phaedrus-B, Wisconsin:
C-I-4-1), this includes plasma heating by fast waves, stabilization of MHD modes
and drift modes, thermal barrier sustainment ('pumping'), and startup andsustainment of mirror plasmas by ICH only.
MHD stabilization in axisymmetric mirrors by ICH was first observed by
the HIEI group (C-I-4-2), and then in Phaedrus (C-I-4-1) and TARA (C-I-3) which
also has a central-cell divertor. These groups found that the RF ponderomotive
force is responsible for MHD stabilization. The scaling of this stabilization has
been extended to ion temperatures in excess of 100 eV; plasma startup andlong-time plasma sustainment by ICH have been demonstrated in HIEI. It has
also been shown that the sustainment of a high beta (~13%) plasma (Phaedrus)
as well as the formation of a thermal barrier (Phaedrus-B) by RF only are
possible in axisymmetric mirror configurations.

IAEA-CN-47/L-I-2
4.

505

REVERSED FIELD PINCHES (RFP)In 1985 and 1986, HBTX-1A (Culham) was upgraded to HBTX-1B.
TPE-1RM-15 (ETL, Japan) was put in operation in 1986. Both machines use
a vertical field to centre the plasma within the vessel. Although the experimental
ranges of toroidal current and electron temperature have remained nearly the same
since 1984, at this Conference the papers presented contained much more elaborate
data which were obtained over wide ranges of parameters and machines. The
simultaneously obtained plasma parameters for the main devices are listed inTable IV.
4.1. Scaling of plasma parameters
The plasma temperature of RFPs was empirically found to be proportional
to the plasma current. However, recent HBTX-1B experiments (D-II-1) indicate
that, while the poloidal beta is high at low current, it decreases with increasing
current (from 20% at 80 kA to 10% at 220 kA in HBTX-1B). In ETA BETA-II(D-II-5), TPE-1 RM-15 (ETL, Japan) and ZT-40M (D-II-3), the poloidal beta
value is kept nearly constant (~10%) up to 150-200 kA. This constancy of beta
is shown by varying the ratio I/N (where I is the current and N is the number ofparticles per unit length of the plasma column), by gas puffing and by multiplepellet injection. ETA BETA-II, ZT-40M and STP-3M (IPP, Nagoya: D-II-2) all
report an increase of Ze^ in the outer regions of the plasma with increasing
current.
4.2. Relaxation process (dynamo effect)
In almost all machines, investigations of long-time self-sustainment of the
RFP magnetic configuration have continued in relation to the observed MHD
fluctuations and anomalous ion heating. The REPUTE-1 (D-II-4), ETA BETA
and HBTX groups all have observed periodic 'current peaking' and 'flattening'
due to the internal kink mode or the resistive tearing mode. It has been shown
on HBTX that the relaxation process extends into the 'scrape-off layer, causing
an additional and sometimes large contribution to the loop voltage. A numberof technical improvements (reduced field error, accurate plasma centring by avertical field, etc.) led to a substantial increase in the energy confinement time(a factor of five).
4.3. Pulse length and energy confinement time
Pulses of long duration, which are possible because of the relaxation process,have been sustained for 36 ms in ZT-40M at low current (60 kA); they are
restricted by the shortage of volt-seconds, not by physics. A slow current ramp-up

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IAEA-CN-47/L-I-2

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with a maximum risetime of 15 ms and an energy confinement time close to 1 ms
have been realized in this machine. It appears that through these improvements
controlled 'current termination' (similar to 'current disruption' in tokamaks) is
now possible. Also, equilibrium control by vertical fields has recently beensuccessfully performed on HBTX and TPE-1RM-15. A pulse length of morethan 10 ms has been observed in OHTE (GA Technologies: D-V-7) with a thinshell having a resistive skin time of 1.5 ms. This is an interesting finding, but
in comparison with the results from REPUTE I there are certain discrepancies
and further clarification seems necessary. The impurity problems are not serious,
except at high plasma densities as in ETA BETA II; however, it may be necessary
to provide divertors in upgraded machines with larger plasma currents (of theorder of mega-amperes).
4.4. Oscillating field current drive
The 'F-0 pumping technique' is one means of increasing the pulse length.
For the first time, audiofrequency current drive has been applied in ZT-40M.
Preliminary results indicate that the plasma current is actually damped when
the phase difference between the toroidal and poloidal RF fields is unfavourable.
There were some indications of a net increase of plasma current under
optimal conditions. At high power operation, however, MHD activity appears
to give rise to impurity production from the wall.

5.

COMPACT TORIThis class is conventionally divided into two groups: spheromaks and fieldreversed configurations (FRCs). Spheromaks have both toroidal and poloidal
magnetic flux, whereas FRCs are considered to have poloidal flux only.
5.1. Spheromaks
The plasma parameters achieved in spheromaks are listed in Table V. At
present, there are two methods by which this magnetic field configuration canbe realized. In the first method the plasma equilibrium is maintained by an
externally applied magnetic field (machines listed in the lower part of the table).
In the other method a magnetized plasma gun type injector and a conducting
wall plasma container ('flux conserver') are used in combination to maintainthe resulting spheromak configuration (machines listed in the upper part of thetable). Only for the flux conserver type have results been presented at theConference.

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The CTX spheromak (D-IV-1-1) has produced significant results:(a) Taylor's minimum energy principle has been quantitatively confirmedin spheromaks, and (b) the global magnetic helicity can be conserved, not only
during the buildup phase but also during the sustainment phase by means of'helicity injection', which implies a staged operation of the power supply on a
coaxial gun type plasma injector. This process led to a toroidal current in excess
of 1 MA and a configuration lifetime of several milliseconds. This is very
interesting because it indicates that nearly steady operation may become
possible in spheromaks with DC power supply.
The largest problem in this system at present, common to all spheromaks
including CTX and CTCC-1 (D-IV-1-2), is a serious energy loss of unknown
physical origin. In CTX, the size scaling of the confinement time has beenstudied by increasing the flux conserver radius by a factor of nearly two.
While the particle confinement time increased, no significant increase in theelectron temperature (below 100 eV) or in the energy confinement time wasapparent. This suggests that strong energy loss mechanisms must be present.
There are speculations regarding the physical origin of this energy loss, butmore convincing identification of these processes and the development of acontrol method are urgently required.

5.2. Field reversed configurations (FRCs)
The plasma parameters achieved in FRCs are listed in Table VI. Most of
the present FRCs are based on theta pinch techniques, in which case the FRCs
usually have a very elongated plasma cross-section. A new technique of forminga nearly spherical FRC (by combining laser produced plasma and field reconnection by fast rising coil currents) has recently been demonstrated at the Universityof Tokyo.
Improvement of the stabilization of the n = 2 rotational instability bysuperimposing a helical multipole field has been pursued further (D-IV-2-2,
D-V-9) and the Osaka group presented a unified theoretical interpretation of
the processes involved. The slow formation of FRC plasmas with flux supplementby an additional longitudinal current was demonstrated at Nihon University(D-IV-2-2).
Empirical confinement and transport scaling has been studied by two groups
in the USA (D-IV-2-1, D-V-9). One group (D-IV-2-1) performed experiments inslow risetime machines with widely varying machine dimensions (four machines)
and operating conditions, and investigated the particle and energy lifetime
scaling with the plasma (separatrix) radius. This group also proposed for the
first time that the transport in FRCs may be explained by the low frequency
drift wave theory.

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6.

SEKIGUCHI

OTHER MACHINES

Table VII gives typical data for various devices. These are discussed in thefollowing sections.

6.1. Screw pinch
The TPE-2 screw pinch device (D-IV-3) provided very interesting data; thedischarges at a low qa-value (about 1.2) with a flat radial current profile produced
an average total beta value of nearly 10%, a global energy confinement time of afew milliseconds at a maximum average temperature of 400 eV, and a plasma
density of 1014 to 1015 cm" 3 . Of particular interest is the fact that the maximum
total beta value of 10% is beyond the 'Troyon limit'. The confinement scaling isvery close to neoclassical. Thus, although a number of problems remain to besolved, further improvement of confinement and higher plasma beta can be
expected.

6.2. REB ring spherator
The possible utilization of an REB ring for magnetic fusion was firstinvestigated in Astron at Livermore some 30 years ago. SPAC-VII (D-V-10) is
a toroidal version which is used in combination with the spherator concept.
A quiescent intensive REB ring, with megaelectronvolt order of energy and a
total current up to 30 kA, could be sustained for longer than 15 ms with very
low poloidal field fluctuations. With the REB ring active, a plasma with
3.7 MHz/100 kW RF power was produced, but only a plasma of low density(5 X 1012 cm" 3 ) and low electron temperature (nearly 10 eV) could be confinedwith a global electron confinement time of about 1 ms.

6.3. Bumpy torus
Because of the discontinuation of the EBT series in the USA, only resultsfrom the Nagoya Bumpy Torus (NBT) (D-III-4) were presented at the Conference.
The main results are as follows: (a) The overall transport can be explained byneoclassical theory, and the energy confinement time (nearly 1 ms) is longer
than the particle collision time by at most a factor of ten (several times smaller
than expected), (b) The mechanism of the self-induced radial electric field aswell as its control and its effect on confinement have been clarified, (c) Whenthe input microwave power is increased, the beta value of the hot electron ringitself remains of the order of several per cent (rather than several ten per cent,IAEA-CN-47/L-I-2

513

as initially anticipated). An average minimum-B configuration has not been
produced, although some stabilization effects have been observed in the vicinityof the ring. Thus, although it could not be proven that the NBT is a valid concept
for the development of a fusion reactor, it may be worth while to consider atransfer of this concept to other magnetic systems, which should be possible if
a hot electron ring with a very much higher beta could be realized.
6.4. Z-pinch and plasma focus
The Imperial College, London (D-IV-4) has performed detailed experimental
and theoretical investigations of operational mechanisms for three different
types of small linear Z-pinches with very high particle densities (10 18 " 19 cm" 3 ;even ~10 2 1 cm" 3 at 'hot spots', which is very near to liquid density). The RoyalInstitute of Technology, Stockholm (D-V-6), reported results for an octopole
stabilized toroidal Z-pinch with much lower density.
Regarding plasma focus experiments (D-IV-5-1, D-IV-5-2), no data for
improved D-D neutron yield (in excess of 1 0 n neutrons per pulse) were reported
at the Conference. Instead, more detailed physical studies were presented, suchas the correlation of generated X-rays, neutrons and high energy deuterons withpinch dynamics and with ion acceleration mechanisms.

7.

CONCLUSIONS

During 1985 and 1986, no new major or upgraded machine in the categoryof alternative magnetic systems has become operational. However, it appears
that helical and open-ended systems as well as RFPs, compact tori and otherexperiments all made steady progress and are being thoroughly explored.
In particular, several machines in the category of helical systems havereached ion temperatures of the order of kiloelectronvolts, and elaborate studies
on various items have been performed in the plasma parameter region next tothat of the tokamak. Regarding the plasma parameters, a new generation oflarger machines appears to be necessary in order to make great strides towards
the fusion target, as was also the case with the tokamak nearly one decade ago.
However, there are still numerous basic physics problems which have to besolved, particularly for transport and beta limit improvement. Thus, it seems
essential to show how to obtain compatibility of the more refined experimental
results needed with the designs of the next-generation larger helical machines
such as that being planned by the group of Japanese universities.
Good progress has also been made in tandem mirrors in the understandingof the physics problems involved and in the extension of the plasma parameters.
In Gamma-10 (Tsukuba, Japan) a potential well very near to 1 kV was attained

514

SEKIGUCHI

at the thermal barrier, with an ion plug potential of more than 1 kV. This led toa perpendicular ion temperature of 1.7 keV, an average beta in excess of 1% andan estimated energy confinement time of more than 0.1 s (assuming goodvacuum), all in the central cell. However, the nr value is still of the order of1011 cm~3-s because of the low plasma density in the central cell (<10 13 cm" 3 ).
The critical issue is to find suitable general schemes for substantially improving
the overall energy confinement. One promising scheme is 'axisymmetrization',
for which an increasing number of results proving its effectiveness have beenreported. However, further investigations are necessary.
The repertoire of RFPs has become much larger and steady progress has
been achieved during 1985-1986. Next-generation machines with currentsof 2-4 MA are now being constructed or planned. These are considered toconstitute milestones on the way to the final goal of 20 MA reactors. However,more detailed work is still needed to solve numerous problems, including theestablishment of effective and practical ways of current drive.
Spheromaks and FRCs have also made good progress in gaining extensive andintensive physics information, but they are still in the exploratory stage. Spheromaks
are currently suffering from heavy energy loss. So far, the magnetic field
generated in spheromaks was at most only 0.6 T. A new machine with amagnetic field in excess of 2 T has been put into operation recently (MS, Univ.
of Maryland, USA); for this, a substantial increase in plasma temperature to
above 100 eV has been anticipated. A larger FRC machine is under constructionin the USA (LSX, see Table VI); this is expected to supply additional information on transport and on MHD (tilting) stability in a more reactor relevant
parameter range. In view of the future potential of compact tori, the continuation of studies of these unique systems is highly desirable, and comparisons
between the two spheromak types FRC and RFP will become necessary in the
not too distant future.
It should be pointed out that it is of utmost importance to show within ashort time that the peaceful use of fusion energy can indeed be realized, at least
scientifically and technologically. Furthermore, since fusion core plasma physics
and reactor technology problems for different magnetic systems are closelyinterrelated, it would be advantageous to use the currently most advanced
tokamaks for the study of problems in other systems; it is hoped that this canbe done at reasonable cost. To reach the ultimate goal, the fusion community
will also have to overcome the hurdles of economic feasibility and social
acceptance. Therefore, it is necessary to provide at least two or three alternatives,
even under the currently stringent resource limitations. It is essential to achievea carefully considered balance between tokamaks and alternative magnetic
systems. Furthermore, to promote development and to enable fast andsubstantial progress to be made, better co-ordinated international planning andexecution of the work is of ever greater importance, especially for the selected
alternative magnetic systems.

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515

ACKNOWLEDGEMENTS
The author is grateful to many colleagues from Japan and other countries
for their great assistance in preparing this summary. Also, he wishes to thankDr. A.A. Newton from the UKAEA Culham Laboratory for improving his
manuscript.
Because of the very limited time available for the preparation of thissummary it was unfortunately not possible to include all work in this field.

IAEA-CN-47/L-I-3

SUMMARY ON
MAGNETIC CONFINEMENT THEORY*
K. LACKNER
Euratom-IPP Association,
Max-Planck-Institut fur Plasmaphysik,
Garching, Federal Republic of Germany

1.

INTRODUCTION

Approximately 40 predominantly theoretical papers on magnetic confinement were presented at the Conference. The following main points werediscussed:- Derivation of the reduced resistive MHD equations in the intermediate andlong mean free path regimes;- Sawtooth modelling;- Effect of fast particles on MHD instabilities;- Enhancement of the critical beta in tokamaks and stellarators.
In these fields, the papers indicated that milestones have been reached. Progress
was reported also in other areas, but of a less definitive nature. Unfortunately,in the area of currently largest concern — the understanding of tokamak confinement — no decisive progress has been reported.

2.

TRANSPORT IN TOKAMAKS

2.1. Energy confinement: Comparison between experiment and theory
The present status of our understanding of tokamak confinement is best
illustrated by three papers which compare the experimental results of Ohmic
and L-regime discharges with the predictions of the currently most advanced
theories of localized anomalous transport, namely those for trapped electronmodes (TEM) in the different collisionality regimes and for the ^-modes'(r?i = d In T;/d In n e ). The essential features of these comparisons and theconclusions of the authors can be summarized as follows:
Paper A-VI-1-1 deals with purely ohmically heated and ICR heated JET
experiments. The experimental electron and ion energy fluxes as a function
This summary talk was also published in Nucl. Fusion 27 (1987) 500.

517

518

LACKNERof radius in the plasma core region are compared with theoretical predictions
for trapped electron modes, rj; modes and microtearing modes. The theoreticalpredictions are found to exceed locally the actual fluxes by more than one
order of magnitude and in particular to show a completely different radial
dependence.
Paper A-VI-1-2 describes simulations of TFTR discharges at different
heating powers and in different plasma regimes, including cases with pellet
injection. In these simulations, the global entropy production through heatdiffusion was derived from first-principle TEM and T?J models, but the predictions for local power flow were modified by a factor constraining the Te
profiles to reproduce the experimentally observed shapes. The reported
agreement (of global confinement time and, evidently, also Te profiles) thusachieved is very good.
Paper A-VI-1-3 describes simulations (of D-III discharges) with TEM andfy-mode transport predictions, but without additional profile adjustments
except through a sawtooth model. This procedure gives reasonable agreement
regarding the electron density and power scaling of energy confinement times,
but it predicts r E « Bt rather than Tg « I p as experimentally observed. Thisis considered to be due to the lack of an adequate transport model for theplasma edge layers which should in particular enhance the heat conduction in
the high-q regions and thus produce a shrinking of the effective discharge radius
for cases of large edge safety factor q a .
The conclusions of the three papers can be summarized as follows: The
papers agree on the non-existence of an adequate transport model for the edge
(say q > 2) region. It can be argued that the transport results of the different
drift wave models are in surprising agreement with the space-averaged transport
in the core region, in particular considering the fact that they are true firstprinciple theories, without 'fudge' factors. It is, however, also clear that they
themselves do not correctly describe the spatial distribution of transport - afact which is obviously made evident much more in a comparison of fluxes(as given in paper A-VI-1-1) than in a comparison of simulated and measured
TE values. The discrepancies observed are intrinsically linked with the resilienceof the plasma temperature profiles to respond to changes in the power deposition profiles - a behaviour commonly called 'profile consistency'. Different
explanations for this are possible. In the discussions during the Conference,in particular regarding the experimental results, it was not possible to come toa decision as to which of the following alternatives is correct:— The action of sawteeth at the centre and an as yet unspecified edge transportmechanism suffice to explain this profile consistency (A-VI-1-3).
— An additional mechanism influences local transport, enforcing the profile
shape but only little affecting its global scaling (A-VI-1-2).
— The agreement with regard to the global transport is coincidental andmechanisms other than drift waves are actually responsible for the observed

IAEA-CN-47/L-I-3

519

energy confinement. In particular, beyond some threshold, the heat
conductivity may just be a steep function of Te gradients, as predicted(for instance in paper A-VI-4) for an rje-driven instability. This in particularwould then explain the confinement degradation with heating power andwould also yield a ('weak') form of profile consistency.
Transport simulations with a quite different aim were presented in paper
E-III-5, in an investigation of the combined effect of thermal and fast particle
energy losses on tokamak performance. The distribution of non-thermal
particles has been studied by combining a conventional transport code with a
Fokker-Planck description (including a spatial loss term). Other situations
studied with this model are the slide-away regime of Ohmic heating and ICR
minority heating, where ripple transport could, for instance, lead to nonthermal particle loss.
2.2. Models of anomalous transport
New first-principle models of anomalous electron energy transport presented
at the Conference focused on the consequences of magnetic braiding due to theinteraction of islands of 'intermediate' m, n number. Large-scale numericalsimulations of the diffusion of magnetic field lines in a model perturbation field
are presented in paper E-III-4. The results of this study should also permit acomparison with the usual expressions for the magnetic field line diffusion suchas those given in Ref. [ 1 ]. Transport models of this type tend to give a strongly
unfavourable temperature dependence of confinement. However, for thesemodels a mechanism is needed which produces the field perturbation. The
authors of paper E-III-4 suggest that such a mechanism could be due to resistivity
differences between the interior region of the islands and the surrounding ergodized zones arising from the mechanism of the heat transport itself. An alternative mechanism for the production of the islands could be due to tearing modes,
which for the considered range of m, n numbers are usually stable within
resistive MHD but could be destabilized by neoclassical effects (bootstrap
current). Such effects are discussed in papers E-II-3-1 and E-II-3-2, presenting
the more relevant long mean free path models. A further possibility is outlined
in paper E-I-4-3, where a non-linear model for ballooning tearing modes is shownto lead to an unconditional instability.
Other papers dealing with local anomalous transport models considered inparticular the plasma edge region where rippling modes may occur. Rogisterand Hasselberg (E-III-1) pointed out that this mode should be further destabilized
in toroidal geometry and should give rise to top-bottom asymmetries in the
resulting fluctuations, as reported from TEXT. The consequences of the temperature and shear dependence of rippling mode produced turbulence are discussed
in paper A-VI-1-3. In particular, it is suggested that this dependence is responsible for the H/L mode bifurcation in divertor tokamaks since it could give rise520

LACKNERto a bistable thermal equilibrium in the high-shear edge regions, with a hightemperature/low x e branch and a low temperature/high x e branch.
Enhanced transport is discussed also in paper I-I-4 which presents a theoryof anomalous momentum transfer. Morozov and Pogutse (I-I-4) also compare
their results and the predictions of the neoclassical, gyroviscous model of Ref. [2]with the experimental findings of beam heated discharges; however, no conclusion
was reached.
Another theoretical approach to the phenomenon of profile consistency,
at present applied by several authors, is to show that the observed plasma
profiles correspond to extremal states of energy or entropy under assumed
constraints on accessibility. Such principles are formulated in paper I-I-5 andapplied in a prediction of j(r) (and thus, in resistive equilibrium, T(r)) andn(T) relations.
Anomalous energy transport in reverse field pinches is analysed in paper
D-V-8. In the inner region of a tokamak, transport is dominated by magnetic
fluctuations, driven by m = 1 tearing modes, but in the outer zone a combination of resistive interchange and ion temperature gradient driven modes isfound to be responsible for confinement. This leads the authors to suggest
pellet injection as a means of controlling density profiles and thereby quenching
this transport enhancement.

3.

MHD STABILITY OF TOKAMAKS

3.1. Simulation of experimentally observed MHD behaviour
A direct comparison of experimental results with MHD stability is attemptedin the following three papers.
Hender (A-V-3) analysed the tearing mode activity in JET with the threedimensional fully resistive MHD code FAR. The coincidence between measured(by Bg loops) and simulated mode patterns is impressive and is better than that
discussed in papers A-VI-2 and A-VI-3 where only the reduced MHD equations
are used.
Chu (A-VI-3) and Gruber (A-VI-2) analysed the approach to the beta limit
on Doublet III and ASDEX, using basically the same theoretical tools, namely
a simulation of the gross MHD behaviour with the CART code, solving the
reduced resistive MHD equations, and an analysis of the ballooning mode stabilityof the experimental profiles. However, a quite different experimental approach
is followed in the two devices. The beta limit experiments on Doublet HI
proceeded in the L-regime and were accompanied by a strong gas puff withconsequent edge cooling and strong drive of the sawtooth fluctuations. The
ASDEX studies were carried out in sawtooth-free hydrogen discharges. In
Doublet III, the peak </3t> values were always realized just before a disruption;

IAEA-CN-47/L-I-3

521

they were preceded by a large sawtooth which destabilized the m = 2 tearing
mode. The ASDEX beta limit studies, on the other hand, did not generally
lead to disruptions but showed a time behaviour of the energy content which
was found to track that of the ideal ballooning mode limit. However, since theactually achieved <|3t> values remained about 20% below this theoretical prediction, the authors suggested that the time behaviour of the energy is caused byresistive ballooning modes.
3.2. Long mean free path effects on resistive MHD stability
Practically all resistive MHD stability studies are based on a fluid model
which is actually not applicable under the long mean free path conditions ofmodern-day tokamaks. One way to apply corrections is to combine the usual
A' analysis of tearing modes with the application of the drift kinetic equation
for the description of the processes in the resistive layer (see paper E-III-1).
A more general approach is discussed in papers E-H-3-1 and E-II-3-2,
which report the derivation of the reduced resistive MHD equations in a form
that is valid in the intermediate regime and the long mean free path regime.
Besides showing that the known results of neoclassical transport theory can berecovered from these equations, the authors also present first results of stability
analyses which indicate a substantial destabilizing effect on intermediate tohigh m, n resistive modes while at the same time leaving unchanged theconventional theoretical low-n mode predictions (which — as shown above tend to agree well with observations).
3.3. Sawtooth modelling
Interest in sawtooth theory, which has been weak for a long time, hasrecently been restimulated by two experimental results that could not be fittedwith the existing theoretical models: the very rapid crash time (about 100 /us)of sawteeth observed in JET, and the very peaked current density profiles
directly measured in TEXTOR.
At the Conference, four papers (E-I-l-1, E-I-l-2, E-I-l-3 and A-VI-4)
analysed with different modelling techniques a basically identical sawtoothmodel. The essential feature of this model is the non-monotonic pre-disruption
q-profile with an off-axis minimum falling below 1 immediately before the onset
of instability. Instability is then a basically ideal, largely pressure driven
m = 1/n = 1 internal kink mode. This model gives a mode pattern with astructure of the cold and hot regions that is in good agreement with SXR
tomography on JET and is also consistent with simulations of the resistive current
diffusion after a sawtooth crash that tends to produce off-axis current maxima
during a transient phase. This resistive current diffusion may also give rise to a

522

LACKNER

more abrupt onset of the instability by allowing the buildup of a larger destabilizing potential before its release. However, complete numerical calculations
presented so far used either a cylindrical model or the reduced MHD equations,both of which are qualitatively inadequate for the pressure driven m = 1/n = 1
mode (some preliminary fully toroidal calculations are reported in paper E-I-l-2).
An appealing feature of the emerging theoretical picture is that variable behaviour
can be expected, depending on the outcome of the competition between pressureand on-axis current peaking as well as the attainment of q = 1 at some radial
position. This would explain the relatively large variety in the appearance ofobserved m = 1 MHD and sawtooth activity. On the other hand, even with thesemodels it is at present still not possible to explain the very fast, precursor-freecrashes observed on JET.
The very peaked current distributions measured by the Faraday rotation
technique in TEXTOR, sometimes corresponding to q0 values significantly
below 1 and persisting throughout the sawtooth cycle, have long been an
embarrassment to MHD theory. A key to their understanding has now been
given by the finding (reported in papers A-V-l and A-V-3) that, as is the casewith higher-m tearing modes, the resistive m = n = 1 internal kink mode canbe stabilized by localized current density flattening in the proximity of theresonant surface, albeit only in toroidal geometry.
3.4. Non-linear MHD stability
A number of rather differing contributions fall under this heading.
Papers A-V-3 and E-I-3 describe applications of three-dimensional non-linear
resistive MHD codes, examining, for instance, the interaction of externallyapplied m = 2/n = 1 fields with the resonant tearing mode. These findings
confirm the results of earlier calculations using a non-linear single-helicity
model [3], leading to the conclusion that direct stabilization can only be obtainedby phase control of the perturbation fields. Dnestrovskij et al. (E-I-4-2) studied(in infinite aspect ratio approximation) the existence of helical equilibria forboundary q-values and plasma current distributions for which the original circularcylindrical equilibria are unstable with regard to the corresponding ideal MHD
kink mode. Such equilibria are indeed found to exist for sufficiently peaked
current density distributions, even with the conducting walls removed to infinity.
Zakharov et al. (E-I-4-1) introduce the 'successive layer principle' as a method
for illustrating and assessing the effect of current profile modifications on
tearing and kink mode stability in the straight cylinder approximation throughan explicit formulation of the contribution of a given layer of plasma currents
to the destabilizing potential.
Reconnection in reversed field pinches (RFP) and tokamaks was studied
with numerical techniques (D-V-l 1) and with analytical techniques (D-V-13).
Sato et al. (D-V-l 1) analysed the differences between total reconnection

IAEA-CN-47/L-I-3

523(|B| = 0 at the reconnection point: the RFP case) and oblique reconnection(finite |B| at the stagnation point: the tokamak case), and studied 'driven'reconnection, i.e. the situation existing when a given plasma inflow speed is
imposed on the boundaries of the computational region. Bhattacharjee et al.
(D-V-13) studied the effective electric field e = <v, X B , ) ( the 'dynamo' field)in the presence of the underlying resistive turbulence, giving expressions for e\\
for the tearing mode and expressions for e^ for the resistive interchange mode.
Applications studied are the compatible steady states for tokamak and RFP
current distributions, and the possibility of current drive through helicity
pumping.
The non-linear dynamics of tearing modes in reversed field pinches,
destabilized by diffusion induced deviations from the Taylor states, wasanalysed in paper D-V-8. The region of instability of m = 1 modes in bothn-number space and radius was found to expand in time because of non-linear
interactions with the m = 0, n = 1 mode and to be accompanied by a dampingof energy to stable modes with shorter wavelengths.
3.5. Effect of suprathermal particles on MHD stability
The interaction of energetic trapped particles with MHD waves was one of
the subjects for which a rather definitive theoretical picture emerged at theConference. This is a complex subject since the interaction depends on threeparameters: co/co^, co/coA and i?(co dh being the precession frequency of hot
trapped particles, coA the toroidal Alfven frequency and 77 the resistivity).
Non-resonant energetic particles have a stabilizing effect, resonant particles
have a destabilizing effect. Likewise, resistivity has a different dominating
effect on low-n internal kink modes (the dissipation of energy in the resonant
layer has a stabilizing effect) and on high m, n ballooning modes (the resistivity
reduces the bending of field lines and hence is destabilizing).
Paper D-V-8 presents a study of the effect of resonant and non-resonant
trapped particles (and also of the excitation of waves by circulating superAlfven particles). On the other hand, paper E-II-2-1 concentrates on resistivity
effects on resonant (fishbone-type) ballooning and kink modes. Finally, themechanisms of phase mixing of the excited Alfven waves and the pumping of
the transferred energy to large k i modes are analysed in paper E-II-2-2.
3.6. Beta limit enhancement in tokamaks
A sizeable number of papers was dedicated to proposals for raising thecritical <(3t) in tokamaks and stellarators. For tokamaks, it is proposed either
to seek a continuous path to the second stability regime or to try to exploit fully
the freedom left by the Troyon formula (jSt> <x I p /aB t to arrive at very large- I p /aB t values by choosing particular plasma shapes or extreme aspect ratios.

524

LACKNERContinuous access to the second stability regime for ideal MHD modes istheoretically possible, either through shaping of the plasma (in particular, paper
E-I-2-1 describes the advantages of the 'crescent' shape over the known beantype plasma cross-sections) or through precise control of plasma current andpressure distributions in conventional circular or D-shaped tokamaks. In thelatter case (E-I-2-1 and E-I-2-2) an extended net shear-free region has to berealized in the plasma interior, with a central q a value of > 1.5. Such current
distributions tend to destabilize free-boundary kink modes, so that a rather
close fitting shell (a wa j]/a plasma = 1.1 -1.2) has to be provided.
Stable <|3t> values in excess of 20% can be reached in the first stability regime
either by providing a strong plasma elongation and curving the resulting belt pinchlike configuration so as to fill the area between two concentric ellipsoidal shells(E-I-2-3) or by using a very low aspect ratio (A « 1.7) (E-I-2-4). For both cases,
parameter sets for the proposed experiments exist (see paper E-I-2-3 and Ref. [4]).
It was also pointed out that the high current density associated with the highly
elongated 'ellipsoidal shell' tokamak gives the best chances for achieving ignition
only by Ohmic heating.
Following a suggestion coming from successful experiments which
demonstrated RF stabilization of modes in mirror machines, Sen et al. (E-II-1-1)
considered the possibility of stabilizing ballooning modes in a tokamak with theponderomotive force exerted by an applied RF field, and concluded this to be arealistic possibility.

4. MODELLING OF CURRENT DRIVE AND RF HEATING
Principally, the papers under this heading can be divided into two groups.
The first group is generally concerned with experimentally well establishedheating techniques and gives elaborate code packages which can be used todetermine power deposition as realistically as possible and to simulate theconsequences for the plasma parameters. The state of technology in this field
is exemplified by paper F-IV-1, reporting a study in which ICRF heating on
JET was modelled using a code package consisting of the following: a twodimensional finite element code for the determination of wave patterns andpower deposition, a combination of a simpler one-dimensional wave propagation
code with a Fokker-Planck analysis of the hot particle distribution functions(to analyse the effect on power absorption by applying the code to the minority
ions or to estimate current drive efficiencies by applying the code to the electrons),
and a conventional transport code coupled with an evaluation of the kink criterion.
A combination of a somewhat simpler (ray tracing) ICRF wave propagation andpower deposition code with a H-dimensional transport code was used by
Swanson et al. (F-IV-4) to predict the expected performance of the planned CIT
ignition experiment. This paper also presents a mathematical model for the

IAEA-CN-47/L-I-3

525

coupling of an incident lower hybrid wave, the ion plasma wave and the ion
Bernstein wave near ion cyclotron harmonics, and gives a solution procedure.
The second group of papers describes the application of an electromagnetic
particle simulation code to more basic problems. The buildup and sustainmentof the toroidal plasma current in a tokamak by injection of a relativistic electron
beam is described in paper E-I-3. The application of electron cyclotron waves
to the heating of overdense plasmas, using the conversion of an X-mode wave toan electron Bernstein wave at the upper hybrid resonance layer, and the stabilization of flute and drift modes by applied RF fields are treated in paper E-H-l-2.

5.

STELLAR ATOR THEORY

The papers on stellarator theory concentrated on the identification ofconfigurations with promise of reactor-relevant MHD-stable (/3> values and on
the analysis of neoclassical transport in these configurations. Two distinctive
approaches have emerged during recent years, one (spearheaded by Heliotron)
involving high rotational transform and strong shear, and the other (referred
to as 'advanced stellarators') trying to minimize the longitudinal plasma currents
and to maximize the magnetic well.
Two possibilities of entering the second stable regime of ideal MHD witha Heliotron E configuration have been discussed: the application of an additionalvertical field (D-V-3) and the superposition of a toroidal field (D-V-2). Wakatani
et al. (D-V-3) also analysed the actually observed stability behaviour of Heliotron E,
which shows good agreement with theoretical predictions for the m = I/n = 1 andm = 3/n = 2 resistive interchange modes. The search for optimal advanced
stellarators described in paper D-I-3 has recently focused on configurations called
HELIAS (a combination of HELIAC and Advanced Stellarator); a case has been
identified which is stable with regard to resistive interchange (argued to be themost restrictive instability type under these conditions) up to <0> = 9%.
Quantitative assessment of neoclassical transport for different stellarator
configurations in the intermediate and long mean free path regimes is still an
important issue. Different numerical approaches exist (solution of the drift
kinetic equation (D-I-4), Monte Carlo simulations (D-I-3)), all of which findlarge differences between their results and those of analytic models and show
a strong effect of the ambipolar radial electric fields on ion transport.
Under normal operating conditions, anomalous transport enhancement
has so far not been found to play a role in the central region of modern
stellarators. A clear deterioration was, however, found in W-VIIA undernearly shear-free operation in well defined regimes of the rotational transform t .
This was generally attributed to externally imposed magnetic field perturbations
and the associated island formation at resonant surfaces and has triggered adetailed analysis (D-I-2) of its dependence on shear, * and perturbation amplitude.

526

LACKNERAn efficient mapping technique was applied in the calculations to analyse magnetic
field braiding, and the consequences for particle transport were studied with aMonte Carlo technique. Of relevance in this context are also results of paper
E-II-3-1 which predict a certain tendency for self-healing due to the additional
currents arising from neoclassical effects.
Because of their complex geometry, stellarators require more involved
techniques for the analysis of the wave patterns and the power absorption in
ICRH. Two approaches to this problem are reported in paper D-III-1: a twodimensional solution method for the partial differential equations and a combination of ray tracing with a boundary layer expansion near cyclotron resonanceregions.

6.

CONCLUSIONS

The achievement of milestones on the way to reactor compatible confinement has been reported at the Conference by authors concerned with experimental tokamak research, but a similar breakthrough has not been achieved in
the theoretical understanding of the major topics. At present, such an understanding seems to be impeded partly by the lack of agreement among experimentalists in the description of the phenomenon of 'profile consistency'. Some
promising ideas — in particular those connected with magnetic braiding andmechanisms of field perturbations — were, however, presented at the Conference
and it is hoped that they will develop to full-fledged theories before the next
Conference.
In other areas, which are at present not so much in the limelight but which
ultimately also require a solution before a reactor can be built, substantial
progress was reported.
Regarding the stellarator, configurations promising reactor relevant <j3>
values have been presented for the two approaches followed at present. A
probably decisive criterion for their ultimate success is the response of thedifferent configurations to externally or internally produced field perturbations.

REFERENCES
[1]
[2]
[3]

[4]

RECHESTER, A.B., ROSENBLUTH, M.N.,Phys. Rev. Lett. 41 (1978) 39.
STACEY, W.M., SIGMAR, D.J., Phys. Fluids 28 (1985) 2800.
MONTICELLO, D.A., WHITE, R.B., ROSENBLUTH, M.N., in Plasma Physics andControlled Nuclear Fusion Research 1978 (Proc. 7th Int. Conf. Innsbruck, 1978),Vol. 1, IAEA, Vienna (1979) 605.
LAZARUS, E.A., ATTENBERGER, S.E., BAYLOR, L.R., et al., Feasibility Study
for the Spherical Torus Experiment, Rep. ORNL-TM-9786, Oak Ridge National
Laboratory, TN (1985).

IAEA-CN-47/L-I-4

SUMMARY ON
INERTIAL CONFINEMENT FUSION*
S.E. BODNER
Naval Research Laboratory,
Washington, D.C.,
United States of America

1.

INTRODUCTION

Inertial confinement fusion (ICF) continues to be an inherently attractive
alternative to magnetic confinement fusion because the high technology component(laser or ion beam) can be located far from the target chamber for easy access andmaintenance [1]. Scientifically and technologically, however, the ICF programme
is not so advanced. Most of the papers presented at this Conference dealt with thebasic physics and driver technology issues, rather than with nrT scaling, ignitionor breakeven.
This summary is organized according to the various ways in which thedriver can be coupled to the pellet: radiation drive with lasers, radiation drivewith light or heavy ions, and direct drive with lasers.

2.

RADIATION DRIVE WITH LASERS

The NOVA facility at Lawrence Livermore National Laboratory (LLNL)
has been completed and is operating. Using 20 kJ at the third harmonic(0.35 nm), the NOVA team has imploded a pellet and achieved a fuel temperature of 1.7 keV with an estimated nr of (2-4) X 1014 s-nT 3 (B-I-2). This is avery positive result, but nr is not by itself a sufficient measure of progress infusion. One additional constraint is the total coupling efficiency of laser energy
to D-T fuel. For electrical energy applications, a high total coupling efficiency
is needed to overcome the inherently low efficiency of laser drivers.
For the radiation drive concept, one component of the total coupling
efficiency is the X-ray conversion efficiency. Laboratories in the USA (B-I-2), in
the Federal Republic of Germany and Japan (B-I-3), and in China (B-III-2) all
reported unexpectedly low X-ray conversion, with the conversion monotonicallydecreasing with increasing laser intensity. For example, LLNL reported an X-ray* The writing of this paper was supported by the Office of Naval Research,
Washington, D.C., USA.
This summary talk was also published in Nucl. Fusion 27 (1987) 505.

527

528

BODNER

conversion efficiency of only 20-30% at a laser intensity of (2-4) X 1015 W- cm" 2 ,rising to 50-60% at an intensity of 10 14 -10 15 W-cm"2.
At the Conference, graphs from LLNL showed that the conversion
efficiency of 50% implies a pellet gain of only about 50 at a laser energy of
5 MJ (B-I-2). For electrical energy applications, the product of the pellet gain
times the driver efficiency must "be at least 10-15. From these numbers we canconclude that a pellet gain of 50 would require a laser efficiency of 20-30%;
this is far above what is expected from any laser. Some of the participants hopeto raise the conversion efficiency by an appropriate choice of the high-Z target
materials, and improvement should be possible. However, for energy applications
there is an additional constraint: a power reactor would use a few tonnes per weekof this high-Z material, and therefore the material has to be both of low cost andnon-radioactive under neutron bombardment.
The Centre d'etudes de Limeil-Valenton (CEL-V) (B-III-1) presented alarge variety of X-ray diagnostics with spatial, temporal and spectral resolution,with an emphasis on very careful absolute calibration. This institute alsopresented results on a new diagnostic technique that can measure radiationprecursors.
The Institute of Laser Engineering (ILE) of the Osaka University and the
Max-Planck-Institut fur Quantenoptik in Garching jointly presented a detailed
study of the radiation flux that is possible in a hohlraum as a function of the
input laser intensity (B-I-3).

3.

RADIATION DRIVE WITH LIGHT IONS OR HEAVY IONSFor radiation drive, light and heavy ions have a substantial advantage overlasers because the driver efficiency is expected to be much higher - approximately
20%. The coupling of these ions to the pellet is also expected to be relatively
advantageous as compared with a laser beam, although the detailed pellet designs
and the techniques for beam propagation and focusing are still under development.
In addition, the cost of a light ion driver is much less than that of a laser. Sandia
National Laboratories (SNL) have already completed a megajoule facility whichaims at the milestone of eventual pellet ignition (B-II-4).
Most current research in ion beam fusion concentrates on driver development
rather than on driver-target coupling. The particle beam fusion accelerator
PBFA-II at SNL is currently in an accelerator activation phase; this is expected tolast another year. A critical test will be the demonstration of the achievement of1014 W-cnT2 on a target. First experiments on targets are expected before thenext IAEA Conference in 1988. At the 1984 IAEA Conference, the generationof a pure lithium source was identified as a critical problem. At this Conference,
SNL announced that a 96% pure lithium source had been developed and was now
being installed on PBFA-II.

IAEA-CN-47/L-I-4

529

ILE announced a three- to fourfold increase in the ion beam intensity on
target, using a non-spherical shaped diode (B-II-5). This is the first confirmationof the shaped diode technique presented by SNL at the last Conference. ILE
also presented results on a new technique that starts with a low voltage pulsed
power source and uses serial connection in the diode to produce a higher voltage.
This concept could lead to repetitive operation of a diode - one of the key
requirements for electrical energy production using light ion fusion.
At the Laboratory of Beam Technology of the Technological University of
Nagaoka (B-III-9) a technique for 'post-acceleration' of ions was demonstrated.
This technique offers the possibility of utilizing intermediate-Z ions rather thanlight ions.
The Tokyo Institute of Technology has developed a cryogenic ion source
for light ion diodes (B-III-7). As an initial test, the Tokyo team has produced
90% pure nitrogen. With colder sources, it expects to produce pure hydrogenand helium in the near future.
At the Instituto de Fusi6n Nuclear (DENIM) in Madrid an advanced 1-Dcomputer code has been developed, primarily for the study of ion beam fusion(B-III-6). This code has capabilities similar to those of the largest computercodes in the world; it has been used to find pellet design regimes for light ion
fusion with enhanced gain.
Research on heavy ion fusion in the USA was presented by Lawrence
Berkeley Laboratory (B-III-3). In a two-year reactor systems assessment it wasestimated that the driver cost could be limited to US $200 per joule with a
3-4 MJ driver, with a net cost of electricity of US $0.05 per kilowatt-hour. This
result is of great importance since the heavy ion fusion programme had previously
been criticized on the grounds that the drivers needed for inertial fusion would be
too large and too expensive.
To obtain a reduction to 3-4 MJ of the ion driver energy, it has beennecessary to redesign the accelerator. Instead of singly charged ions, the new
design uses triply charged ions and multiple parallel beam lines in the sameaccelerator tube. Preliminary experiments with the Single Beam Transport
Experiment support the concept of using multiple charged ions. A proof-ofprinciple induction Linac experiment is under way to test the concepts of current
amplification and acceleration of multiple beam lines.

4.

DIRECT DRIVE WITH LASERS

The direct drive approach to inertial fusion is attractive because of thesimple geometry. Since the target only consists of the pellet, there is less physicalcomplexity than in the radiation drive concept and therefore an inherently lower
risk of failure. In addition, the direct drive approach has one energy conversion
step less, with the laser light directly heating the pellet corona. Each energy

530

BODNER

conversion introduces inefficiencies; with direct drive the pellet gain is therefore
potentially higher and the energy necessary for ignition lower. Ignition andbreakeven are estimated to be in the range of 50-200 kJ. With a 5 MJ laser,
direct drive pellet gains are estimated to be in the range of 150-300, depending
upon pellet design details such as shell thickness. These gains are sufficient forelectrical energy production, assuming a KrF laser with an efficiency of 6-7%.
Because of the high cost of lasers, the direct drive laser fusion concept willprobably proceed in a set of steps:
(a)
(b)
(c)
(d)

Test of critical physics elements (0.1-10 kJ lasers)
Hydrodynamically equivalent implosion (10-100 kJ lasers)
Breakeven and ignition (0.1-1 MJ lasers)
High gain (1 -5 MJ lasers).

Currently, the direct drive laser fusion programme is still in phase (a) (as is
the programme using radiation drive with lasers). In the next phase, the pellet
would be designed to behave hydrodynamically in the same way as a high gainpellet. In other words, it would have the same ratio of pellet thickness to radius,
the same convergence ratio, the same final shell velocity, the same peak laser
intensity and pulse shape, the same sensitivity to non-uniformities, and so on. If
the implosions of phase (b) are successful, the next stage, with a factor of ten
more laser energy, will aim for breakeven and ignition. For a direct drive target
this is expected to require a laser of about 50-200 kJ. The final step before
engineering would then be the demonstration of high pellet gain.
To make a rough comparison, the tokamak programme is now completingthe hydrodynamically equivalent geometry and is ready to attempt ignition.
The direct drive programme is still testing the critical physics elements of highgain, but it seems that the progress reported at this Conference and elsewhere
has been sufficient to justify an attempt at implosion of a hydrodynamicallyequivalent pellet.
The direct drive laser fusion papers presented at this Conference can begrouped according to the various critical physics elements, namely preheat control,
laser symmetry, coupling efficiency, ablative Rayleigh-Taylor instability, and thecentral ignition concept. These five physics issues are arranged in the following
sections according to our level of understanding; preheat control is the best
understood topic, the ignition concept is the least understood. In addition tothese five topics, the requirements for laser development, pellet fabrication,
reactor chamber design, and various supporting physics are briefly discussed.
4.1. Preheat control
Directly driven laser fusion pellets are extremely sensitive to fast electronpreheat; controlling this preheat will probably require a short laser wavelength of
0.25 jum and a laser intensity of less than 1015 W-cm"2. At this Conference, theIAEA-CN-47/L-I-4

531

Laboratoire de physique des milieux ionises of the Ecole polytechnique at
Palaiseau presented experimental data (B-II-3) indicative of very low preheat
with 0.26 ixm laser light. Further experiments are planned with longer scale
length plasmas. The Naval Research Laboratory in Washington (NRL) has
previously presented experimental results showing that their induced spatial
incoherence (ISI) optical smoothing technique reduces most plasma instabilities
if the laser wavelength is less than 1 jim. At this Conference, NRL presented
data (B-I-5) showing that ISI with both spatial and temporal incoherence also
reduces the Raman instability, while the use of only spatial incoherence without
temporal averaging (the random phase plate) slightly enhances the Raman modeover an ordinary laser beam.
4.2. Laser symmetry
To achieve a high pellet gain, the implosion symmetry must be uniform to
1-2%. This uniformity can be achieved in principle with a laser that uses oneof several optical techniques which add spatial and temporal incoherence to thelaser beam. The smoothing has been experimentally demonstrated on an
individual laser beam, and calculations indicate that the uniformity on an
imploding sphere should be sufficient. In the future, these optical smoothing
techniques will probably be installed on various spherical implosion facilities,
but this will require further time and money. At the Conference, some laboratories presented status reports on spherical implosion experiments that did not
utilize any optical incoherence for smoothing.
In the Laboratory for Laser Energetics (LLE) of the University of Rochester,
an ongoing programme aims at improving the laser uniformity sufficiently toobtain an implosion to 100-200 times solid density (B-I-5). This demonstration
would then be followed by a larger facility that would incorporate spatial andtemporal incoherence smoothing techniques. Currently, the OMEGA laser at
LLE can produce 15% RMS non-uniformity, summed over all surface harmonicmodes. With improvements now under way, LLE expects to reduce this nonuniformity by a factor of three in the near future.
At the Electrotechnical Laboratory in Tsukuba thermal smoothing has been
experimentally measured with both an intensity modulated laser beam and alaser beam with a random phase mask (B-III-8). Consistent with theory, significant
thermal smoothing of higher wavenumber disturbances was found with 0.53 /nm
laser light but negligible smoothing with 0.26 jum laser light.
4.3. Coupling efficiency
LLE and NRL reported that they have found pellet design regimes with ahydrodynamic rocket efficiency of 10-15%, the higher values requiring thinnerpellet shells (B-I-5). Thinner shells are of course more sensitive to the ablative
Rayleigh-Taylor instability. This leads naturally to the next topic.

532

BODNER4.4. Ablative Rayleigh-Taylor instability
When a high gain direct drive pellet implodes, the in-flight aspect ratio,
R/AR, can approach 100 to 150, the shell thickness being defined by the 1/edensity surfaces. One of the outstanding issues in laser fusion is whether infact such thin pellet shells can be stably driven. To achieve stable drive, thegrowth rate, 7, of the ablative Rayleigh-Taylor instability must be reduced
below the classical value, 7 = y'kg. Computer simulations at LLE, the Rutherford
Appleton Laboratory in Chilton, ILE and NRL all predict that the growth rate is
reduced. Calculations at ILE and NRL indicate that the growth rate can be reduced
to 30% of the classical value with 0.25 /im laser light. So far, however, there havebeen very few experimental data on the growth rate of this mode. At this Conference, two laboratories, Rutherford Appleton and NRL, announced new experimental methods and results (a third laboratory, ILE, did not present their recent
Rayleigh-Taylor experiments).
The Rutherford Appleton Laboratory has used an ingenious alpha particle
backlighting of an ablatively accelerating foil to measure the eventual breakupof that foil (B-II-1). The alpha particles came from the thermonuclear burn ofa separate imploding pellet.
NRL directly measured the linear growth rate of the Rayleigh-Taylor modeby using two separate backlighters shining through an ablatively accelerated
foil onto X-ray streak cameras (B-I-5). One source was used to measure theacceleration of the foil, the other source measured the changes in the mass perunit area in the thick and thin regions of the foil. The growth rate was consistent
with the computer simulations for perturbations of 100 and 150 jum, but it wasmuch less than that of the non-ISI simulations for 50 Mm perturbation. Theseexperiments were performed with 1/^m laser light. Future experiments areplanned with 0.53 /urn laser light and perhaps with 0.25 //m KrF laser light.
4.5. Ignition concept
Here, the phrase 'ignition concept' refers to the point design of a high gainpellet, the creation of a central hot spot that can ignite the fuel, and the RayleighTaylor mix of the hot fuel with the surrounding cold and denser fuel. Several
laboratories reported spherical implosion experiments that attempted to create
optimum fuel conditions, given the constraint of limited laser energy.
ILE (B-I-4) and LLNL (B-I-2) have both obtained about 1013 neutrons bycarefully tailoring the shock waves in a high aspect ratio pellet. ILE emphasized
the use of the technique of stagnation-free implosions for a minimization of theshell/fuel mix, and suggested that these implosions are the equivalent of thetokamak 'supershots'. The type of pellet used in these high neutron yield experiments cannot be extrapolated to high gain, but the experiments do demonstrate
substantial control of the implosion process, and there is the possibility of longterm application of this stagnation-free technique.

IAEA-CN-47/L-I-4

533

LLE has an ongoing campaign to compress fuel to about 100-200 times
solid density. Progress is limited by laser asymmetry, as discussed in Section 3.2.
At this Conference, LLE reported (B-I-5) on various complementary diagnostic
techniques that are being used for ensuring that the measurement of high density
is accurate; these diagnostics include secondary reaction products, nuclear activation, and X-ray spectroscopy.
The Ecole polytechnique at Palaiseau has performed pellet implosion using
its unique 0.26 pm spherical implosion facility (the fourth harmonic of Nd:glass).
It claims that the implosion was degraded by the fuel/shell mix which was in turndriven by laser non-uniformities on the imploding shell (B-II-3). The Ecolepolytechnique is now trying to improve the laser uniformity.
CEL-V reported results of planar experiments designed to isolate the physics
of a Rayleigh-Taylor mix (B-I-l). Extrapolating a technique first utilized by ILE,
CEL-V built a target of aluminium layered with a thin coating of gold on each side.
By accelerating the target with one laser and burning through the other side witha second laser beam, CEL-V was able to demonstrate that the acceleration process
produced a mix on the rear of the foil. This technique should be powerful for
future studies of the Rayleigh-Taylor mix.
4.6. Laser research and development
Both LANL (B-II-2) and the Tsukuba Electro technical Laboratory (B-III-8)
reported briefly on their ongoing effort to build KrF lasers. Both laboratories
will probably incorporate ISI optical smoothing, and both will perform one-sided
flat target experiments.
4.7. Pellet fabrication
LLNL announced the development of a low density foam pellet that can'wick' up the D-T fuel. This is a very significant development, since previously
there had been no way to build a thick cryogenic D-T fuel layer into a pellet shell.
4.8. Reactor chamber design
A reactor design that may be applicable to direct drive laser fusion is being
developed by the University of Wisconsin. This reactor would have a simple drychamber wall and a gas fill. However, no direct-drive reactor designs were
presented at this Conference.
4.9. Supporting research
There is a variety of important fundamental research that does not fall into
any of the above programmatic categories but that is necessary if we are to534

BODNER

succeed in laser fusion. For example, the University of California, Los Angeles,
has completed a microwave simulation of laser-plasma interactions (B-III-4).
With the longer wavelength microwaves, the saturation mechanisms of ion waves
driven by two electromagnetic waves can be studied. At the University of British
Columbia, detailed quantitative studies of radiative and shock heating in simple
planar targets have been performed (B-III-10).
Special mention should be made of the effort made at several laboratories
to understand thermal transport, which led to a non-local transport theory!
Early in the history of the laser fusion programme, several laboratories notedthat transport in layered targets did not match the classical Spitzer-Harm formula,even though the gradient scale lengths were many times the electron mean freepath. This was followed by years of careful experimentation under various
conditions at many laboratories, and the development of some crude transportmodels that are best forgotten. But eventually this led to a sophisticated FokkerPlanck study at the Rutherford Appleton Laboratory demonstrating the importance of electrons with a velocity several times the thermal velocity. At thisConference, a sophisticated non-local transport model was presented by a scientificgroup working with the Laboratoire de physique des milieux ionises of the Ecolepolytechnique at Palaiseau (B-II-3). (Other non-local transport models have beendeveloped at ILE and at LLNL, but they were not presented at the Conference.)

5.

CONCLUSIONS

The inertial fusion concept has made substantial progress since its inception.
Large Nd: glass lasers and pulsed power devices have been successfully developed,
beyond most scientists' initial imagination. Plasma coupling problems that led
to fast electron preheat have been essentially solved. The initial p'ellet designconcepts that started the programme were flawed, but they have been replacedby superior pellet design concepts. Optical techniques have solved the fundamental problem of laser non-uniformity. Some aspects of the programme are
now in the process of shifting, from studies of plasma instabilities to the studyof fluid instabilities, from basic plasma physics to hydrodynamically equivalent
implosions, from pulsed power development to ion beam target experiments,from Nd: glass laser research to, hopefully, the development of KrF lasers andheavy ion beams. As could be noticed at this Conference, inertial confinement
fusion is healthy and lively, with several potential candidates for attractive
electrical energy applications.

REFERENCE[1]

Review of the Department of Energy's Inertial Confinement Fusion Program, National
Academy of Sciences, Washington, D.C. (March 1986).

IAEA-CN-47/L-I-5

SUMMARY ON
TECHNOLOGY AND REACTOR CONCEPTS*
R. HANCOX
Euratom-UKAEA Fusion Association,
Culham Laboratory,Abingdon, Oxfordshire,
United Kingdom

1.

INTRODUCTION

Technology and reactor concepts were covered in fifteen presentations.
A broad range of topics was discussed, including alternative confinement systems,
alternative fuel cycles, a range of technologies, and safety and environmental
aspects of fusion.
Technology and reactor concepts represent only a small part of thisConference and therefore the papers discussed here cover only a small part of
the worldwide activity in these areas. This is especially so because in 1986 the
IAEA also organized two technical meetings relating to fusion reactors - theTechnical Committee on Fusion Reactor Design (Yalta, 26 May to 6 June 1986)
and the Technical Committee on Fusion Reactor Safety (Culham, 3 to 7 November1986). It is therefore appropriate to incorporate some background material fromthese meetings into this review.

2.

NEAR TERM TOKAMAK REACTORS

Each of the four countries, or group of countries, involved in fusion
research and development is studying at least one Near Term Tokamak Reactor
which will follow the present generation of large experiments. The objectivesof these devices vary, from the simple physics objective of demonstrating an
ignited and burning plasma with minimum technology to the demonstrationof long term plasma burn and testing of technological components, with sufficient neutron flux to be regarded as a significant step towards the requirementsof a demonstration reactor. Progress on two of these reactors, CIT and NET,
was discussed at this Conference.
The Compact Ignition Tokamak (CIT) (H-l-2) has the primary objectiveof satisfying the physics requirements for ignition with the shortest pulse lengthand the smallest major radius, and hence minimum cost, consistent with prudent* This summary talk was also published in Nucl. Fusion 27 (1987) 510.

535

536

HANCOX

engineering choices. These objectives lead to a high field machine with a magnetic
field in the plasma of 10.4 T, generated by copper coils pre-cooled to liquid
nitrogen temperature. The technology has already been proved in the Alcator C
and Frascati FT high field tokamaks. In the present design the average stress inthe toroidal field coils is allowed to approach 85% of the yield strength of thecomposite copper and steel material and, as a result, the machine will be limited
to 3000 full power burn pulses although 50 000 pulses would be possible at 70%of the full field and current. The neutron wall loading in such a machine is high
(6.8 MW- m~2) and thus wall protection and plasma exhaust are major problems.
The design of this machine is well advanced and has been demonstrated to befeasible, although, as is the case with all near term tokamaks, it has beennecessary to extrapolate considerably from present plasma physics experienceand in particular to assume H-mode operation with high levels of auxiliary
heating. If funded in 1987, CIT could be in operation in conjunction withTFTR at Princeton in 1993.
Before considering the NET design, it is interesting to compare the four
devices of this type currently at the conceptual design stage. These are:Next European Torus (NET)Fusion Engineering Reactor (FER)
Tokamak Ignition Experiment (TIBER II)
Fusion Test Reactor (OTR)- Euratom- Japan- USA- USSRAll four groups also collaborate in the INTOR Workshops.
The parameters of these machines, as presented at Yalta, are shown in
Table I. There is a relatively narrow range of maximum magnetic field andplasma current. All reactors have elongated plasmas and poloidal divertors, andall have fusion power levels in the range of 300-600 MW. Only three devices,
however, aim at breeding tritium and only OTR aims at full breeding; OTR also
has provision for fissile fuel production.
The physics objectives of the NET device (H-I-l) are to produce a plasma
with parameters and performance relevant to a fusion reactor, and this implies
substantially longer plasma burn times than in CIT. In addition, the machine
has extensive technological objectives, including an integrated test of reactor
components and systems as well as engineering tests and the development of
in-vessel components, and therefore a minimum neutron fluence of about
0.8 MWa-m" 2 is judged necessary. Thus an integrated burn time of 108 s isrequired at the design wall loading of 1 MW-m"2, with a duty cycle greater than
90% in the final phase of operation compared to a value of less than 0.2% for
CIT. Two variants of the reference concept were presented, involving singlenull or double-null poloidal divertors, the latter leading to a physically smallermachine because of the larger plasma elongation and triangularity. A further

537

IAEA-CN-47/L-I-S

TABLE I. NEAR TERM TOKAMAK REACTORS
OTR

FER

TIBER-II

NET

INTOR

R(m)

6.2

5.2

3.0

5.2

4.9

a(m)

1.5

1.1

0.8

1.4

1.2

b/a

1.5

1.5

2.2

2.2

1.6
5.5

B0(T)

5.6

5.3

6.0

5.0

I (MA)

8.2

5.9

10.0

10.8

8.0

T(S)

600

2000

s.s

600

200

Power (MW)

520

297

290

600

570

0

?

~0.4

>0.6

TBR

1.05

Notes: TBR = tritium breeding ratio
s.s = steady state? = to be decided.

possible variant is operation of the machine in its first phase with a blanket/
shield of reduced thickness, allowing a substantially larger plasma (15 MA
instead of 10 MA) for plasma optimization experiments. The NET activity hasentered the pre-design phase in 1986, and a device could be constructed for
operation in the late 1990s.
Before leaving the topic of Near Term Tokamak Reactors, mention must
be made of the American initiative announced at this Conference by J.F. Clarkein his Artsimovich Memorial Lecture [1], proposing international co-operationin the preparation of a conceptual design report for an engineering test reactor.
The USA has indicated its willingness to commit the necessary resources to thisactivity for a period of three years, and proposes that the objective should beachieved through already established mechanisms in the IAEA. If accepted,
this initiative would represent a significant step forward towards the construction of a tokamak reactor.

3.

ALTERNATIVE CONFINEMENT SYSTEMS FOR REACTORS

Looking further ahead to conceptual designs of power reactors, it isnoticeable that no commercial tokamak reactor designs were reported at thisConference. Indeed, at the Yalta meeting it was noted that no major commercial tokamak reactor design study had been started since the previous Fusion
Reactor Design Meeting held at Tokyo in 1981. This does not mean that there
has been no interest in power reactors; a demonstration reactor study has been

538

HANCOX

undertaken in the UK,and in the USA the STARFIRE study has beenreconsidered and the substantial MARS and mini-MARS mirror reactor studies
have been completed. There has also been significant interest in other nontokamak reactors, with a major compact RFP study (TITAN) in progress in the
USA and several smaller stellarator reactor studies. At this Conference there
were four papers on power reactor concepts based on alternative confinement
systems: a dense Z-pinch reactor, an ICF reactor, a muon catalysed hybrid
reactor and a compact torus.
The dense Z-pinch reactor study undertaken by Imperial College in London(H-III-7) has the objective of reducing the overall size and complexity of thefusion system. Such a pulsed device would have a 0.1 m long plasma with a
50 nm radius, lasting for less than 1 /JLS. An energy release of 5 MJ per pulse
would require a repetition rate of 100 Hz for a 500 MW thermal reactor. The
power supply for such a reactor must produce repetitive current pulses witha very fast rate of rise (dl/dt ~ 1013 A-s" 1 ) using passive elements and with anenergy recovery efficiency of at least 80%. Other major engineering problems
include electrode erosion due to both thermal losses from the plasma and alphaparticle deposition, and a blanket structure to withstand repetitive shocks. One
nuclear blanket configuration considered involves the use of a vortex of liquid
lithium acting as breeder and coolant. Since the D-T fuel is at high pressure,a pressure vessel similar to that required for a PWR would surround the reactor
structure. Whilst the study is far from completion, it was suggested that, subject
to certain assumptions, the Z-pinch might lead to a relatively small and simple
magnetic fusion reactor with a power density somewhat greater than that of a
PWR, but there was no evidence to support the claim that this system would be
economically attractive.
The novel feature of the advanced ICF reactor Takanawa-I (H-III-4) is thatit uses a D-T ignitor inside a D-D fuel pellet. After compression of the target,
ignition occurs at a temperature of around 10 keV and the D-T reactions heat
the remainder of the target to temperatures in excess of 100 keV for the D-D
reactions to become effective. The majority of tritium generated in thesereactions will be burnt within the pellet, the remainder being sufficient tomaintain the fuel cycle without external breeding. Simulations of the pellet
burn show that the required amount of tritium is less than 3% of the quantityof deuterium fuel and that an internal tritium breeding ratio higher than unity
can be achieved when PdRd is greater than 10 or 15 g-cm" 2 , although thisrequirement is a factor of three higher than is necessary for a conventional
D-T pellet. The fusion gain in the pellet is relatively low (~200) and therefore
a high efficiency driver (~20%) is required for such a system; a particle beam
driver was briefly discussed as one of the candidates. Another novel feature of
the reactor is the proposed use of silicon carbide for the reactor vessel which
would be protected by a falling curtain of liquid lead. A claimed advantage ofthis construction is that all the materials have low induced radioactivity

IAEA-CN-47/L-I-5

539(lCT6 Ci-W"1 after one year) and therefore problems of remote maintenanceare mitigated, but the calculations did not include the effects of impurity
elements which might be very significant.
Another new approach to fusion reactors is the possibility of muon
catalysed fusion, and this topic was treated at the Conference in a paper
presenting a muon catalysed hybrid fusion-fission reactor (H-II-1). It has been
known for several years that it is possible to catalyse fusion reactions in moleculesby reducing their size by binding them with negative muons. The prospect ofachieving a net energy gain in this way has improved in recent years as new
theoretical work and experimental evidence suggested that a muon may catalyse
as many as 100 fusion reactions during its lifetime, which is typically 2 ps. It is
now suggested that the additional energy multiplication possible in a hybrid
system is sufficient for such a combination to lead to a net energy gain. A
synergistic concept was proposed in which a 3 GeV tritium beam was injected
into high pressure D-T gas flowing through a magnetic mirror. Pions, which are
created within the magnetic field, are confined and decay into muons. A 1 GW
power reactor would need about 2 kg of tritium contained in a 20 cm radiusmirror at half the density of liquid hydrogen, which corresponds to a pressureof 1500 atm. The nuclear blanket would breed sufficient fuel to support fivesimilarly sized fission reactors, and with an accelerator efficiency of 60% and arecirculating power fraction of 13% an overall energy gain of 13 appears possible.
If the latest measurements of 160 fusion reactions per muon were assumed, thisgain would be correspondingly higher.
The fourth reactor system based on an alternative confinement configuration is a compact torus formed by the injection of plasma rings at either end ofa magnetic mirror (H-II-2). Such a system should be able to contain a high beta
plasma, but the energy confinement times would be reduced by lower hybrid
drift instabilities caused by the steep density gradients. Nevertheless, parameters
were proposed for a quasi-stationary reactor in which plasma injection at afrequency of 1 Hz resulted in limitation of cyclic variations in power output
to 15%; such a reactor would have a magnetic energy requirement an order ofmagnitude less than a tokamak reactor of equivalent output. A further development of this concept involved the possibility of a transition from low temperature operation (10 keV) with D-T fuel to high temperature operation (50 keV)with D- 3 He fuel over a period of many cycles.
In summary, for all such alternative reactor concepts, either very large
extrapolations of physics performance or the solution of difficult engineering
problems must be assumed. Despite optimistic assumptions regarding theefficiencies of the major components or processes, none of these concepts has
substantiated claims to generate cheaper fusion power. The development ofsuch reactor concepts should be encouraged, but, so far, no alternative toreactors based on the conventional magnetic confinement systems appears
convincing.

540
4.

HANCOXALTERNATIVE FUEL CYCLES

Mention has already been made of the use of D-T drivers in D-D targets
for an ICF reactor, the use of D- 3 He fuel in a compact torus, and the incorporation of fissile fuel production in muon catalysed reactors. There remains thediscussion of D-D based fuel cycles, which have the advantage that full tritium
breeding is no longer an essential requirement. However, the difficulty withsuch concepts is that more exacting plasma confinement criteria must be
achieved. A discussion of deuterium based fuel cycles was presented at theConference (H-III-9) which is general in that it is applicable to any steady state
confinement geometry but detailed in that it includes all possible plasmacomponents. Calculations of thermal and suprathermal energy distributions
and of synchrotron radiation were included and for the numerical computations
an extensive data file was developed which includes the relevant cross-sections
and reaction rate parameters. Results were presented for various types of fueland operational modes, but the major conclusion was that only a few deuterium
based fusion alternatives deserved further attention unless highly specialized
applications are envisaged.

5.

TECHNOLOGICAL DEVELOPMENTS

Three major areas of technology were covered - superconducting magnets,
neutral beams and high heat flux components such as those required in a divertor.
In addition, practical experience in commissioning the large fusion experiment
MFTF-B was reported.
There has been significant progress in the development of superconducting
magnets, and activities reported at this Conference include the development of
new materials and technologies required for the Japanese Fusion Experimental
Reactor (FER) (H-III-6), results from the international Large Coil Task (LCT)at Oak Ridge National Laboratory (H-I-4), and the initial operation of thesuperconducting tokamak TRIAM-1M (H-II-3).
The steady state toroidal field coils proposed for FER (H-III-6) are
9 m X 11 m in size and the pulsed poloidal field coils are up to 20 m in diameter.
The toroidal field coils will have a peak magnetic field of 12 T, and the present
design employs intermetallic niobium-tin conductors. However, alternative superconductors have been investigated, including modifications of the alloys niobiumtitanium or vanadium-titanium which must be operated at superfluid heliumtemperatures to obtain high enough critical fields, and the less well known Lavesphase compound V2(Hf,Zr) which also offers attractive current densities at these
low temperatures. Development of these materials is still in the early stages, butthey have the potential of improved performance for future applications. The
performance of coils cooled with superfluid helium may be more dependable

IAEA-CN-47/L-I-5

541

than that of coils cooled by pool boiling, and alloy superconductors may be
more reliable than the brittle intermetallic compounds in large reactor
components.
Large scale testing of toroidal field coils is being undertaken in the LCT(H-I-4), which includes six coils of 4.5 m X 3.5 m size, each of which weighs
40 t. The coils have been built in the USA, the Federal Republic of Germany,
Switzerland and Japan. Both niobium-titanium and niobium-tin superconductors have been used and the coils are operated with both pool boilingand internal cooling with liquid helium at a pressure of 15 atm. There are otheralternatives of construction — pancake or layer windings, stainless steel oraluminium structures, adiabatic or cryogenic stabilization - and all are fully
instrumented. Since testing began in February 1986, each of the six coils has
been tested individually at its full rated current, although in this situation thecoils do not experience the full design magnetic field. Each of the coils has been
subjected to local internal heating in order to study stability and to test protection systems. More recently, all six coils have been successfully tested together
at full current, generating a field of 8 T, and they will soon be subjected topulsed external fields in a simulation of normal tokamak operation. Successful
completion of the programme will go a long way towards demonstrating thatsuitable superconducting technology is available for a Next Step Tokamak.
Another important demonstration of the successful development of superconducting magnets is the operation of the TRIAM-1M tokamak at Kyushu
University (H-II-3). The sixteen D-shaped toroidal field coils of this tokamak
experiment have a bore of 0.44 m X 0.67 m and a major radius of 0.85 m, andgenerate a magnetic field on the plasma axis of 8 T and a peak field of 11 T.
The niobium-tin conductor is bonded to a high-purity aluminium stabilizer
and is cooled by pool boiling helium. A notable feature is the high average
current density of 57 A-mm" 2 . First operation was in June 1986, and themagnet has been operated at full field with plasma currents up to 150 kA. For
comparison, the only other operating tokamak with superconducting toroidal
field winding is T-7 at the Kurchatov Institute, which has niobium-titanium
conductors producing a magnetic field on the plasma axis of 3T (H-III-10).
However, in 1988, both the Tore-Supra tokamak at CEN de Cadarache and the
T-15 in Moscow should be operational.
Neutral beam injectors based on positive ion sources are a fully developed
technology and are used in the majority of the world's fusion experiments,
including injection into the three large tokamak experiments. For example,
on JET, 80 keV deuterium beams are injected at a power level of 10 MW for
pulse lengths of up to 7 s. However, Next Step tokamaks and power reactors
will require higher injection energies because of their larger physical size and,since the neutralizer efficiency for positive ions decreases with increasing ion
energy, it will be necessary either to add efficient energy recovery of the nonneutralized beam or to use negative ion sources which allow substantially higher

542

HANCOX

neutralization efficiencies. For FER it is proposed to inject a 200-500 keVbeam at a power level of 20 MW for both plasma heating and plasma current
drive, with steady state capability (H-I-3). Preliminary experimental results
suggest that a suitable negative ion source can be developed. Furthermore,because of the low beam divergence of 0.3°, a 20 m long beam tube/neutralizer
could be used so that the ion source would be far removed from the reactor
and would suffer minimal radiation damage. With this configuration it is also
possible to adjust the beam profile by a magnetic controller and thus to
influence the beam current profile which affects beta limits, sawtooth activity
and disruptions. Conceptual design studies have also considered the possibilityof steady state current drive in FER for which 70 MW would be required todrive a plasma current of 5.7 MA.
A far less advanced technology is the use of high heat flux materials andcomponents to overcome plasma-wall interactions, either on the wall of theplasma containment vessel or on a divertor. The demands on divertor target
materials are most severe because the particle flux and its associated power
loading are not only concentrated on a small area but also poorly defined since
they depend on plasma processes in the cool edge of the plasma and in thedivertor chamber. One solution to the latter problem has been examined anda 'super high density' regime of divertor operation has been proposed for a
tokamak reactor in which a dense radiating plasma is formed in the divertor
chamber but remote from the target (H-III-10). Under these conditions theplasma close to the divertor target is a cool, weakly ionized gas blanket in which
recombination and multi-body interactions occur and from which neutral gas
can be exhausted. This regime is more extreme than the similar regime proposed
for the INTOR divertor. However, at the present time, all such possibilities can only
be examined in computer models based on general assumptions about theboundary plasma and remain to be demonstrated in practice.
The problem of surface coatings for high heat flux components was discussed
and illustrated by experiments on titanium carbide deposited on molybdenum(H-III-5). High heat fluxes and high temperatures result in a variety of phenomena, including vaporization, melting and chemical interactions between thelayer and the substrate. Under steady state conditions, diffusion results in theformation of complex compounds which may have high vapour pressures; in thisparticular case, titanium was lost from the coating. At higher heat fluxes theinterface compound may melt, enhancing chemical reactions and the propagationof the diffusion zone, with a rapid increase of the loss of titanium. It is concluded
that additional test facilities are essential for further investigation of these problems
and their solution.
Two of the three technologies described in the above paragraphs - large
superconducting magnets and high power neutral beam injectors - have been
demonstrated in the MFTF-B acceptance tests (H-III-2). The completion ofthis major tandem mirror experiment, with a vacuum vessel 58 m in length andIAEA-CN-47/L-I-S

543

a superconducting magnet weighing 1050 t, has been a major engineering achievement. The machine contains 24 superconducting coils in a linear array with acentral field of 1.26 T and 16 end-mirrors with magnetic fields up to 12 T. Despite
substantial magnet movements during cooldown and owing to magnetic loading,the largest field alignment error under full operating conditions was only 3 mm,
as determined with an electron beam technique. Absolute values of themagnetic field, determined with nuclear magnetic resonance techniques, were
found to be within 2% of the predicted values. Two of the 80 kV neutral beam
injectors were tested at a power level in excess of 5 MW for half-second pulses.
The satisfactory operation of these and other components demonstrates thehigh level of engineering which is now achievable in large and complex fusion
devices.

6.

ENVIRONMENTAL ISSUESThe environmental and economic implications of fusion are subjects whichhave received widespread attention in 1986. It is no longer possible to justify
fusion research by simple statements that fusion is a clean and cheap source ofenergy. There is pressure from the public and from governments who want
quantified statements of the risk and cost of fusion power. A report on this
topic was published by the Atomic Energy Society of Japan in March 1986.
In the European Community, statements are being prepared at the request of
the European Parliament, and a major document comparing fission and fusion
is in preparation in the USA. Public awareness of the hazards of nuclear power
as a result of the Chernobyl accident causes questions to be asked about fusion
as well as fission power, and the fusion community will be wise to listen to thesequestions and to answer them carefully and accurately.
On the positive side, it can be demonstrated that most proposed fusion
reactor designs have significant safety and environmental advantages over fission
reactors. For example, preliminary information from the American study on
Environmental, Safety, and Economic Aspects of Magnetic Fusion Energy(ESECOM) presented at the Culham meeting suggests that in an accident releasing
a given fraction of the radioactive inventory the resulting whole-body dose at thesite boundary would be a factor of 10-100 lower for a fusion reactor than for aliquid metal fast breeder reactor. Furthermore, 'intruder doses' resulting fromburied wastes, for a period of between 100 and 1000 years, are expected to bea factor of 10-100 lower, provided low-activation steels or more advanced lowactivation structural materials are used.
Of the several areas of concern, including the tritium inventory and its
containment, structural activity and its implications for occupational and public
exposures, as well as possible accidents and their consequences, only one area
was considered at the Conference (H-I-5). It was shown that the activity of

544

HANCOX

structural materials such as steel can be reduced by elemental modifications,but that the benefits relate to long-term issues, such as waste disposal and thepossibility of material recycling, rather than to reduced exposures during
maintenance or in accident situations.

7.

CONCLUSIONSIn summary, technology has developed on a broad front as a result of boththe successful construction of large plasma physics experiments and programmes
specifically designed to prepare for the needs of Near Term Tokamak Reactors.
The prospects for building these reactors are still reasonably good, subject tocontinuing progress in plasma physics, but progress may be slower than enthusiasts
would wish and at a cost that governments may not easily accept. The prospects
for future commercial reactors appear less certain, and reactor studies have notprogressed markedly in the past two years. The tokamak is not an ideal reactor
system and many alternatives are still being proposed, but in most cases either
the physics basis or the engineering embodiment are too weak to support
convincing conceptual designs. Finally, there is a growing need to quantify thesafety and environmental advantages of fusion. If the task of developing a safeand economic fusion reactor appears daunting, however, it should be rememberedthat major advances in plasma physics and fusion technology have been
demonstrated at this Conference and that continuing progress can bring success.

REFERENCE[1]

CLARKE, J., Nucl. Fusion 27 (1987) 475.

CHAIRMEN OF SESSIONS
Volume 1
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. ZAKHAROVUSSRSession F-I

O.S. PAVLICHENKO

USSR

Session F-II

R.R. PARKER

USA

Session F-III

F.G. PREVOT

FranceVolume 2
Session E-I

D.B. NELSON

USA

Session E-JJ

R.R. PARKER

USA

Session C-I

D.D. RYUTOV

USSR

Session D-I

L.M. KOVRIZHNYKH

USSR

Session D-H

T. UCHIDA

Japan

Session D-III

G. GRIEGER

Germany, Fed. Rep

Session D-IV

F.L. RIBE

USA

Volume 3
Session B-I

R.L. SCHRIEVER

USA

Session B-II

C. YAMANAKA

Japan

Session G-I

B.B. KADOMTSEV

USSR

Session H-I

K. TOMABECHI

Japan

Session H-II

R.W. CONN

USA

Session K

R.S. PEASE

UK
545

SECRETARIAT OF THE CONFERENCE
Scientific
Secretaries:

M. LEISER
A.A. SHURYGIN

Division of Research andLaboratories, IAEA

Administrative
Secretary:

Edith PILLER

Division of External
Relations, IAEA

Editors:

J.W. WEIL
Maria DEMIR

Division of Publications,
IAEA

Records Officer:

J. RICHARDSONDivision of Languages,
IAEA

546

LIST OF PARTICIPANTS
Abe, H.

Department of Electronics, Kyoto University,
Kitashirakawa, Sakyo-ku,
Kyoto 606, Japan

Aikawa, H.

Naka Fusion Energy Research Establishment,
Japan Atomic Energy Research Institute,801-1 Mukoyama, Naka-machi, Naka-gun,Ibaraki-ken 311-02, Japan

Airoldi, A.

Associazione Euratom-ENEA, Istituto di Fisica del Plasma,
Consiglio Nazionale delle Ricerche,
Via Bassini 15,
1-20133 Milan, Italy

Alexander, K.F.

Zentralinstitut fur Elektronenphysik,
Akademie der Wissenschaften der DDR,
Hausvogteiplatz 5-7,
D-1086 Berlin, German Democratic Republic

Amano, T.

Institute of Plasma Physics, Nagoya University,
Furo-cho, Chikusa-ku,
Nagoya 464, Japan

Amemiya, N.

Sekiguchi Laboratory,
Department of Electrical Engineering, University of Tokyo,
7-3-1 Hongo, Bunkyo-ku,'
Tokyo 113, Japan

Ando, A.

Department of Physics, Kyoto University,
Kitashirakawa, Sakyo-ku,
Kyoto 606, Japan

Ashida, H.

Electrotechnical Laboratory,
1-1-4 Umezono, Sakura-mura, Niihari-gun,Ibaraki-ken 305, Japan

Azechi, H.

Institute of Laser Engineering, Osaka University,
2-6 Yamada-oka, Suita,
Osaka 565, Japan

Azodi, H.

Plasma Physics Group, Nuclear Research Centre,
Atomic Energy Organization,
P.O. Box 11365-8486,
Tehran, Iran

547

548

LIST OF PARTICIPANTS

Azumi, M.

Naka Fusion Energy Research Establishment,
Japan Atomic Energy Research Institute,801-1 Mukoyama, Naka-machi, Naka-gun,Ibaraki-ken 311-02, JapanBaba, T.

Plasma Physics Laboratory, Kyoto University,
Gokasho, Uji,
Kyoto 611, Japan

Bakos, J.S.

Central Research Institute for Physics,
P.O. Box 49,
H-1525 Budapest, Hungary

Baldwin, D.E.

Lawrence Livermore National Laboratory,
University of California,
P.O. Box 5511,
Livermore, CA 94550, United States of America

Bamiere, C.G.H.

Secretariat general de la defense nationale (AST),
51, boulevard de Latour Maubourg,
F-75700 Paris, France

Barnes, D.C.

Science Applications International Corporation,
1406 Camp Craft Road, Ste. 104,
Austin, TX 78746, United States of America

Becker, G.

Max-PIanck-Institut fur Plasmaphysik,
Boltzmannstrasse 2,
D-8046 Garching, Federal Republic of Germany

Behringer, K.

JET Joint Undertaking,
Abingdon, Oxfordshire OX14 3EA, United Kingdom

Bellan, P.M.

California Institute of Technology,
Mail Station 128-95,
Pasadena, CA 91125, United States of America

Berge, G.

Department of Applied Mathematics, University of Bergen,
Allegb. 53/55,
N-5000 Bergen, Norway

Bemabei, S.

Princeton Plasma Physics Laboratory, Princeton University,
P.O. Box 451,
Princeton, NJ 08544, United States of America

Bertschinger, H.G.

Euratom-KFA Association, Institut fur Plasmaphysik,
Kernforschungsanlage Jiilich GmbH,
P.O. Box 1913,
D-5170 Jiilich, Federal Republic of Germany

LIST OF PARTICIPANTS

549

Besshou, S.

Plasma Physics Laboratory, Kyoto University,
Gokasho, Uji,
Kyoto 611, Japan

Bhattacharjee, A.

Department of Applied Physics and Nuclear Engineering,
Columbia University,
New York, NY 10027, United States of America

Eiskamp, D.

Max-Planck-Institut fur Plasmaphysik,
Boltzmannstrasse 2,
D-8046 Garching, Federal Republic of Germany

Bobeldijk, C.

Division of Scientific and Technical Information,
International Atomic Energy Agency,
P.O. Box 100, A-1400 Vienna, Austria

Bodin, H.A.B.

United Kingdom Atomic Energy Authority,
Culham Laboratory,Abingdon, Oxfordshire, 0X14 3DB, United Kingdom

Bodner, S.E.

Naval Research Laboratory,
4555 Overlook Avenue, S.W.,
Washington, DC 20375, United States of America

Bol, K.

Princeton Plasma Physics Laboratory, Princeton University,
P.O. Box 451,
Princeton, NJ 08544, United States of America

Bolt, H.

Department of Nuclear Engineering, Osaka University,
2-1 Yamada-oka, Suita,
Osaka 565, Japan

Bolton, R.A.

Institut de recherche d'Hydro-Que'bec,
C.P. 1000,
Varennes, Quebec JOL 2P0, Canada

Borghi, C.A.

Faculta di Ingegneria, Universita di Bologna,
Viale Risorgimento 2,
1-40136 Bologna, Italy

Braams, C M .

FOM Instituut voor Plasmafysica 'Rynhuizen',
Edisonbaan 14, P.O. Box 1207,
NL-3430 BE Nieuwegein, Netherlands

Brennan, M.H.

Plasma Physics Department, School of Physics,
University of Sydney,
Sydney, NSW 2006, Australia

Breun, R.A.

Nuclear Engineering Department, University of Wisconsin,
Engineering Research Building, 1500 Johnson Drive,
Madison, WI 53706, United States of America

550

LIST OF PARTICIPANTS

Brunelli, B.

Associazione Euratom-ENEA,
Centro Ricerche Energia Frascati,
C.P. 65,
1-00044 Frascati, Rome, Italy

Brusati, M.

JET Joint Undertaking,
Abingdon, Oxfordshire OX14 3EA, United Kingdom

Buffa, A.

Associazione Euratom-ENEA, Istituto Gas Ionizzati,
Consiglio Nazionale delle Ricerche,
Via Gradenigo 6A,
1-35131 Padua, Italy

Bures, M.

JET Joint Undertaking,
Abingdon, Oxfordshire OX14 3EA, United Kingdom

Bussac, M.N.

Centre de physique the'orique, Ecole polytechnique,
F-91128 Palaiseau, France

Callen, J.D.

Nuclear Engineering Department, University of Wisconsin,
1500 Johnson Drive,
Madison, WI 53706, United States of America

Campbell, D.J.

JET Joint Undertaking,
Abingdon, Oxfordshire OX 14 3EA, United Kingdom

Canobbio, E.

Commission of the European Communities (DG XU/Fusion),
200, rue de la Loi,
B-1049 Brussels, Belgium

Carreras, B.A.

Fusion Energy Division, Oak Ridge National Laboratory,
P.O. Box Y,
Oak Ridge, TN 37831, United States of America

Chang, C.T.

Ris0 National Laboratory,
P.O. Box 49,
DK-4000 Roskilde, Denmark

Chatelier, M.

Commissariat a l'energie atomique (DRFC),
Centre d'&udes nucleaires de Fontenay-aux-Roses,
B.P. 6,F-92260, Fontenay-aux-Roses, France

Chen, L.

Princeton Plasma Physics Laboratory, Princeton University,
P.O. Box 451,
Princeton, NJ 08544, United States of America

Chen, Y.S.

Institute of Applied Physics and Computational Mathematics,
P.O. Box 8009,
Beijing, China

LIST OF PARTICIPANTS

Cho, T.

Plasma Research Centre, University of Tsukuba,
Sakura-mura, Niihari-gun,Ibaraki-ken 305, Japan

Choi, Duk-In

Institute for Fusion Studies, University of Texas,
Austin, TX 78731, United States of America

Chu, C.K.

Department of Applied Physics and Nuclear Engineering,
Columbia University,
500 West 120th Street,
New York, NY 10027, United States of America

Chu, Mingsheng

GA Technologies Inc.,
P.O. Box 85608,
San Diego, CA 92138, United States of America

Clarke, J.F.

Office of Energy Research,
United States Department of Energy,
Washington, DC 20545, United States of America

Coensgen, F.H.

Lawrence Livermore National Laboratory,
University of California,
P.O. Box 5511,
Livermore, CA 94550, United States of AmericaConn, R.W.

Center for Plasma Physics and Fusion Engineering,
University of California,
6291 Boelter Hall, 405 Hilgard Avenue,
Los Angeles, CA 90024, United States of America

Connor, J.W.

United Kingdom Atomic Energy Authority,
Culham Laboratory,Abingdon, Oxfordshire OX 14 3DB, United Kingdom

Conrads, J.P.F.

Euratom-KFA Association, Institut fur Plasmaphysik,
Kernforschungsanlage Julich GmbH,
Postfach 1913,
D-5170 Julich, Federal Republic of Germany

Coppi, B.

Tecnomasio Italiano Brown Boveri,
P. le Looli 3,1-20137 Milan, Italy

Coppi, B.

Massachusetts Institute of Technology,
77 Massachusetts Avenue,
Cambridge, MA 02139, United States of America

Cordey, J.G.

JET Joint Undertaking,
Abingdon, Oxfordshire OX14 3EA, United Kingdom

551

552

LIST OF PARTICIPANTS

Coutant, J.

Commissariat a I'energie atomique,Centre d'emdes de Limeil-Valenton,
B.P. 27,
F-94190 Villeneuve-Saint-Georges, France

Czekaj, S.

Institute of Plasma Physics and Laser Microfusion,
P.O. Box 49,
PL-00-908 Warsaw 49, Poland

Dahlburg, J.

Naval Research Laboratory,
4555 Overlook Avenue,
Washington, DC 20375-5000, United States of America

Daido, H.

Institute of Laser Engineering, Osaka University,
2-6 Yamada-oka, Suita,
Osaka 565, Japan

Daughney, C.C.

Division of Energy, National Research Council,
Montreal Road,
Ottawa, Ontario K1A 0R6, Canada

Dautovich, D.P.

Ontario Hydro,
2700 Lakeshore Road West,
Mississauga, Ontario L5J 1K3, Canada

Dautray, R.I.

Commissariat a l'energie atomique,Centre d'^tudes de Limeil-Valenton,
B.P. 27,
F-94190 Villeneuve-Saint-Georges, FranceDavidson, R.

Massachusetts Institute of Technology,
77 Massachusetts Avenue,
Cambridge, MA 02139, United States of America

Dawson, J.

Los Alamos National Laboratory,
P.O. Box 1663,
Los Alamos, NM 87545, United States of America

Dean, S.O.

Fusion Power Associates,
2 Professional Drive 249,
Gaithersburg, MD 20879, United States of America

Decroisette, M.

Commissariat a l'6nergie atomique,Centre d'ftudes de Limeil-Valenton,
B.P. 27,
F-94190 Villeneuve-Saint-Georges, France

De Grassie, J.S.

GA Technologies Inc.,
P.O. Box 85608,
San Diego, CA 92138, United States of America

LIST OF PARTICIPANTS

De Marco, F.

Associazione Euratom-ENEA,
Centro Ricerche Energia Frascati,
C.P. 65,
1-00044 Frascati, Rome, Italy

Denus, S.

Institute of Plasma Physics and Laser Microfusion,
P.O. Box 49,
PL-OO-908 Warsaw, Poland

Diamond, P.H.

Institute for Fusion Studies,
The University of Texas at Austin,
Austin, TX 78712, United States of America

Dietz, K.J.

JET Joint Undertaking,
Abingdon, Oxfordshire OX14 3EA, United Kingdom

DiMarco, J.N.

Los Alamos National Laboratory,
P.O. Box 1663,
Los Alamos, NM 87545, United States of America

Dippel, K.H.

Euratom-KFA Association, Institut fur Plasmaphysik,
Kernforschungsanlage Jiilich GmbH,
Postfach 1913,
D-5170 Julich, Federal Republic of Germany

Dos Santos, R.

Centro de Informacoes Nucleares,
Rua General Severiano 90, Botafogo,
22.294 Rio de Janeiro, RJ, Brazil

Drake, J.F.

Laboratory for Plasma and Fusion Energy Studies,
University of Maryland,College Park, MD 20742, United States of America

Dubois, M.A.

Commissariat a l'energie atomique (DRFC),
Centre d'&udes nucleaires de Fontenay-aux-Roses,
B.P. 6,F-92260 Fontenay-aux-Roses, France

Dilchs, D.F.

JET Joint Undertaking,
Abingdon, Oxfordshire OX14 3EA, United Kingdom

Duesing, G.

JET Joint Undertaking,
Abingdon, Oxfordshire OX14 3EA, United Kingdom

Edenstrasser, J.W.

Institut fur Theoretische Physik, Universitat Innsbruck,
Technikerstrasse 15,
A-6020 Innsbruck, Austria

Edwards, A.

JET Joint Undertaking,
Abingdon, Oxfordshire 0X14 3EA, United Kingdom

553

554

LIST OF PARTICIPANTS

Eggleston, D.L.

Physics Department, University of California,
La Jolla, CA 92093, United States of America

El-Khalafawy, T.A.

Plasma Physics and Accelerators Department,
Nuclear Research Centre, Atomic Energy Authority,13759 Cairo, Egypt

Eliezer, S.

Plasma Physics Department,
Soreq Nuclear Research Center,
IL-70600, Yavne, Israel

Engelhardt, W.

JET Joint Undertaking,
Abingdon, Oxfordshire OX14 3EA, United Kingdom

Engelmann, F.

NET Team, Max-Planck-Institut fur Plasmaphysik,
Boltzmannstrasse 2,
D-8046 Garching, Federal Republic of Germany

Eninger, J.E.

Department of Plasma Physics and Fusion Research,
Royal Institute of Technology,
S-10044 Stockholm 70, Sweden

Enoto, T.

Faculty of Engineering, Hokkaido University,
Nishi-8, Kita-13, Kita-ku,
Sapporo 060, Japan

Erckmann, V.

Max-Planck-Institut fur Plasmaphysik,
Boltzmannstrasse 2,
D-8046 Garching, Federal Republic of Germany

Esipchuk, Yu.V.

I.V. Kurchatov Institute of Atomic Enery,
42 Ulitsa Kurchatova, P.O. Box 3402,
123182 Moscow, Union of Soviet Socialist Republics

Fabre, E.

GRECO Interaction Laser-Matiere,
Laboratoire de physique des milieux ionises,
Ecole polytechnique,
F-91128 Palaiseau Cedex, France

Ferronsky, V.I.

Division of Research and Laboratories,
International Atomic Energy Agency,
P.O. Box 100, A-1400 Vienna, Austria

Finn, J.M.

Laboratory for Plasma and Fusion Energy Studies,
University of Maryland,College Park, MD 20742, United States of America

Fontanesi, M.

Associazione Euratom-ENEA, Istituto di Fisica del Plasma,
Consiglio Nazionale delle Ricerche,
Via Bassini 15,
1-20133 Milan, Italy

LIST OF PARTICIPANTS

Fowler, T.K.

Lawrence Livermore National Laboratory,
University of California,
P.O. Box 5511,
Livermore, CA 94550, United States of America

Fujii, M.

Nuclear Systems Division,
Kawasaki Heavy Industries Ltd,
4-25-2 Minamisuna-cho, Koto-ku,
Tokyo 136, Japan

Fujii, T.

Division of Thermonuclear Fusion Research,
Japan Atomic Energy Research Institute,Naka-machi, Naka-gun,Ibaraki-ken 311-02, Japan

Fujikura, T.

Power and Fusion Technology Development Department,
Toshiba Corporation,
1-6-1 Uchisaiwai-cho, Chiyoda-ku,
Tokyo 100, Japan

Fujisawa, N.

Naka Fusion Energy Research Establishment,
Japan Atomic Energy Research Institute,801-1 Mukoyama, Naka-machi, Naka-gun,Ibaraki-ken 311-02, Japan

Fujita, H.

Institute of Laser Engineering, Osaka University,
2-6 Yamada-oka, Suita,
Osaka 565, Japan

Fujita, J.

Institute of Plasma Physics, Nagoya University,
Furo-cho, Chikusa-ku,
Nagoya 464, Japan

Fujita, T.

Department of Nuclear Engineering, University of Tokyo,
7-3-1 Hongo, Bunkyo-ku,
Tokyo 113, Japan

Fujiwara, M.

Institute of Plasma Physics, Nagoya University,
Furo-cho, Chikusa-ku,
Nagoya 464, Japan

Fukai, Y.

Nuclear Energy Group, Toshiba Corporation,
1-6-1 Uchisaiwai-cho, Chiyuda-ku,
Tokyo 100, Japan

Fukuyama, A.

Department of Electronics, Okayama University,
3-1-1 Tsushima-naka, Okayama 700, Japan

555

556

LIST OF PARTICIPANTS

Funahashi, A.

Experimental Plasma Physics Laboratory,
Department of Thermonuclear Fusion Research,
Japan Atomic Energy Research Institute,
Tokai-mura, Naka-gun,Ibaraki-ken 319-11, Japan

Furth, H.P.

Princeton Plasma Physics Laboratory, Princeton University,
P.O. Box 451,
Princeton, NJ 08544, United States of America

Garner, H.

GA Technologies Inc.,
P.O. Box 856X18,
San Diego, CA 92138, United States of America

Geiger, J.S.

Chalk River Nuclear Laboratories,
Atomic Energy of Canada Ltd,
Chalk River, Ontario KOJ 1J0, Canada

Ghorannevis, M.

Plasma Physics Group, Nuclear Research Centre,
Atomic Energy Organization,
P.O. Box 11365-8486,
Tehran, Iran

Gibson, A.

JET Joint Undertaking,
Abingdon, Oxfordshire 0X14 3EA, United Kingdom

Gierke, G. von

Max-Planck-Institut fur Plasmaphysik,
Boltzmannstrasse 2,
D-8046 Garching, Federal Republic of Germany

Gilleland, J.R.

GA Technologies Inc.,
P.O. Box 85608,
San Diego, CA 92138, United States of America

Goldston, R.J.

Princeton Plasma Physics Laboratory, Princeton University,
P.O. Box 451,
Princeton, NJ 08544, United States of America

Goldsworthy, M.

University of New South Wales,
P.O. Box 1,
Kensington 2033, Australia

Gondhalekar, A.

JET Joint Undertaking,
Abingdon, Oxfordshire OX14 3EA, United Kingdom

Goto, S.

Plasma Physics Laboratory, Osaka University,
2-6 Yamada-oka, Suita,
Osaka 565, Japan

LIST OF PARTICIPANTS

557

Goto, Y.

Department of Electrical Engineering, Tohoku University,
Aoba, Aramaki,
Sendai 980, Japan

Gottlieb, H.B.

Grumman Aerospace Corporation,
14 Mershon Drive,
Princeton, NJ 08540, United States of America

Greenwald, M.

Plasma Fusion Center, Massachusetts Institute of Technology,
167 Albany Street NW 16-274,
Cambridge, MA 02139, United States of America

Grieger, G.

Max-Planck-Institut fur Plasmaphysik,
Boltzmannstrasse 2,
D-8046 Garching, Federal Republic of Germany

Gross, R.A.

School of Engineering and Applied Science,
Columbia University,
New York, NY 10027, United States of America

Grove, D.J.

Princeton Plasma Physics Laboratory, Princeton University,
P.O. Box 451,
Princeton, NJ 08544, United States of America

Gruber, O.

Max-Planck-Institut fur Plasmaphysik,
Boltzmannstrasse 2,
D-8046 Garching, Federal Republic of Germany

Guasp, J.

Divisi6n de Fusi6n, Junta de Energia Nuclear,
Avda. Complutense 22,
E-28040 Madrid, Spain

Guo, Wenkang

Institute of Plasma Physics, Academia Sinica,
P.O. Box 26,
Hefei, Anhui, China

Haines, M.G.

Centre for Fusion Studies, Blackett Laboratory,
Imperial College of Science and Technology,
Prince Consort Road,
London SW7 2BZ, United Kingdom

Hamada, S.

College of Science and Technology,
Department of Physics, Nihon University,
1-8 Kanda-Surugadai, Chiyoda-ku,
Tokyo 101, Japan

Hamada, S.

Department of Electronics, Kyoto University,
Kitashirakawa, Sakyo-ku,
Kyoto 606, Japan558

LIST OF PARTICIPANTS

Hamada, Y.

Institute of Plasma Physics, Nagoya University,
Furo-cho, Chikusa-ku,
Nagoya 464, Japan

Hamamatsu, K.

Naka Fusion Energy Research Establishment,
Japan Atomic Energy Research Institute,801-1 Mukoyama, Naka-machi, Naka-gun,Ibaraki-ken 311-02, Japan

Hamasaki, S.

JAYCOR,
11011 Torreyana Road, P.O. Box 85154,
San Diego, CA 92138, United States of America

Hamberger, S.M.

Research School of Physical Sciences,
Institute of Advanced Studies, Australian National University,
P.O. Box 4,
Canberra 2601 AC, Australia

Hanatani, K.

Plasma Physics Laboratory, Kyoto University,
Gokasho, Uji,
Kyoto 611, Japan

Hancox, R.

Euratom-UKAEA Fusion Association,
Culham Laboratory,Abingdon, Oxfordshire 0X14 3DB, United Kingdom

Harafuji, K.

Institute for Fusion Theory, Hiroshima University,
1-1-89 Higashisenda-machi, Naka-ku,
Hiroshima 730, Japan

Hasegawa, A.

AT&T Bell Laboratories,
Room 1E-351,
Murray Hill, NJ 07974, United States of America

Hashiguchi, R.

Faculty of Engineering, Science University of Tokyo,
1-3 Kagurazaka, Shinjuku-ku,
Tokyo 162, Japan

Hatakeyama, R.

Department of Electronic Engineering, Tohoku University,
Aoba, Aramaki,
Sendai 980, Japan

Hattori, K.

Department of Nuclear Engineering, University of Tokyo,
7-3-1 Hongo, Bunkyo-ku,
Tokyo 113, Japan

Hawryluk, R.J.

Princeton Plasma Physics Laboratory, Princeton University,
P.O. Box 451,
Princeton, NJ 08544, United States of America

LIST OF PARTICIPANTS

559

Hayase, K.

Electrotechnical Laboratory,
1-1-4 Umezono, Sakura-mura, Niihari-gun,Ibaraki-ken 305, Japan

Hayashi, T.

Institute for Fusion Theory, Hiroshima University,
1-1-89 Higashisenda-machi, Naka-ku,
Hiroshima 730, Japan

Heindler, M.

Institut fiir Theoretische Physik, Technische Universitat Graz,
Petersgasse 16,
A-8010 Graz, Austria

Hender, T.C.

United Kingdom Atomic Energy Authority,
Culham Laboratory,Abingdon, Oxfordshire 0X14 3DB, United Kingdom

Hershkowitz, N.

Nuclear Engineering Department,
University of Wisconsin-Madison,1500 Johnson Drive,
Madison, WI 53706, United States of America

Hill, K.W.

Princeton Plasma Physics Laboratory, Princeton University,
P.O. Box 451,
Princeton, NJ 08544, United States of America

Hintz, E.

Euratom-KFA Association, Institut fur Plasmaphysik,
Kernforschungsanlage Ju'lich GmbH,
Postfach 1913,
D-5170 Jiilich, Federal Republic of Germany

Hiraki, N.

Research Institute for Applied Mechanics,
Kyushu University,
87 Kasauga,
Fukuoka 816, Japan

Hirano, K.

Department of Electronic Engineering, Gunma University,
1-5-1 Tenjin-cho, Kiryu,
Gunma 376, Japan

Hirano, Y.

Electrotechnical Laboratory,
1-1-4 Umezono, Sakura-mura, Niihari-gun,Ibaraki-ken 305, Japan

Hirayama, T.

Naka Fusion Energy Research Establishment,
Japan Atomic Energy Research Institute,801-1 Mukoyama, Naka-machi, Naka-gun,Ibaraki-ken 311-02, Japan

Hirose, A.

Department of Physics, University of Saskatchewan,
Saskatoon, Saskatchewan S7N 0W0, Canada

560

LIST OF PARTICIPANTS

Hirota, I.

Electrotechnical Laboratory,
1-1-4 Umezono, Sakura-mura, Niihari-gun,Ibaraki-ken 305, Japan

Hoffman, A.L.

Spectra Technology, Inc.,
2755 Northup Way,
Bellevue, WA 98004, United States of America

Honda, Y.

Department of Nuclear Engineering, Osaka University,
2-1 Yamada-oka, Suita,
Osaka 565, Japan

Honma, T.

Department of Electrical Engineering, Hokkaido University,
Nishi-8, Kita-13, Kita-ku,
Sapporo 060, Japan

Hora, H.

University of New South Wales,
P.O. Box 1,
Kensington 2033, Australia

Horiike, H.

Naka Fusion Energy Research Establishment,
Japan Atomic Energy Research Institute,801-1 Mukoyama, Naka-machi, Naka-gun,Ibaraki-ken 3.11-02, Japan

Horioka, K.

Department of Energy Sciences,
Tokyo Institute of Technology,
4259 Nagatsuta-cho, Midori-ku,
Yokohama 227, Japan

Horiuchi, R.

Science Project Corporation,
3 Kuchita-minami, Asakita-ku,
Hiroshima 730, Japan

Hosea, J.C.

Princeton Plasma Physics Laboratory, Princeton University,
P.O. Box 451,
Princeton, NJ 08544, United States of America

Hoshino, K.

Experimental Plasma Physics Laboratory,
Department of Thermonuclear Fusion Research,
Japan Atomic Energy Research Institute,
Tokai-mura, Naka-gun,Ibaraki-ken 319-11, Japan

Hosogane, N.

Naka Fusion Energy Research Establishment,
Japan Atomic Energy Research Institute,801-1 Mukoyama, Naka-machi, Naka-gun,Ibaraki-ken 311-02, Japan

LIST OF PARTICIPANTS
Hugill, J.

United Kingdom Atomic Energy Authority,
Culham Laboratory,Abingdon, Oxfordshire 0X14 3DB, United Kingdom

Huo, Yuping

Institute of Plasma Physics, Academia Sinica,
P.O. Box 26,
Hefei, Anhui, China

Hushimi, K.

State Councillor,2-803 Ojkura-yama-haimu, Futoh-cho, Kohoku-ku,
Yokohama 227, Japan

Ichac, J.P.

Thomson-CSF, Division tubes e"Iectroniques,
38, rue Vauthier, B.P. 305,
F-92102 Boulogne-Billancourt Cedex, France

Ichiguchi, K.

Plasma Physics Laboratory, Kyoto University,
Gokasho, Uji,
Kyoto 611, Japan

Ichikawa, Y.H.

Institute of Plasma Physics, Nagoya University,
Furo-cho, Chikusa-ku,
Nagoya 464, Japan

Ichimura, M.

Plasma Research Centre, University of Tsukuba,
Sakura-mura, Niihari-gun,Ibaraki-ken 305, Japan

Ida, K.

Institute of Plasma Physics, Nagoya University,
Furo-cho, Chikusa-ku,
Nagoya 464, Japan

Ide, S.

Department of Physics, Kyoto University,
Kitashirakawa, Sakyo-ku,
Kyoto 606, Japan

Idehara, T.

Department of Applied Physics, University of Fukui,
Fukui 910, Japan

Ido, S.

Institute of Laser Engineering, Osaka University,
2-6 Yamada-oka, Suita,
Osaka 565, Japan

Iguchi, H.

Institute of Plasma Physics, Nagoya University,
Furo-cho, Chikusa-ku,
Nagoya 464, Japan

Iida, M.

Department of Physics, Kyoto University,
Kitashirakawa, Sakyo-ku,
Kyoto 606, Japan

561

562

LIST OF PARTICIPANTS

Iijima, T.

Naka Fusion Energy Research Establishment,
Japan Atomic Energy Research Institute,801-1 Mukoyama, Naka-machi, Naka-gun,Ibaraki-ken 311-02, Japanlima, S.

Plasma Physics Laboratory, Kyoto University,
Gokasho, Uji,
Kyoto 611, Japan

Iiyoshi, A.

Plasma Physics Laboratory, Kyoto University,
Gokasho, Uji,
Kyoto 611, Japan

Ijiri, Y.

Plasma Physics Laboratory, Kyoto University,
Gokasho, Uji,
Kyoto 611, Japan

Ikeda, N.

Electrotechnical Laboratory,
1-1-4 Umezono, Sakura-mura, Niihari-gun,Ibaraki-ken 305, Japan

Ikegami, H.

Institute of Plasma Physics, Nagoya University,
Furo-cho, Chikusa-ku,
Nagoya 464, Japan

Imai, T.

Department of Thermonuclear Fusion Research,
Japan Atomic Energy Research Institute,801-1 Mukoyama, Naka-machi, Naka-gun,Ibaraki-ken 311-02, Japan

Imasaki, K.

Institute of Laser Engineering, Osaka University,
2-6 Yamada-oka, Suita,
Osaka 565, Japan

Imazu, S.

Kanazawa Institute of Technology,
Nonoichi, Kanazawa-South,
Ishikawa 921, Japan

Inoue, K.

Naka Fusion Energy Research Establishment,
Japan Atomic Energy Research Institute,801-1 Mukoyama, Naka-machi, Naka-gun,Ibaraki-ken 311-02, Japan

Inoue, N.

Department of Nuclear Engineering, University of Tokyo,
7-3-1 Hongo, Bunkyo-ku,
Tokyo 113, Japan

Inoue, S.

Department of Electronics, Kyoto University,
Kitashirakawa, Sakyo-ku,
Kyoto 606, Japan

LIST OF PARTICIPANTS

563

Inutake, M.

Plasma Research Centre, University of Tsukuba,
Sakura-mura, Niihari-gun,Ibaraki-ken 305, Japan

Ioki, K.

Mitsubishi Atomic Power Industries,
4-1-2 Shibakouen-cho, Minato-ku,
Tokyo 105, Japan

Ishihara, O.

Department of Electrical Engineering/Computer Science,
Texas Tech University,
Lubbock, TX 79409^1439, United States of America

Ishii, K.

Plasma Research Centre, University of Tsukuba,
Sakura-mura, Niihari-gun,Ibaraki-ken 305, Japan

Ishii, S.

Tokyo Institute of Technology,
2-12-1 O-okayama, Meguro-ku,
Tokyo 152, Japan

Ishimura, T.

Plasma Physics Laboratory, Osaka University,
2-6 Yamada-oka, Suita,
Osaka 565, Japan

Itakura, A.

Plasma Research Centre, University of Tsukuba,
Sakura-mura, Niihari-gun,Ibaraki-ken 305, Japan

Itami, K.

Department of Nuclear Engineering, University of Tokyo,
7-3-1 Hongo, Bunkyo-ku,
Tokyo 113, Japan

Itatani, R.

Department of Electronics, Kyoto University,
Kitashirakawa, Sakyo-ku,
Kyoto 606, Japan

Ito, H.

Plasma Physics Laboratory, Osaka University,
2-6 Yamada-oka, Suita,
Osaka 565, Japan

Ito, Y.

Plasma Physics Laboratory, Osaka University,
2-6 Yamada-oka, Suita,
Osaka 565, Japan

Itoh, K.

Plasma Physics Laboratory, Kyoto University,
Gokasho, Uji,
Kyoto 611, Japan

Itoh, S.

Research Institute for Applied Mechanics, Kyushu University,
87 Kasuga,
Fukuoka 816, Japan

564

LIST OF PARTICIPANTS

Itoh, S.-I.

Institute for Fusion Theory, Hiroshima University,
1-1-89 Higashisenda-machi, Naka-ku,
Hiroshima 730, Japan

Iwamura, K.

Department of Physics, Kyoto University,
Kitashirakawa, Sakyo-ku,
Kyoto 606, Japan

Izawa, Y.

Institute of Laser Engineering, Osaka University,
2-6 Yamada-oka, Suita,
Osaka 565, Japan

Jackson, D.P.

Atomic Energy of Canada Ltd,
Chalk River Nuclear Laboratories,
Chalk River, Ontario KOJ 1J0, Canada

Jacquinot, J.

JET Joint Undertaking,
Abingdon, Oxfordshire OX14 3EA, United Kingdom

Jarvis, O.N.

JET Joint Undertaking,
Abingdon, Oxfordshire OX14 3EA, United Kingdom

Jensen, V.O.

Rise National Laboratory,
P.O. Box 49,
DK-4000 Roskilde, Denmark

Jimbo, K.

Institute of Atomic Energy, Kyoto University,
Gokasho, Uji,
Kyoto 611, Japan

Johnson, P.C.

United Kingdom Atomic Energy Authority,
Culham Laboratory,Abingdon, Oxfordshire 0X14 3DB, United Kingdom

Kadomtsev, B.B.

I.V. Kurchatov Institute of Atomic Energy,
42 Ulitsa Kurchatova, P.O. Box 3402,
123182 Moscow, Union of Soviet Socialist Republics

Kako, M.

Plasma Physics Laboratory, Osaka University,
2-6 Yamada-oka, Suita,
Osaka 565, Japan

Kamada, K.

Institute of Plasma Physics, Nagoya University,
Furo-cho, Chikusa-ku,
Nagoya 464, Japan

Kamada, Y.

Department of Nuclear Engineering, University of Tokyo,
7-3-1 Hongo, Bunkyo-ku,
Tokyo 113, Japan

LIST OF PARTICIPANTS

565

Kameari, A.

Mitsubishi Atomic Power Industries,
4-1-2 Shibakouen-cho, Minato-ku,
Tokyo 105,Japan

Kamimura, T.

Institute of Plasma Physics, Nagoya University,
Furo-cho, Chikusa-ku,
Nagoya 464, Japan

Kamitani, A.

Department of Applied Physics, University of Tokyo,
7-3-1 Hongo, Bunkyo-ku,
Tokyo 113, Japan

Kaneko, H.

Plasma Physics Laboratory, Kyoto University,
Gokasho, Uji,
Kyoto 611, Japan

Kaneko, S.

Department of Applied Physics, University of Tokyo,
7-3-1 Hongo, Bunkyo-ku,
Tokyo 113, Japan

Karger, F.

Max-Planck-Institut fur Plasmaphysik,
Boltzmannstrasse 2,
D-8046 Garching, Federal Republic of Germany

Kasahara, T.

Energy Conversion Research Institute, Hokkaido University,
Nishi-8, Kita-13, Kita-ku,
Sapporo 060, Japan

Kasai, M.

Mitsubishi Atomic Power Industries,
4-1-2 Shibakouen-cho, Minato-ku,
Tokyo 105, Japan

Kasai, S.

Experimental Plasma Physics Laboratory,
Department of Thermonuclear Fusion Research,
Japan Atomic Energy Research Institute,
Tokai-mura, Naka-gun,Ibaraki-ken 319-11, Japan

Kasuya, K.

Department of Energy Sciences,
Tokyo Institute of Technology,
4259 Nagatsuta-cho, Midori-ku,
Yokohama 227, Japan

Katanuma, I.

Plasma Research Centre, University of Tsukuba,
Sakura-mura, Niihari-gun,Ibaraki-ken 305, Japan

Kato, Y.

Department of Nuclear Engineering, Osaka University,
2-1 Yamada-oka, Suita,
Osaka 565, Japan

566

LIST OF PARTICIPANTS

Kato, Y.

Institute of Laser Engineering, Osaka University,
2-6 Yamada-oka, Suita,
Osaka 565, Japan

Kaufmann, M.

Max-Planck-Institut fur Plasmaphysik,
Boltzmannstrasse 2,
D-8046 Garching, Federal Republic of Germany

Kaw, P.K.

Plasma Physics Programme, Physical Research Laboratory,
Navrangpura, Ahmedabad 380009, India

Kawabata, N.

Plasma Physics Laboratory, Kyoto University,
Gokasho, Uji,
Kyoto 611, Japan

Kawabe.T.

Plasma Research Centre, University of Tsukuba,
Sakura-mura, Niihari-gun,Ibaraki-ken 305, Japan

Kawahata, K.

Institute of Plasma Physics, Nagoya University,
Furo-cho, Chikusa-ku,
Nagoya 464, Japan

Kawai, Y.

Interdisciplinary Graduate School of Engineering Sciences,
Kyushu University,
6-1 Kasuga Koen, Kasuga,
Fukuoka 816, Japan

Kawakami, I.

Atomic Energy Research Institute, Nihon University,
1-8 Kanda-Surugadai, Chiyoda-ku,
Tokyo 101, Japan

Kawamura, T.

Institute of Plasma Physics, Nagoya University,
Furo-cho, Chikusa-ku,
Nagoya 464, Japan

Keen, B.E.

JET Joint Undertaking,
Abingdon, Oxfordshire OX14 3EA, United Kingdom

Keilhacker, M.

JET Joint Undertaking,
Abingdon, Oxfordshire OX 14 3EA; United Kingdom

Kernbichler, W.

Institut fur Theoretische Physik, Technische Universitat Graz,
Petersgasse 16,
A-8010 Graz, Austria

Kikuchi, H.

College of Science and Technology, Nihon University,
1-8 Kanda Surugadai, Chiyoda-ku,
Tokyo 101, Japan

LIST OF PARTICIPANTS

Killeen, J.

National MFE Computer Center,
Lawrence Livermore National Laboratory,
University of California,
P.O. Box 5509,
Livermore, CA 94550, United States of America

Kimura, H.

Naka Fusion Energy Research Establishment,
Japan Atomic Energy Research Institute,801-1 Mukoyama, Naka-machi, Naka-gun,Ibaraki-ken 311-02, Japan

Kimura, T.

Naka Fusion Energy Research Establishment,
Japan Atomic Energy Research Institute,801-1 Mukoyama, Naka-machi, Naka-gun,Ibaraki-ken 311-02, Japan

Kishimoto, H.

Naka Fusion Energy Research Establishment,
Japan Atomic Energy Research Institute,801-1 Mukoyama, Naka-machi, Naka-gun,Ibaraki-ken 311-02, Japan

Kishimoto, Y.

Tokai Research Establishment,
Japan Atomic Energy Research Institute,
Tokai-mura, Naka-gun,Ibaraki-ken 319-11, Japan

Kitagawa, S.

Institute of Plasma Physics, Nagoya University,
Furo-cho, Chikusa-ku,
Nagoya 464, Japan

Kitagawa, Y.

Institute of Laser Engineering, Osaka University,
2-6 Yamada-oka, Suita,
Osaka 565, Japan

Kito, M.

Electrotechnical Laboratory,
1-1-4 Umezono, Sakura-mura, Niihari-gun,Ibaraki-ken 305, Japan

Kitsunezaki, A.

Naka Fusion Energy Research Establishment,
Japan Atomic Energy Research Institute,801-1 Mukoyama, Naka-machi, Naka-gun,Ibaraki-ken 311-02, Japan

Kawamoto, Y.

Institute of Applied Physics,
Plasma Research Centre, University of Tsukuba,
Sakura-mura, Niihari-gun,Ibaraki-ken 305, Japan

Kiyama, H.

Electrotechnical Laboratory,
1-1-4 Umezono, Sakura-mura, Niihari-gun,Ibaraki-ken 305, Japan

567

568

LIST OF PARTICIPANTS

Kiyama, S.

Electrotechnical Laboratory,
1-1-4 Umezono, Sakura-mura, Niihari-gun,Ibaraki-ken 305, Japan

Kobayashi, S.

Plasma Physics Laboratory, Kyoto University,
Gokasho, Uji,
Kyoto 611, JapanKoch, R.

Ecole Royale Militaire,
30, avenue de la Renaissance,
B-1040 Brussels, Belgium

Kogoshi, S.

Department of Electrical Engineering, University of Tokyo,
7-3-1 Hongo, Bunkyo-ku,
Tokyo 113, Japan

Kojevin, B.M.

State Committee for the Utilization of Atomic Energy,
Staromonetnyj Pereulok 26,
Moscow Zh-180, Union of Soviet Socialist Republics

Komarek, P.

Komori, A.

Kondo, I.

Institut fur Technische Physik,
Kernforschungszentrum Karlsruhe GmbH,
Postfach 3640,
D-7500 Karlsruhe 1, Federal Republic of Germany
Interdisciplinary Graduate School of Engineering Sciences,
Kyushu University,
6-1 Kasuga Koen, Kasuga,
Fukuoka 816, Japan
Naka Fusion Energy Research Establishment,
Japan Atomic Energy Research Institute,801-1 Mukoyama, Naka-machi, Naka-gun,Ibaraki-ken 311-02, Japan

Kondo, K.

Plasma Physics Laboratory, Kyoto University,
Gokasho, Uji,
Kyoto 611, Japan

Konoshima, S.

Naka Fusion Energy Research Establishment,
Japan Atomic Energy Research Institute,801-1 Mukoyama, Naka-machi, Naka-gun,Ibaraki-ken 311-02, Japan

Koppendorfer, W.

Max-Planck-Institut fur Plasmaphysik,
Boltzmannstrasse 2,
D-8046 Garching, Federal Republic of Germany

Kotschenreuther, M.

Institute for Fusion Studies,
The University of Texas at Austin,
Austin, TX 78712, United States of America

LIST OF PARTICIPANTS
Kovrizhnykh, L.M.

Institute of General Physics,
Academy of Sciences of the USSR,
Ulitsa Vavilova 38, 117924, GSP,
Moscow V-333, Union of Soviet Socialist Republics

Koyama, K.

Electrotechnical Laboratory,
1-1-4 Umezono, Sakuia-mura, Niihari-gun,Ibaraki-ken 305, Japan

Kozhevin, V.M.

D.V. Efremov Scientific Research Instituteof Electrophysical Apparatus,
Leningrad, Union of Soviet Socialist Republics

Krall, N.A.

JAYCOR,
11011 Torreyana Road, P.O. Box 85154,
San Diego, CA 92138, United States of America

Krasheninnikov, S.I.

I.V. Kurchatov Institute of Atomic Energy,
42 Ulitsa Kurchatova, P.O. Box 3402,
123182 Moscow, Union of Soviet Socialist Republics

Kubo, S.

Institute of Plasma Physics, Nagoya University,
Furo-cho, Chikusa-ku,
Nagoya 464, Japan

Kumazawa, R.

Institute of Plasma Physics, Nagoya University,
Furo-cho, Chikusa-ku,
Nagoya 464, Japan

Kurasawa, T.

Department of Physics, Tokyo Institute of Technology,
2-12-1 O-okayama, Meguro-ku,
Tokyo 152, Japan

Kurita, G.

Department of Thermonuclear Fusion Research,
Tokai Research Establishment,
Japan Atomic Energy Research Institute,
Tokai-mura, Naka-gun,Ibaraki-ken 319-11, Japan

Kuriyama, M.

Naka Fusion Energy Research Establishment,
Japan Atomic Energy Research Institute,801-1 Mukoyama, Naka-machi, Naka-gun,Ibaraki-ken 311-02, Japan

Kusano, K.

Department of Physics, Hiroshima University,
1-1-89 Higashisenda-machi, Naka-ku,
Hiroshima 730, Japan

Kuwahara, H.

Plasma Physics Laboratory, University of Saskatchewan,
Saskatoon, Saskatchewan S7N 0W0, Canada

569

570

LIST OF PARTICIPANTS

Lackner, K.

Max-Planck-Institut fiiir Plasmaphysik,
Boltzmannstrasse 2,
D-8046 Garching, Federal Republic of Germany

Lee, E.P.

Lawrence Berkeley Laboratory, University of California,
1 Cyclotron Road,
Berkeley, CA 94720, United States of America

Lehnert, B.P.

Department of Plasma Physics and Fusion Research,
Royal Institute of Technology,
S-10044 Stockholm 70, Sweden

Leuterer, F.

Max-Planck-Institut fur Plasmaphysik,
Boltzmannstrasse 2,
D-8046 Garching, Federal Republic of Germany

Li, Zhengwu

Southwestern Institute of Physics,
P.O. Box 15,
Leshan, Sichuan, China

Linford, R.K.

Los Alamos National Laboratory,
P.O. Box 1663,
Los Alamos, NM 87545, United States of America

Lister, J.B.

Centre de recherches en physique des plasmas,
Ecole polytechnique fe'de'rale de Lausanne,
21, avenue des Bains,
CH-1007 Lausanne, Switzerland

Logan, B.G.

Lawrence Livermore National Laboratory,
University of California,
P.O. Box 5511,
Livermore, CA 94550, United States of America

Lomer, W.M.

United Kingdom Atomic Energy Authority,
Culham Laboratory,Abingdon, Oxfordshire OX 14 3DB, United Kingdom

Lubell, M.S.

Oak Ridge National Laboratory,
P.O. Box Y,
Oak Ridge, TN 37831, United States of America

Luhmann, N.C., Jr.

Electrical Engineering Department,
University of California,
7535 Boelter Hall,
Los Angeles, CA 90024, United States of America

Luxon, J.

GA Technologies Inc.,
P.O. Box 85608,
San Diego, CA 92138, United States of America

LIST OF PARTICIPANTS

571

Maassberg, H.

Max-Planck-Institut fur Plasmaphysik,
Boltzmannstrasse 2,
D-8046 Garching, Federal Republic of Germany

Machida, M.

Instituto de Fi'sica "Gleb Wataghin",
Universidade Estadual de Campinas,
CP 1170 Unicamp,
13100 Campinas, SP, Brazil

Maeda, H.

Tokai Research Establishment,
Japan Atomic Energy Research Institute,
Tokai-mura, Naka-gun,Ibaraki-ken 319-11, Japan

Maekawa, T.

Department of Physics, Kyoto University,
Kitashirakawa, Sakyo-ku,
Kyoto 606, Japan

Maeno, K.

Naka Fusion Energy Research Establishment,
Japan Atomic Energy Research Institute,801-1 Mukoyama, Naka-machi, Naka-gun,Ibaraki-ken 311-02, Japan

Maisonnier, Ch.

Commission of the European Communities (DG XII/Fusion),
200, rue de la Loi,
B-1049 Brussels, Belgium

Malesani, G.

Associazione Euratom-ENEA, Istituto Gas Ionizzati,
Consiglio Nazionale delle Ricerche,
Via Gradenigo 6A,
1-35131 Padua, Italy

Masamune, S.

Department of Electrical Engineering,
Kyoto Institute of Technology,
Matsugasaki, Sakyo-ku,
Kyoto 606, Japan

Mase, A.

Plasma Research Centre, University of Tsukuba,
Sakura-mura, Niihari-gun,Ibaraki-ken 305, Japan

Matoba, T.

Experimental Plasma Physics Laboratory,
Department of Thermonuclear Fusion Research,
Japan Atomic Energy Research Institute,
Tokai-mura, Naka-gun,Ibaraki-ken 319-11, Japan

Matsuda, S.

Naka Fusion Energy Research Establishment,
Japan Atomic Energy Research Institute,801-1 Mukoyama, Naka-machi, Naka-gun,Ibaraki-ken 311-02, Japan

572

LIST OF PARTICIPANTS

Matsumoto, H.

Experimental Plasma Physics Laboratory,
Department of Thermonuclear Fusion Research,
Japan Atomic Energy Research Institute,
Tokai-mura, Naka-gun,Ibaraki-ken 319-11, Japan

Matsumoto, M.

Faculty of Engineering, Yamanashi University,
4-3-11 Takeda, Kofu,
Yamanashi 400, Japan

Matsuoka, A.

Department of Electrical Engineering, Gunma University,
1-5-1 Tenjin-cho, Kiryu,
Gunma 376, Japan

Matsuoka, K.

Institute of Plasma Physics, Nagoya University,
Furo-cho, Chikusa-ku,
Nagoya 464, Japan

Matsuura, K.

Institute of Plasma Physics, Nagoya University,
Furo-cho, Chikusa-ku,
Nagoya 464, Japan

Matsuzaki, Y.

Facility Operation and Engineering Division,
Department of Thermonuclear Fusion Research,
Japan Atomic Energy Research Institute,
Tokai-mura, Naka-gun,Ibaraki-ken 319-11, Japan

Mattoo, S.K.

Physical Research Laboratory,
Navrangpura, Ahmedabad 380009, India

McCarthy, A.L.

School of Physical Sciences,
The Flinders University of South Australia,
Bedford Park, South Australia 5042, Australia

McCrory, R.L.

Laboratory for Laser Energetics, University of Rochester,
250 East River Road,
Rochester, NY 14623, United States of America

McGuire, K.

Princeton Plasma Physics Laboratory, Princeton University,
P.O. Box 451,
Princeton, NJ 08544, United States of America

Meade, D.M.

Princeton Plasma Physics Laboratory, Princeton University,
P.O. Box 451,
Princeton, NJ 08544, United States of America

Mieno, T.

Department of Electronic Engineering, Tohoku University,
Aoba, Aramaki,
Sendai 980, Japan

LIST OF PARTICIPANTS

573

Miley, G.H.

Fusion Studies Laboratory, University of Illinois,
103 South Goodwin Avenue,
Urbana, IL 61801, United States of America

Milora, S.L.

Oak Ridge National Laboratory,
P.O. Box Y,
Oak Ridge, TN 37831, United States of America

Mima, K,

Institute of Laser Engineering, Osaka University,
2-6 Yamada-oka, Suita,
Osaka 565, Japan

Minami, K.

Department of Electrical Engineering, Niigata University,
Ikarashi 2-no-cho,
Niigata 950-21, Japan

Minardi, E.

Associazione Euratom-ENEA, Istituto di Fisica del Plasma,
Consiglio Nazionale delle Ricerche,
Via Bassini 15,
1-20133 Milan, Italy

Mitarai, O.

Department of Electrical Engineering,
Kumamoto Institute of Technology,
4-22-1 Ikeda-cho,
Kumamoto 860, Japan

Miyahara, A.

Institute of Plasma Physics, Nagoya University,
Furo-cho, Chikusa-ku,
Nagoya 464, Japan

Miyamoto, K.

Department of Physics, University of Tokyo,
7-3-1 Hongo, Bunkyo-ku,
Tokyo 113, Japan

Miyamoto, S.

Institute of Laser Engineering, Osaka University,
2-6 Yamada-oka, Suita,
Osaka 565, Japan

Miyamoto, T.

Atomic Energy Research Institute, Nihon University,
1-8 Kanda-Surugadai, Chiyoda-ku,
Tokyo 101, Japan

Miyanaga, N.

Institute of Laser Engineering, Osaka University,
2-6 Yamada-oka, Suita,
Osaka 565, Japan

Miyoshi, S.

Plasma Research Centre, University of Tsukuba,
Sakura-mura, Niihari-gun,Ibaraki-ken 305, Japan

574

LIST OF PARTICIPANTS

Mizuuchi, T.

Plasma Physics Laboratory, Kyoto University,
Gokasho, Uji,
Kyoto 611, Japan

Mochizuki, T.

Institute of Laser Engineering, Osaka University,
2-6 Yamada-oka, Suita,
Osaka 565, Japan

Mohamed, Ahmad Mohsin

Scientific Research Council,
Electronic and Computers Research Centre,
Jadriya, P.O. Box 2131,
Baghdad, Iraq

Mohri, A.

Institute of Plasma Physics, Nagoya University,
Furo-cho, Chikusa-ku,
Nagoya 464, Japan

Momota, H.

Institute of Plasma Physics, Nagoya University,
Furo-cho, Chikusa-ku,
Nagoya 464, Japan

Morgan, O.B.

Oak Ridge National Laboratory,
P.O. Box Y,
Oak Ridge, TN 37831, United States of America

Mori, I.

Technical College, University of Tokushima,
2-1 Minami-josanjima-cho,
Tokushima 770, JapanMori, M.

Experimental Plasma Physics Laboratory,
Department of Thermonuclear Fusion Research,
Japan Atomic Energy Research Institute,
Tokai-mura, Naka-gun,Ibaraki-ken 319-11, JapanMori, S.

Japan Atomic Energy Research Institute,
2-2-2 Uchisaiwai-cho, Chiyoda-ku,
Tokyo 100, Japan

Morikawa, T.

Mitsubishi Electric Corporation,
2-3-2 Marunouchi-cho, Chiyoda-ku,
Tokyo 100, Japan

Morimoto, S.

Plasma Physics Laboratory, Kyoto University,
Gokasho, Uji,
Kyoto 611, Japan

Morino, N.

Nuclear Fusion Project Division, Hitachi Ltd,
6-4 Kanda-Surugadai-cho, Chiyoda-ku,
Tokyo 101, Japan

LIST OF PARTICIPANTS
Morita, S.

Institute of Plasma Physics, Nagoya University,
Furo-cho, Chikusa-ku,
Nagoya 464, Japan

Motojima, O.

Plasma Physics Laboratory, Kyoto University,
Gokasho, Uji,
Kyoto 611, Japan

Murakami, Y.

Department of Nuclear Engineering, University of Tokyo,
7-3-1 Hongo, Bunkyo-ku,
Tokyo 113, Japan

Muraoka, K.

Kyushu University,
6-1 Kasuga Koen, Kasuga,
Fukuoka 816, Japan

Mutoh, T.

Plasma Physics Laboratory, Kyoto University,
Gokasho, Uji,
Kyoto 611, Japan

Nagami, M.

Naka Fusion Energy Research Establishment,
Japan Atomic Energy Research Institute,801-1 Mukoyama, Naka-machi, Naka-gun,Ibaraki-ken 311-02, Japan

Nagao, A.

Research Institute for Applied Mechanics,
Kyushu University,
87 Kasuga,
Fukuoka 816, Japan

Nagao, S.

Engineering College, Chubu University,
1200 Matsumoto-cho, Kasugai-shi,
Aichi-ken 487, Japan

Nagata, M.

Department of Nuclear Engineering, Osaka University,
2-1 Yamada-oka, Suita,
Osaka 565, Japan

Nagayama, Y.

Department of Physics, University of Tokyo,
7-3-1 Hongo, Bunkyo-ku,
Tokyo 113, Japan

Naitou, H.

Institute of Plasma Physics, Nagoya University,
Furo-cho, Chikusa-ku,
Nagoya 464, Japan

Nakai, M.

Institute of Laser Engineering, Osaka University,
2-6 Yamada-oka, Suita,
Osaka 565, Japan

575

576

LIST OF PARTICIPANTS

Nakai, S.

Institute of Laser Engineering, Osaka University,
2-6 Yamada-oka, Suita,
Osaka 565, Japan

Nakajima, S.

Department of Electronics, Kyoto University,
Kitashirakawa, Sakyo-ku,
Kyoto 606, Japan

Nakamura, H.

Naka Fusion Energy Research Establishment,
Japan Atomic Energy Research Institute,801-1 Mukoyama, Naka-machi, Naka-gun,Ibaraki-ken 311-02, Japan

Nakamura, K.

Research Institute for Applied Mechanics,
Kyushu University,
87 Kasuga,
Fukuoka 816, Japan

Nakamura, M.

Osaka Institute of Technology,
Ohmiya, Asahi-ku,
Osaka 535, Japan

Nakamura, Y.

Department of Nuclear Engineering, Kyoto University,
Yoshida-Hin-machi, Sakyo-ku,
Kyoto 606, Japan

Nakamura, Y.

Research Institute for Applied Mechanics,
Kyushu University,
87 Kasuga,
Fukuoka 816, Japan

Nakano, T.

Department of Physics, Kyoto University,
Kitashirakawa, Sakyo-ku,
Kyoto 606, Japan

Nakao, Y.

Department of Nuclear Engineering,
Kyushu University,
6-10-1 Hakozaki, Higashi-ku,
Fukuoka 812, Japan

Nakashima, H.

Department of Energy Conversion Engineering,
Kyushu University,
87 Kasuga,
Fukuoka 816, Japan

Nakashima, Y.

Plasma Research Centre, University of Tsukuba,
Sakura-mura, Niihari-gun,Ibaraki-ken 305, Japan

LIST OF PARTICIPANTS
Nakasuga, M.

Plasma Physics Laboratory, Kyoto University,
Gokasho, Uji,
Kyoto 611, Japan

Nakatsuka, M.

Institute of Laser Engineering, Osaka University,
2-6 Yamada-oka, Suita,
Osaka 565, Japan

Nakayama, Y.

Department of Physics, Hiroshima University,
1-1-89 Higashisenda-machi, Naka-ku,
Hiroshima 730, Japan

Namba, C.

Institute of Plasma Physics, Nagoya University,
Furo-cho, Chikusa-ku,
Nagoya 464, Japan

Narihara, K.

Institute of Plasma Physics, Nagoya University,
Furo-cho, Chikusa-ku,
Nagoya 464, Japan

Nascimento, I.C.

Institute de Fi'sica, Universidade de Sao Paulo,
Caixa Postal 20516,
01498 Sao Paulo, Brazil

Navratil, G.A.

Department of Applied Physics andNuclear Engineering,
500 West 120th Street,
New York, NY 10027, United States of America

Negrini, F.

Faculta di Ingegneria, Universita di Bologna,
Viale Risorgimento 2,
1-40136 Bologna, Italy

Nelson, D.B.

Office of Fusion Energy, Office of Energy Research,
United States Department of Energy,
Washington, DC 20545, United States of America

Nevins, W.M.

Lawrence Livermore National Laboratory,
University of California,
P.O. Box 5511,
Livermore, CA 94550, United States of America

Newstead, C.

United States Department of State,
OES, Room 7820,
Washington, DC 20520, United States of America

Newton, A.A.

United Kingdom Atomic Energy Authority,
Culham Laboratory,Abingdon, Oxfordshire OX14 3DB, United Kingdom

577

578

LIST OF PARTICIPANTS

Ng, A.

Physics Department, University of British Columbia,
6224 Agriculture Road,
Vancouver, BC, Canada

Niedermeyer, H.

Max-Planck-Institut fur Plasmaphysik,
Boltzmannstrasse 2,
D-8046 Garching, Federal Republic of Germany

Ninomiya, H.

Naka Fusion Energy Research Establishment,
Japan Atomic Energy Research Institute,801-1 Mukoyama, Naka-machi, Naka-gun,Ibaraki-ken 311-02, Japan

Nishida, Y.

Department of Electrical Engineering,
Utsunomiya University,
2753 Ishii-machi, Utsunomiya,Tochigi 321, Japan

Nishihara, K.

Institute of Laser Engineering, Osaka University,
2-6 Yamada-oka, Suita,
Osaka 565, Japan

Nishikawa, K.

Institute for Fusion Theory, Hiroshima University,
1-1-89 Higashisenda-machi, Naka-ku,
Hiroshima 730, Japan

Nishikawa, M.

Department of Nuclear Engineering, Osaka University,
2-6 Yamada-oka, Suita,
Osaka 565, Japan

Nishikawa, M.

Mitsubishi Atomic Power Industries,
4-1-2 Shibakouen-cho, Minato-ku,
Tokyo 105, Japan

Nishimura, H.

Institute of Laser Engineering, Osaka University,
2-6 Yamada-oka, Suita,
Osaka 565, Japan

Nishimura, K.

Institute of Plasma Physics, Nagoya University,
Furo-cho, Chikusa-ku,
Nagoya 464, Japan

Niu, K.

Tokyo Institute of Technology,
4259 Nagatsuta-cho, Midori-ku,
Yokohama 227, Japan

Nogi, Y.

College of Science and Technology, Nihon University,
1-8 Kanda Surugadai, Chiyoda-ku,
Tokyo 101, Japan

LIST OF PARTICIPANTS
Noll, P.

JET Joint Undertaking,
Abingdon, Oxfordshire 0X14 3EA, United Kingdom

Norimatsu, T.

Institute of Laser Engineering, Osaka University,
2-6 Yamada-oka, Suita,
Osaka 565, Japan

Noterdaeme, J.-M.

Max-Planck-Institut fur Plasmaphysik,
Boltzmannstrasse 2,
D-8046 Garching, Federal Republic of Germany

Nuhrenberg, J.

Max-Planck-Institut ffir Plasmaphysik,
Boltzmannstrasse 2,
D-8046 Garching, Federal Republic of Germany

Obayashi, H.

Institute of Plasma Physics, Nagoya University,
Furo-cho, Chikusa-ku,
Nagoya 464, Japan

Obiki, T.

Plasma Physics Laboratory, Kyoto University,
Gokasho, Uji,
Kyoto 611, Japan

O'Connor, G.T.

Interscience,
12850 Middlebrook Road,
Germantown, MD 20874, United States of America

Odajima, K.

Experimental Plasma Physics Laboratory,
Japan Atomic Energy Research Institute,
Tokai-mura, Naka-gun,Ibaraki-ken 319-11, Japan

Ogawa, K.

Electrotechnical Laboratory,
1-1-4 Umezono, Sakura-mura, Niihari-gun,Ibaraki-ken 305, Japan

Ogawa, Y.

Institute of Plasma Physics, Nagoya University,
Furo-cho, Chikusa-ku,
Nagoya 464, Japan

Ogura, K.

Department of Physics, Kyoto University,
Kitashirakawa, Sakyo-ku,
Kyoto 606, Japan

Ohara, Y.

Naka Fusion Energy Research Establishment,
Japan Atomic Energy Research Institute,801-1 Mukoyama, Naka-machi, Naka-gun,Ibaraki-ken 311-02, Japan

579

580

LIST OF PARTICIPANTS

Ohi, S.

Plasma Physics Laboratory, Osaka University,
2-6 Yamada-oka, Suita,
Osaka 565, Japan

Ohkawa, T.

GA Technologies Inc.,
P.O. Box 85608,
San Diego, CA 92138, United States of America

Ohkubo, K.

Institute of Plasma Physics, Nagoya University,
Furo-cho, Chikusa-ku,
Nagoya 464, Japan

Ohnishi, M.

Institute of Atomic Energy, Kyoto University,
Gokasho, Uji,
Kyoto 611, Japan

Oho, K.

Department of Physics, Kyoto University,
Kitashirakawa, Sakyo-ku,
Kyoto 606, Japan

Ohsaki, H.

Sekiguchi Laboratory,
Department of Electrical Engineering, University of Tokyo,
7-3-1 Hongo, Bunkyo-ku,
Tokyo 113, Japan

Ohta, M.

Naka Fusion Energy Research Establishment,
Japan Atomic Energy Research Institute,801-1 Mukoyama, Naka-machi, Naka-gun,Ibaraki-ken 311-02, Japan

Ohtake, K.

Plasma Physics Laboratory, Kyoto University,
Gokasho, Uji,
Kyoto 611, Japan

Ohtsuka, H.

Plasma Engineering Laboratory,
Department of Thermonuclear Fusion Research,
Japan Atomic Energy Research Institute,
Tokai-mura, Naka-gun,Ibaraki-ken 319-11, Japan

Okabayashi, M.

Princeton Plasma Physics Laboratory, Princeton University,
P.O. Box 451,
Princeton, NJ 08544, United States of America

Okada, H.

Plasma Physics Laboratory, Kyoto University,
Gokasho, Uji,
Kyoto 611, Japan

Okada, M.

National Research Institute for Metals, Tsukuba Laboratory,
1-2-1 Sengen, Sakura-mura, Niihari-gun,Ibaraki-ken 305, Japan

LIST OF PARTICIPANTS

581

Okada, O.

Central Research Laboratory, Hitachi Ltd,
1-280 Higashikoigakubo, Kokubunji,
Tokyo 185, Japan

Okada, S.

Plasma Physics Laboratory, Osaka University,
2-1 Yamada-oka, Suita,
Osaka 565, Japan

Okamoto, M.

Institute of Plasma Physics, Nagoya University,
Furo-cho, Chikusa-ku,
Nagoya 464, Japan

Okamura, S.

Institute of Plasma Physics, Nagoya University,
Furo-cho, Chikusa-ku,
Nagoya 464, Japan

Okano, K.

Toshiba Research and Development Centre,
Toshiba Corporation,
4-1 Ukishima-cho, Kawasaki-ku,
Kawasaki, Kanagawa 210, Japan

Okazaki, T.

Energy Research Laboratory, Hitachi Ltd,1168 Moriyama-cho, Hitachi-shi,
Ibaraki-ken 316, Japan

Ono, M.

Princeton Plasma Physics Laboratory, Princeton University,
P.O. Box 451,
Princeton, NJ 08544, United States of America

Oomens, A.A.M.

FOM Instituut voor Plasmafysica 'Rynhuizen',
Edisonbaan 14, P.O. Box 1207,
NL-3430 BE Nieuwegein, Netherlands

Ortolani, S.

Istituto Gas Ionizzati,
Consiglio Nazionale delle Ricerche,
Via Gradenigo 6A,
1-35131 Padua, Italy

Oshiyama, H.

Department of Electrical Engineering,
Kyoto Institute of Technology,
Matsugasaki, Sakyo-ku,
Kyoto 606, Japan

Overskei, D.

GA Technologies Inc.,
P.O. Box 85608,
San Diego, CA 92138, United States of America

Owadano, Y.

Electrotechnical Laboratory,
1-1-4 Umezono, Sakura-mura, Niihari-gun,Ibaraki-ken 305, Japan

582

LIST OF PARTICIPANTS

Ozaki, A.

Power and Fusion Technology Development Department,
Toshiba Corporation,
1-6-1 Uchisaiwai-cho, Chiyoda-ku,
Tokyo 100, Japan

Ozaki, S.

Department of Physics, Kyoto University,
Kitashirakawa, Sakyo-ku,
Kyoto 606, Japan

Pacher, H.D.

INRS Energie,
C.P. 1020,
Varennes, Quebec JOL 2P0, Canada

Parker, R.R.

Plasma Fusion Center,
Massachusetts Institute of Technology,
Cambridge, MA 02139, United States of America

Parlange, F.

Commissariat a l'energie atomique (SIG),
Centre d'etudes nucleaires de Grenoble,
F-38041 Grenoble Cedex, France

Pavlichenko, O.S.

Kharkov Physico-Technical Institute,
Ukrainian SSR Academy of Sciences,
Kharkov 310108, Union of Soviet Socialist Republics

Pease, R.S.

United Kingdom Atomic Energy Authority,
Culham Laboratory,Abingdon, Oxfordshire OX14 3DB, United Kingdom

Perlado, J.M.

Institute de Fusion Nuclear,
Universidad Politecnica de Madrid,
P° Castellana 80,
E-28046 Madrid, Spain

Pfirsch, P.

Max-Planck-Instltut fur Plasmaphysik,
Boltzmannstrasse 2,
D-8046 Garching, Federal Republic of Germany

Phillips, G J .

Atomic Energy of Canada Ltd,
Chalk River Nuclear Laboratories,
Chalk River, Ontario KOJ U0, Canada

Pinkau, K.

Max-Planck-Institut fur Plasmaphysik,
Boltzmannstrasse 2,
D-8046 Garching, Federal Republic of Germany

Porkolab, M.

Plasma Fusion Center, Massachusetts Institute of Technology,
Cambridge, MA 02139, United States of America

LIST OF PARTICIPANTS

583

Post, D.E.

Princeton Plasma Physics Laboratory, Princeton University,
P.O. Box 451,
Princeton, NJ 08544, United States of America

Post, R.S.

Plasma Fusion Center, Massachusetts Institute of Technology,
Cambridge, MA 02139, United States of America

Prater, R.

GA Technologies Inc.,
P.O. Box 85608,
San Diego, CA 92138, United States of America

Prevot, F.G.

Commissariat a 1'energie atomique,Centre d'&udes nucleaires de Cadarache,
B.P. 1,
F-13108 Saint-Paul-lez-Durance, France

Qian, ShangjieSouthwestern Institute of Physics,
P.O. Box 15,
Leshan, Sichuan, China

Quinn, W.E.

Los Alamos National Laboratory,
P.O. Box 1663,
Los Alamos, NM 87545, United States of America

Raeder, J.

NET Team, Max-Planck-Institut fflr Plasmaphysik,
Boltzmannstrasse 2,
D-8046 Garching, Federal Republic of Germany

Rager, J.-P.

Commission of the European Communities (DG XII/Fusion),
200, rue de la Loi,
B-1049 Brussels, Belgium

Rau, F.

Max-PJanck-Institut fur Plasmaphysik,
Boltzmannstrasse 2,
D-8046 Garching, Federal Republic of Germany

Rebut, P.H.

JET Joint Undertaking,
Abingdon, Oxfordshire OX14 3EA, United Kingdom

Renner, H.

Max-Planck-Institut fur Plasmaphysik,
Boltzmannstrasse 2,
D-8046 Garching, Federal Republic of Germany

Rewoldt, G.

Princeton Plasma Physics Laboratory, Princeton University,
P.O. Box 451,
Princeton, NJ 08544, United States of America

Ribe, F.L.

University of Washington,
Benson Hall BF-10,
Seattle, WA 98195, United States of America

584

LIST OF PARTICIPANTS

Ringler, H.

Max-Planck-Institut fur Plasmaphysik,
Boltzmannstrasse 2,
D-8046 Garching, Federal Republic of Germany

Ripin, B.H.

Naval Research Laboratory,
4555 Overlook Avenue,
Washington, DC 20375-5000, United States of America

Riviere, A.C.

United Kingdom Atomic Energy Authority,
Culham Laboratory,Abingdon, Oxfordshire 0X14 3DB, United Kingdom

Robertson, T.

United States Department of State,
Washington, DC 20520, United States of America

Robinson, D.C.

United Kingdom Atomic Energy Authority,
Culham Laboratory,Abingdon, Oxfordshire OX14 3DB, United Kingdom

Rodrigo, A.

Comisi6n Nacional de Energfa Attfrnica (Dept. Prospectiva),
Avda. del Libertador 8250,
Buenos Aires, Argentina

Rogister, A.

Euratom-KFA Association, Institut fur Plasmaphysik,
Kernforschungsanlage Jfllich GmbH,
Postfach 1913,
D-5170 Julich, Federal Republic of Germany

Rosenbluth, M.N.

Institute for Fusion Studies,
University of Texas at Austin,
Austin, TX 7812, United States of America

Rosocha, L.A.

Los Alamos National Laboratory,
P.O. Box 1663,
Los Alamos, NM 87545, United States of America

Rostagni, G.

Associazione Euratom-ENEA, Istituto Gas Ionizzati,
Consiglio Nazionale delle Ricerche,
Via Gradenigo 6A,
1-35131 Padua, Italy

Roubin, J.-P.

Commissariat a l'e'nergie atomique,Centre d'e'tudes nucllaires de Cadarache,
B.P. 1,
F-13108 Saint-Paul-lez-Durance, France

Rutherford, P.H.

Princeton Plasma Physics Laboratory, Princeton University,
P.O. Box 451,
Princeton, NJ 08544, United States of America

LIST OF PARTICIPANTS
Ryutov, D.D.

Nuclear Physics Institute,
Siberian Branch of the USSR Academy of Sciences,
Prospekt Nauki 11,
630090 Novosibirsk, Union of Soviet Socialist Republics

Saito, R.

Mitsubishi Electronic Corporation,
2-3-2 Marunouchi-cho, Chiyoda-ku,
Tokyo 100, Japan

Saito, T.

Plasma Research Centre, University of Tsukuba,
Sakura-mura, Niihari-gun,Ibaraki-ken 305, Japan

Sakabe, S.

Institute of Laser Engineering, Osaka University,
2-6 Yamada-oka, Suita,
Osaka 565, Japan

Sakanaka, P.H.

Institutes de Ffsica "Gleb Wataghin",
Universidade Estadual de Campinas,
C.P. 1170 Unicamp,
13100 Campinas, SP, Brazil

Sano, F.

Plasma Physics Laboratory, Kyoto University,
Gokasho, Uji,
Kyoto 611, Japan

Santini, F.

Associazione Euratom-ENEA,
Centro Ricerche Energia Frascati,
C.P. 65,
1-00044 Frascati, Rome, Italy

Sanuki, H.

Institute of Plasma Physics, Nagoya University,
Furo-cho, Chikusa-ku,
Nagoya 464, Japan

Sasaki, T.

Institute of Laser Engineering, Osaka University,
2-6 Yamada-oka, Suita,
Osaka 565, Japan

Sasao, M.

Institute of Plasma Physics, Nagoya University,
Furo-cho, Chikusa-ku,
Nagoya 464, Japan

Sato, K.I.

Institute of Plasma Physics, Nagoya University,
Furo-cho, Chikusa-ku,
Nagoya 464, Japan

Sato, K.N.

Institute of Plasma Physics, Nagoya University,
Furo-cho, Chikusa-ku,
Nagoya 464, Japan

585

586

LIST OF PARTICIPANTS

Sato, M.

Department of Physics, Nihon University,
1-8 Kanda-Surugadai, Chiyoda-ku,
Tokyo 101, Japan

Sato, M.

Plasma Physics Laboratory, Kyoto University,
Gokasho, Uji,
Kyoto 611, Japan

Sato, T.

Institute for Fusion Theory, Hiroshima University,
1-1-89 Higashisenda-machi, Naka-ku,
Hiroshima 730, Japan

Sato, T.

Institute of Plasma Physics, Nagoya University,
Furo-cho, Chikusa-ku,
Nagoya 464, Japan

Sato, Y.

Electrotechnical Laboratory,
1-1-4 Umezono, Sakura-mura, Niihari-gun,Ibaraki-ken 305, Japan

Satomi, N.

Department of Nuclear Engineering, Osaka University,
2-6 Yamada-oka, Suita,
Osaka 565, Japan

Sawada, Y.

Toshiba Corporation,
2-4 Suehiro-cho, Tsurumi-ku,
Yokohama 230, Japan

Scarin, P.

Associazione Euratom-ENEA, Istituto Gas Ionizzati,
Consiglio Nazionale delle Ricerche,
Via Gradenigo 6A,
1-35131 Padua, Italy

Schep, T.J.

FOM Instituut voor Plasmafysica 'Rynhuizen',
Edisonbaan 14, P.O. Box 1207,
NL-3430 BE Nieuwegein, Netherlands

Schliiter, J.

Euratom-KFA Association, Institut fur Plasmaphysik,
Kernforschungsanlage Julich GmbH,
Postfach 1913,
D-5170 Julich, Federal Republic of Germany

Schmidt, J.

Princeton Plasma Physics Laboratory, Princeton University,
P.O. Box 451,
Princeton, NJ 08544, United States of America

Schoenberg, K.F.

Los Alamos National Laboratory,
P.O. Box 1663,
Los Alamos, NM 87545, United States of America

LIST OF PARTICIPANTS

587

Schopf, K.

Institut fiir Theoretische Physik, Universitat Innsbruck,
Technikerstrasse 15,
A-6020 Innsbruck, Austria

Schriever, R.L.

Office of Inertial Fusion,
United States Department of Energy,
Washington, DC 20545, United States of America

Schuller, F.C.

JET Joint Undertaking,
Abingdon, Oxfordshire 0X14 3EA, United Kingdom

Seki, H.

Department of Electrical Engineering, University of Tokyo,
7-3-1 Hongo, Bunkyo-ku,
Tokyo 113, Japan

Seki, S.

Seki, Y.

Naka Fusion Energy Research Establishment,
Japan Atomic Energy Research Institute,801-1 Mukoyama, Naka-machi, Naka-gun,Ibaraki-ken 311-02, Japan
Naka Fusion Energy Research Establishment,
Japan Atomic Energy Research Institute,801-1 Mukoyama, Naka-machi, Naka-gun,Ibaraki-ken 311-02, Japan

Sekiguchi, T.

Department of Electrical Engineering, University of Tokyo,
7-3-1 Hongo, Bunkyo-ku,
Tokyo 113, Japan

Seligman, H.

Internationa] Atomic Energy Agency,
P.O. Box 100, A-1400 Vienna, Austria

Sen, A.

Physical Research Laboratory,
Navrangpura, Ahmedabad 380009, India

Sengoku, S.

Department of Thermonuclear Fusion Research,
Japan Atomic Energy Research Institute,
Tokai-mura, Naka-gun,Ibaraki-ken 319-11, Japan

Sgro, A.G.

Los Alamos National Laboratory,
P.O. Box 1663,
Los Alamos, NM 87545, United States of America

Shaing, K.C.

Oak Ridge National Laboratory,
P.O. Box Y,
Oak Ridge, TN 37831, United States of America

Shan, Zhenkui

Southwestern Institute of Physics,
P.O. Box 15,
Leshan, Sichuan, China

588

LIST OF PARTICIPANTS

Shigeta, K.

Department of Electronics, Kyoto University,
Kitashirakawa, Sakyo-ku,
Kyoto 606, Japan

Shiina, S.

Atomic Energy Research Institute, Nihon University,
1-8 Kanda-Surugadai, Chiyoda-ku,
Tokyo 101, Japan

Shikama, T.

National Research Institute for Metals, Tsukuba Laboratory,
1-2-1 Sengen, Sakura-mura, Niihari-gun,Ibaraki-ken 305, Japan

Shimada, R.

Naka Fusion Energy Research Establishment,
Japan Atomic Energy Research Institute,801-1 Mukoyama, Naka-machi, Naka-gun,Ibaraki-ken 311-02, Japan

Shimada, T.

Electrotechnical Laboratory,
1-1-4 Umezono, Sakura-mura, Niihari-gun,Ibaraki-ken 305, Japan

Shimamoto, S.

Naka Fusion Energy Research Establishment,
Japan Atomic Energy Research Institute,801-1 Mukoyama, Naka-machi, Naka-gun,Ibaraki-ken 311-02, Japan

Shimamura, S.

College of Science and Technology, Nihon University,
1-8 Kanda-Surugadai, Chiyoda-ku,
Tokyo 101, Japan

Shimizu, K.

Naka Fusion Energy Research Establishment,
Japan Atomic Energy Research Institute,801-1 Mukoyama, Naka-machi, Naka-gun,Ibaraki-ken 311-02, Japan

Shimomura, Y.

Naka Fusion Energy Research Establishment,
Japan Atomic Energy Research Institute,801-1 Mukoyama, Naka-machi, Naka-gun,Ibaraki-ken 311-02, Japan

Shindo, H.

Faculty of Engineering, Hiroshima University,
Shitami Saijo-cho, Higashi,
Hiroshima 724, Japan

Shinohara, H.

Power and Fusion Technology Development Department,
Toshiba Corporation,
1-6-1 Uchisaiwai-cho, Chiyoda-ku,
Tokyo 100, Japan

LIST OF PARTICIPANTS
Shinya, K.

Toshiba Research and Development Centre,
Toshiba Corporation,
4-1 Ukishima-cho, Kawasaki-ku,
Kawasaki 210, Japan

Shiraga, H.

Institute of Laser Engineering, Osaka University,
2-6 Yamada-oka, Suita,
Osaka 565, Japan

Shirakata, H.

Naka Fusion Energy Research Establishment,
Japan Atomic Energy Research Institute,801-1 Mukoyama, Naka-machi, Naka-gun,Ibaraki-ken 311-02, Japan

Simonen, T.C.

Lawrence Livermore National Laboratory,
University of California,
P.O. Box* 5511,
Livermore, CA 94550, United States of America

Sinman, A.

Physics Department, Nuclear Fusion Laboratory,
Ankara Nuclear Research and Training Centre,
Besevler, Ankara, Turkey

Sinman, S.

Electrical and Electronics Engineering Department,
Plasma Engineering Laboratory,
Middle East Technical University,
Ankara, Turkey

Skarsgard, H.M.

Physics Department, University of Saskatchewan,
Saskatoon, Saskatchewan S7N 0W0, Canada

Smirnov, V.P.

I.V. Kurchatov Institute of Atomic Energy.
42 Ulitsa Kurchatova, P.O. Box 3402,
123182 Moscow, Union of Soviet Socialist Republics

Soldner, F.X.

Max-Planck-Institut fur Plasmaphysik,
Boltzmannstrasse 2,
D-8046 Garching, Federal Republic of Germany

Soltwisch, H.

Euratom-KFA Association, Institut fur Plasmaphysik,
Kernforschungsanlage Julich GmbH,
Postfach 1913,
D-5170 Julich, Federal Republic of Germany

Soures, J.M.

Laboratory for Laser Energetics, University of Rochester,
250 East River Road,
Rochester, NY 14623-1299, United States of America

Speth, E.

Max-Planck-Institut fiir Plasmaphysik,
Boltzmannstrasse 2,
D-8046 Garching, Federal Republic of Germany

589

590

LIST OF PARTICIPANTS

Steinmetz, K.

Max-Planck-Institut fflr Plasmaphysik,
Boltzmannstrasse 2,
D-8046 Garching, Federal Republic of Germany

Stork, D.

JET Joint Undertaking,
Abingdon, Oxfordshire 0X14 3EA, United Kingdom

Storm, E.

Lawrence Livermore National Laboratory,
University of California,
P.O. Box 5508,
Livermore, CA 94550, United States of America

Sudo, S.

Plasma Physics Laboratory, Kyoto University,
Gokasho, Uji,
Kyoto 611, Japan

Sugama, H.

Plasma Physics Laboratory, Kyoto University,
Gokasho, Uji,
Kyoto 611, Japan

Sugawa, M.

Department of Physics, Faculty of Science,
Ehime University,
Bunkyo-cho, Matsuyama,
Ehime 790, Japan

Sugawara, T.

Toshiba Research and Development Centre,
Toshiba Corporation,
4-1 Ukishima-cho, Kawasaki-lcu,
Kawasaki 210, Japan

Sugihara, R.

Institute of Plasma Physics, Nagoya University,
Furo-cho, Chikusa-ku,
Nagoya 464, Japan

Sugimoto, S.

Plasma Physics Laboratory, Osaka University,
2-6 Yamada-oka, Suita,
Osaka 565, Japan

Sugiyama, L.E.

Plasma Fusion Center, Massachusetts Institute of Technology,
Cambridge, MA 02139, United States of America

Sunako, K.

Department of Physics, Tokai University,
1117 Kitakaname, Hiratsuka,
Kanagawa 259-12, Japan

Suzuki, N.

Ishikawajima-Harima Heavy Industries,
Tokyo-Chuo Bldg.,
6-2-1 Marunouchi-cho, Chiyoda-ku,
Tokyo 100, Japan

LIST OF PARTICIPANTS

591

Suzuki, N.

Department of Thermonuclear Fusion Research,
Japan Atomic Energy Research Institute,
Tokai-mura, Naka-gun,Ibaraki-ken 319-11, Japan

Suzuki, Y.

Naka Fusion Energy Research Establishment,
Japan Atomic Energy Research Institute,801-1 Mukoyama, Naka-machi, Naka-gun,Ibaraki-ken 311-02, Japan

Swanson, D.G.

Physics Department, Auburn University,
Auburn, AL 36849, United States of America

Sweetman, D.R.

United Kingdom Atomic Energy Authority,
Culham Laboratory,Abingdon, Oxfordshire OX 14 3DB, United Kingdom

Taguchi, M.

Department of Applied Physics, University of Tokyo,
7-3-1 Hongo, Bunkyo-ku,
Tokyo 113, Japan

Takabe, H.

Institute of Laser Engineering, Osaka University,
2-6 Yamada-oka, Suita,
Osaka 565, Japan

Takahashi, H.

Princeton Plasma Physics Laboratory, Princeton University,
P.O. Box 451,
Princeton, NJ 08544, United States of America

Takahashi, R.

Plasma Physics Laboratory, Kyoto University,
Gokasho, Uji,
Kyoto 611, Japan

Takamura, S.

Department of Electrical Engineering, Nagoya University,
Furo-cho, Chikusa-ku,
Nagoya 464, Japan

Takasugi, K.

Atomic Energy Research Institute, Nihon University,
1-8 Kanda-Surugadai, Chiyoda-ku,
Tokyo 101, Japan

Takeda, S.

Electrotechnical Laboratory,
1-1-4 Umezono, Sakura-mura, Niihari-gun,Ibaraki-ken 305, Japan

Takeda, T.

Department of Thermonuclear Fusion Research,
Tokai Research Establishment,
Japan Atomic Energy Research Institute,
Tokai-mura, Naka-gun,Ibaraki-ken 319-11, Japan

592

LIST OF PARTICIPANTS

Takeiri, Y.

Plasma Physics Laboratory, Kyoto University,
Gokasho, Uji,
Kyoto 611, Japan

Takeuchi, H.

Naka Fusion Energy Research Establishment,
Japan Atomic Energy Research Institute,801-1 Mukoyama, Naka-machi, Naka-gun,Ibaraki-ken 311-02, Japan

Takeuchi, S.

Faculty of Engineering, Yamanashi University,
4-3-11 Takeda, Kofii,
Yamanashi 400, Japan

Takizuka, T.

Japan Atomic Energy Research Institute,
Tokai-mura, Naka-gun,Ibaraki-ken 319-11, Japan

Tamano, T.

GA Technologies Inc.,
P.O. Box 85608,
San Diego, CA 92138, United States of America

Tamaru, T.

Faculty of Engineering, Hiroshima University,
Shitami, Saijo-cho, Higashi,
Hiroshima 724, Japan

Tanaka, H.

Department of Physics, Kyoto University,
Kitashirakawa, Sakyo-ku,
Kyoto 606, Japan

Tanaka, K.

Institute of Laser Engineering, Osaka University,
2-6 Yamada-oka, Suita,
Osaka 565, Japan

Tanaka, M.

Institute for Fusion Theory, Hiroshima University,
1-1-89 Higashisenda-machi, Naka-ku,
Hiroshima 730, Japan

Tanaka, M.

Interdisciplinary Graduate School of Engineering Sciences,
Kyushu University,
6-1 Kasuga Koen, Kasuga,
Fukuoka 816, Japan

Tanaka, M.

Naka Fusion Energy Research Establishment,
Japan Atomic Energy Research Institute,801-1 Mukoyama, Naka-machi, Naka-gun,Ibaraki-ken 311-02, Japan

Tanaka, S.

Department of Physics, Kyoto University,
Kitashirakawa, Sakyo-ku,
Kyoto 606, Japan

LIST OF PARTICIPANTS

593

Tanaka, S.

Department of Electrical and Electronics Engineering,
Sophia University,
7-1 Kioi-cho, Chiyoda-ku,
Tokyo 102, Japan

Tang, W.M.

Princeton Plasma Physics Laboratory, Princeton University,
P.O. Box 451,
Princeton, NJ 08544, United States of America

Tanga, A.

JET Joint Undertaking,
Abingdon, Oxfordshire OX14 3EA, United Kingdom

Tani, K.

Naka Fusion Energy Research Establishment,
Japan Atomic Energy Research Institute,801-1 Mukoyama, Naka-machi, Naka-gun,Ibaraki-ken 311-02, Japan

Tani, T.

Facility Operation and Engineering Division,
Department of Thermonuclear Fusion Research,
Japan Atomic Energy Research Institute,
Tokai-mura, Naka-gun,Ibaraki-ken 319-11, Japan

Taroni, A.

JET Joint Undertaking,
Abingdon, Oxfordshire 0X14 3EA, United Kingdom

Tartari, U.

Associazione Euratom-ENEA, Istituto di Fisica del Plasma,
Consiglio Nazionale delle Ricerche,
Via Bassini 15,
1-20133 Milan, Italy

Taylor, R J .

Tokamak Fusion Laboratory, University of California,
1540 Boelter Hall,
Los Angeles, CA 90024, United States of America

Tazima, T.

Institute of Plasma Physics, Nagoya University,
Furo-cho, Chikusa-ku,
Nagoya 464, Japan

Terasawa, S.

Nuclear Power Generation Division, Hitachi, Ltd,
6-4 Kanda-Surugadai-cho, Chiyoda-ku,
Tokyo 101, Japan

Terashima, Y.

Institute of Plasma Physics, Nagoya University,
Furo-cho, Chikusa-ku,
Nagoya 464, Japan

Tereshin, V.

Division of Research and Laboratories,
International Atomic Energy Agency,
P.O. Box 100,
A-1400 Vienna, Austria

594

LIST OF PARTICIPANTS

Terumichi, Y.

Department of Physics, Kyoto University,
Kitashirakawa, Sakyo-ku,
Kyoto 606, Japan

Thomassen, K.

Lawrence Livermore National Laboratory,
University of California,
P.O. Box 5508
Livermore, CA 94550, United States of America

Todd, A.M.M.

Grumman Corporation,
105 College Road East,
Princeton, NJ 08540, United States of America

Todd, T.N.

United Kingdom Atomic Energy Authority,
Culham Laboratory,Abingdon, Oxfordshire OX14 3DB, United Kingdom

Todoroki, J.

Institute of Plasma Physics, Nagoya University,
Furo-cho, Chikusa-ku,
Nagoya 464, Japan

Toi, K.

Institute of Plasma Physics, Nagoya University,
Furo-cho, Chikusa-ku,
Nagoya 464, Japan

Tokuda, S.

Department of Thermonuclear Fusion Research,
Tokai Research Establishment,
Japan Atomic Energy Research Institute,
Tokai-mura, Naka-gun,Ibaraki-ken 319-11, Japan

Tomabechi, K.

Naka Fusion Energy Research Establishment,
Japan Atomic Energy Research Institute,801-1 Mukoyama, Naka-machi, Naka-gun,Ibaraki-ken 311-02, Japan

Tominaga, S.

Mitsubishi Electric Corporation,
2-3-2 Marunouchi-cho, Chiyoda-ku,
Tokyo 100, Japan

Tomita, Y.

Institute of Plasma Physics, Nagoya University,
Furo-cho, Chikusa-ku,
Nagoya 464, Japan

Tone, T.

Naka Fusion Energy Research Establishment,
Japan Atomic Energy Research Institute,801-1 Mukoyama, Naka-machi, Naka-gun,Ibaraki-ken 311-02, Japan

LIST OF PARTICIPANTS

595

Toyama, H.

Department of Physics, University of Tokyo,
7-3-1 Hongo, Bunkyo-ku,
Tokyo 113, Japan

Trivelpiece, A.W.

Office of Energy Research,
United States Department of Energy,
Washington, DC 20585, United States of America

Tsai, Shihtung

Institute of Physics, Chinese Academy of Sciences,
P.O. Box 603,
Beijing, China

Tsukishima, T.

Faculty of Engineering, Nagoya University,
Furo-cho, Chikusa-ku,
Nagoya 464, Japan

Tsunematsu, T.

Department.of Thermonuclear Fusion Research,
Tokai Research Establishment,
Japan Atomic Energy Research Institute,
Tokai-mura, Naka-gun,Ibaraki-ken 319-1.1, Japan

Tsushima, A.

Department of Electronic Engineering, Tohoku University,
Aoba, Aramaki,
Sendai 980, Japan

Tsutsui, H.

Department of Applied Physics, University of Tokyo,
7-3-1 Hongo, Bunkyo-ku,
Tokyo 113, Japan

Tuczek, H.

Fachbereich Physik, Universitat GH Essen,
Universitatsstrasse 5,
D-4300 Essen, Federal Republic of Germany

Tuda, T.

Department of Thermonuclear Fusion Research,
Tokai Research Establishment,
Japan Atomic Energy Research Institute,
Tokai-mura, Naka-gun,Ibaraki-ken 319-11, Japan

Uchida, T.

Institute of Plasma Physics, Nagoya University,
Furo-cho, Chikusa-ku,
Nagoya 464, Japan

Uehara, K.

Naka Fusion Energy Research Establishment,
Japan Atomic Energy Research Institute,801-1 Mukoyama, Naka-machi, Naka-gun,Ibaraki-ken 311-02, Japan

596

LIST OF PARTICIPANTS

Uo, K.

Plasma Physics Laboratory, Kyoto University,
Gokasho, Uji,
Kyoto 611, Japan

Uyama, T.

Course of Electromagnetic Energy Engineering,Osaka University,
2-1 Yamada-oka, Suita,
Osaka 565, Japan

Van Dam, J.W.

Insitute for Fusion Studies,
University of Texas at Austin,
Austin, TX 78712, United States of America

Vandenplas, P.E.

Ecole Royale Militaire,
30, avenue de la Renaissance,
B-1040 Brussels, Belgium

VanDevender, J.P.

Pulsed Power Sciences, Sandia National Laboratories,
P.O. Box 5800,
Albuquerque, NM 87185-5800, United States of America

Velarde, G.

Institute de Fusi<5n Nuclear,
Escuela Tecnica Superior de Ingenieros Industrials,
Universidad Politecnica de Madrid,
P° Castellana 80,
E-28046 Madrid, Spain

Vlases, G.C.

Nuclear Engineering Department (FL-10),
University of Washington,
Seattle, WA 98195, United States of America

Vojtenko, D.A.

I.N. Vekua Institute of Physics and Technology,
State Committee for the Utilization of Atomic Energy,
Sukhumi, Union of Soviet Socialist Republics

Wada, H.

National Research Institute for Metals, Tsukuba Laboratory,
1-2-1 Sengen, Sakura-mura, Niihari-gun,Ibaraki-ken 305, Japan

Waelbroeck, F.

Euratom-KFA Association, Institut fur Plasmaphysik,
Kernforschungsanlage Julich GmbH,
Postfach 1913,
D-5170 Julich, Federal Republic of Germany

Wagner, F.

Max-Planck-Institut fur Plasmaphysik,
Boltzmannstrasse 2,
D-8046 Garching, Federal Republic of Germany

Wakatani, M.

Plasma Physics Laboratory, Kyoto University,
Gokasho, Uji,
Kyoto 611, Japan

LIST OF PARTICIPANTS

597

Waltz, R.E.

GA Technologies Inc.,
P.O. Box 85608,
San Diego, CA 92138, United States of America

Watanabe, H.

Faculty of Engineering, Tohoku University,
Aoba, Aramaki,
Sendai 980, Japan

Watanabe, K.

Course of Electromagnetic Energy Engineering,Osaka University,
2-1 Yamada-oka, Suita,
Osaka 565, Japan

Watanabe, Y.

Department of Electronic Engineering, Tohoku University,
Aoba, Aramaki,
Sendai 980, Japan

Watari, T.

Institute of Plasma Physics, Nagoya University,
Furo-cho, Chikusa-ku,
Nagoya 464, Japan

Watkins, M.L.

JET Joint Undertaking,
Abingdon, Oxfordshire OX14 3EA, United Kingdom

Watteau, J.-P.

Commissariat a l'energie atomique,Centre d'6tudes de Limeil-Valenton,
B.P. 27,
F-94190 Villeneuve-Saint-Georges, France

Wegrowe, J.G.

NET Team, Max-Planck-Institut fiir Plasmaphysik,
Boltzmannstrasse 2,
D-8046 Garching, Federal Republic of Germany

Weitzner, H.

Magneto-Fluid Dynamics Division,
Courant Institute of Mathematical Sciences,
New York University,
251 Mercer Street,
New York, NY 10012, United States of America

Wesson, J.A.

JET Joint Undertaking,
Abingdon, Oxfordshire OX14 3EA, United Kingdom

Weynants, R.R.

Ecole Royale Militaire,
30, avenue de la Renaissance,
B-1040 Brussels, Belgium

Willis, J.

United States Department of Energy (ER-55),
Washington, DC 20545, United States of America

598

LIST OF PARTICIPANTS

Winter, H.

Institut fur Allgemeine Physik, Technische Universita't,
Karlsplatz 13,
A-1040 Vienna, Austria

Witkovski, S.

Max-Planck-Institut fur Quantenoptik,
Postfach 1513,
D-8046 Garching, Federal Republic of Germany

Wobig, H.

Max-Planck-Institut fur Plasmaphysik,
Boltzmannstrasse 2,
D-8046 Garching, Federal Republic of Germany

Wolf, G.H.

Euratom-KFA Association, Institut fur Plasmaphysik,
Kernforschungsanlage Julich GmbH,
Postfach 1913,
D-5170 Julich, Federal Republic of Germany

Wootton, A.J.

Fusion Research Center,
The University of Texas at Austin,
Austin, TX 78712, United States of America

Wright, B.L.

Los Alamos National Laboratory,
P.O. Box 1663,
Los Alamos, NM 87545, United States of America

Xie, Jikang

Institute of Plasma Physics, Academia Sinica,
P.O. Box 26,
Hefei, Anhui, China

Xiong, Zirong

Southwestern Institute of Physics,
P.O. Box 15,
Leshan, Sichuan, China

Yabe, T.

Institute of Laser Engineering, Osaka University,
2-6 Yamada-oka, Suita,
Osaka 565, Japan

Yagi, Y.

Electrotechnical Laboratory,
1-1-4 Umezono, Sakura-mura, Niihari-gun,Ibaraki-ken 305, Japan

Yahagi, E.

Electrotechnical Laboratory,
1-1-4 Umezono, Sakura-mura, Niihari-gun,Ibaraki-ken 305, Japan

Yamada, H.

Department of Nuclear Engineering, University of Tokyo,
7-3-1 Hongo, Bunkyo-ku,
Tokyo 113, Japan

LIST OF PARTICIPANTS

599

Yamada, M.

Princeton Plasma Physics Laboratory, Princeton University,
P.O. Box 451,
Princeton, NJ 08544, United States of America

Yamagiwa, K.

Department of Physics, Shizuoka University,
836 Ohta, Shizuoka 422, Japan

Yamaguchi, N.

Plasma Research Centre, University of Tsukuba,
Sakura-mura, Niihari-gun,Ibaraki-ken 305, Japan

Yamamoto, K.

Japan Atomic Industrial Forum, Inc.,
1-1-13 Shinbashi, Minato-ku,
Tokyo 105, Japan

Yamamoto, S.

Naka Fusion Energy Research Establishment,
Japan Atomic Energy Research Institute,801-1 Mukoyama, Naka-machi, Naka-gun,Ibaraki-ken 311-02, Japan

Yamamoto, T.

Naka Fusion Energy Research Establishment,
Japan Atomic Energy Research Institute,801-1 Mukoyama, Naka-machi, Naka-gun,Ibaraki-ken 311-02, Japan

Yamamoto, T.

Department of Electronic Engineering, Gunma University,
1-5-1 Tenjin-cho, Kiryu,
Gunma 376, Japan

Yamamoto, Y.

Institute of Atomic Energy, Kyoto University,
Gokasho, Uji,
Kyoto 611, Japan

Yamanaka, C.

Institute of Laser Engineering, Osaka University,
2-6 Yamada-oka, Suita,
Osaka 565, Japan

Yamanaka, M.

Institute of Laser Engineering, Osaka University,
2-6 Yamada-oka, Suita,
Osaka 565, Japan

Yamanaka, T.

Institute of Laser Engineering, Osaka University,
2-6 Yamada-oka, Suita,
Osaka 565, Japan

Yamato, H.

Toshiba Research and Development Centre,
Toshiba Corporation,
4-1 Ukishima-cho, Kawasaki-ku,
Kawasaki 210, Japan

600

LIST OF PARTICIPANTS

Yamazaki, A.

Department of Physics, Kyoto University,
Kitashirakawa, Sakyo-ku,
Kyoto 606, Japan

Yamazaki, K.

Institute of Plasma Physics, Nagoya University,
Furo-cho, Chikusa-ku,
Nagoya 464, Japan

Yano, M.

Electrotechnical Laboratory,
1-1-4 Umezono, Sakura-mura, Niihari-gun,Ibaraki-ken 305, Japan

Yasaka, Y.

Department of Electronics, Kyoto University,
Kitashirakawa, Sakyo-ku,
Kyoto 606, Japan

Yatsu, K.

Plasma Research Centre, University of Tsukuba,
Sakura-mura, Niihari-gun,Ibaraki-ken 305, Japan

Yatsu, S.

Energy Conversion Research Institute, Hokkaido University,
Nishi-8, Kita-13, Kita-ku,
Sapporo 060, Japan

Yatsui, K.

Laboratory of Beam Technology,
Technological University of Nagaoka,
Nagaoka, Niigata 940-21, Japan

Yokomizo, H.

Naka Fusion Energy Research Establishment,
Japan Atomic Energy Research Institute,801-1 Mukoyama, Naka-machi, Naka-gun,Ibaraki-ken 311-02, Japan

Yokoyama, M.

Institute of Laser Engineering, Osaka University,
2-6 Yamada-oka, Suita,
Osaka 565, Japan

Yoshida, Z.

Department of Nuclear Engineering, University of Tokyo,
7-3-1 Hongo, Bunkyo-ku,
Tokyo 113, Japan

Yoshikawa, K.

Institute of Atomic Energy, Kyoto University,
Gokasho, Uji,
Kyoto 611, Japan

Yoshikawa, M.

Naka Fusion Energy Research Establishment,
Japan Atomic Energy Research Institute,801-1 Mukoyama, Naka-machi, Naka-gun,Ibaraki-ken 311-02, Japan

LIST OF PARTICIPANTS

601

Yoshioka, K.

Energy Research Laboratory, Hitachi Ltd,1168 Moriyama-cho, Hitachi-shi,
Ibaraki-ken 316, Japan

ZScek, F.

Institute of Plasma Physics,
Czechoslovak Academy of Sciences,
Pod vodanrenskou vezf 4,
CS-182 11 Prague 8, Czechoslovakia

Zakharov, L.E.

I.V. Kurchatov Institute of Atomic Energy,
42 Ulitsa Kurchatova, P.O. Box 3402,
123182 Moscow, Union of Soviet Socialist Republics

Zarnstorff, M.C.

Princeton Plasma Physics Laboratory, Princeton University,
P.O. Box 451,
Princeton, NJ 08544, United States of America

Zushi, H.

Plasma Physics Laboratory, Kyoto University,
Gokasho, Uji,
Kyoto 611, Japan

AUTHOR INDEXIn the index entries the Roman numeral refers to the volume number;the following numerals refer to the page within that volume

Abe, H.: 1553; II 109
Abe, T.: I 11, 89, 217, 563
Abe, Y.: I 309
Adams, F.P.: III 187
Adati, K.: II 343
Aikawa, H.: I 11, 89, 217, 563
Ainsworth, N.R.: I 575
Akaishi, K.: H 355, 473
Akaoka, H.: I 11, 89, 217, 563
Akasaka, H.: I 11, 89, 217, 563
Akazaki, M.: II 355
Akiba, M.: I 11, 89, 217, 563
Akino, N.: I 11, 89, 217, 563
Akiyama, R.: 1309, 495
Akiyama, T.: I 11, 89, 217, 563
Akulina, D.K.: II 481
Alexander, K.F.: I 237
Alikaev, V.V.: I 111, 237
Allen, J.: I 227
Allen, S.L.: II 231
Alper, B.: II 399
Altmann, H.: I 31, 449
Amano, T.: 1309; II 27, 413
Amemiya, A.: II 355
Amemiya, H.: II 355
Amiranoff, F.: Ill 79
An, Z.G.: II 663
Anderson, P.: I 159
Anderson, R.J.: I 31, 449
Ando, A.: I 553
Ando, R.: I 309, 495
Ando, T.: I 11, 89, 217, 563;
III 357
Andre", A.: Ill 3

Andrews, P.: II 343
Andryukhina, E.D.: II 481
Annoh, K.: I 11, 89, 217, 563
Antoni, V.: II 399, 441
Antonsen, T.M., Jr.: I 407
Aoki, T.: II 343
Aoyagi, N.: I 11, 89, 217, 563
Apte, P.: Ill 55
Aragones, J.M.: III 153
Arakawa, K.: I 11, 89, 217, 563
Araki, M.: I 11, 89, 217, 563
Araki, Y.: Ill 177
Arbez, J.: I 31, 449
Arimoto, H.: II 413
Arimoto, K.: I 11, 89, 217, 563
Armstrong, W.T.: II
Arunasalam, V.: I 51, 75, 171, 207,
421
Arzhannikov, A.V.: II 323
Asakura, N.: II 433
Asami, N.: Ill 343
Askinasi, L.G.: I 607
Astrelin, V.T.: II 323
Audebert, P.: Ill 47
Austin, G.E.: I 227
Avrorov, A.P.: II 323
Axon, K.B.: I 227
Aydemir, A.: II 149
Azechi, H.: ffl 25, 33
Azumi, M.: 111, 89,217, 563; II27;
IE 267Baba, T.: II 355
Bagdasarov, A.A.: 1111
Bagryanskij, P.A.: Ill 467
603

604

AUTHOR INDEX

Bailey, W.: I 31, 449
Baker, D.A.: II 423
Baldis, H.: in 79
Baldock, P.: II 573
Baldwin, D.E.: II 231, 293
Baransky, Y.: I 299
Bard, W.D.: H 655
Barnes, C.W.: II 519
Barnes, D.C.: II 673
Barnouin, O.: Ill 47
Barnsley, R.: I 227
Barter, J.D.: H 231
Bartlett, D.V.: I 31, 65, 99, 449
Bassett, D.: Ill 55
Batanov, G.M.: II 481
Batchelor, D.B.: II 463
Bateman, G.: Ill 259
Batty, F.: I 159
Batyuk, V.A.: II 481
Baumler, I.: 1461
Baxi, C : I 159
Bay, H.L.: 1249, 485
Beasley, C O . , Jr.: II 391
Beaumont, B.: I 31, 449
Beaumont, G.: I 31, 449
Becker, G.: I 125, 357, 461
Behringer, K.: I 31, 65, 197, 449
Beiersdorfer, P.: I 477
Beklemishev, A.D.: II 127
Bell, A.R.: III 55
Bell, M.G.: I 51, 75, 171, 207, 421
Bell, R.: 1477, 503
Bellan, P.M.: I 407
Bender, S.E.: I 607
Bengtson, R.D.: I 187
Berezhetskij, M. S.: II 481
Berezhnyj, V.L.: II 497
Berezovskij, E.L.: I 111, 237
Berk, H.L.: H 135, 719
Berlizov, A.B.: 1111
Bernabei, S.: 1477, 503
Berthier, E.: Ill 3
Bertoldi, P.: I 31, 449

Bertolini, E.: I 31, 449
Bertschinger, G.: I 249, 485
Berzins, L.V.: II 231
Besedin, N.T.: II 497
Besen, M.: I 139
Besshou, S.: II 355, 473
Besson, G.: I 617
Best, C.H.: 131, 449
Bevir, M.K.: H 399
Bhatnagar, V.: I 31, 99, 449
Bhattacharjee, A.: I 299; II 711
Bickerton, R.J.: I 31, 65, 99, 449
Bieger, W.: I 249, 485
Biglary, H.: II 119
Bishop, C M . : 1337
Biskamp, D.: n 11; IE 447
Bitter, M.: I 51, 75, 171, 207, 421
Blackfield, D.: HI 259
Blackwell, B.D.: II 511
Blanchard, W.R.: I 51, 75, 207, 421
Blokh, M.A.: H 481
Bobrovskij, G.A.: I 111
Bodin, H.A.B.: II 399
Bodner, S.E.: IH 47, 527
Boedo, J.A.: I 187
Boehly, T.: Ill 25
Bogdanova, N.E.: I 607
Bogen, P.: I 249, 485
Boghosian, B.M.: II 283
Boileau, A.: 1 3 1 , 197,449
Bol, K.: I 275
Bolton, H.R.: HI 367
Bombarda, F.: I 31, 197, 449
Bonicelli, T.: 131, 449
Bonoli, P.: 1509
Booth, S.: I 31, 449
Boozer, A.H.: II 711
Borshchegovskij, A.A.: I 111, 533
Bosch, H.S.: I 125, 357, 461
Boschi, A.: I 31, 449
Bosia, G.: 131, 449
Botman, M.: I 31, 449
Bottollier-Curtet, H.: I 525

AUTHOR INDEX

Bourgade, J.L.: in 113
Boyd, D.A.: I 387
Brambilla, M.: I 461
Bramson, G.: I 159
Brau, K.: II 251
Braun, F.: I 461
Bravenec, R.V.: I 187
Brejzman, B.N.: II 323
Brelen, H.: I 31, 449
Bretz, N.L.: I 51, 75, 207
Breun, R.A.: II 263
Briand, J.: HI 79
Briffod, G.: I 525
Brinkschulte, H.: I 31, 449
Brocken, H.: I 125, 357, 461
Bromberg, L.: i n 259
Brooker, P.: II 263
Brooks, N.: I 159
Brooks, R.D.: II 673
Brouchous, D.: II 263
Brower, D.L.: I 187
Brown, B.: I 159
Brown, T.: IH 259
Browne, M.L.: I 31, 449
Browning, J.: II 263
Brusati, M.: I 31, 325, 449; II 187
Brussau, W.: 1249, 485
Buchenauer, C.J.: II 423
Biichl, K.: I 125
Budd, T.: 131, 449
Budny, R.: I 51, 75, 207, 421
Buffa, A.: H 441
Bunting, C.A.: II 399
Burdakov, A.V.: H 323
Bures, M.: HI 457
Buresi, E.: HI 3
Burkhardt, L.C.: II 423
Burley, B.: I 159
Burmasov, V.S.: U 323
Burrell, K.H.: I 159, 373
Burtsev, V.A.: UI 309
Bush, C.E.: I 51, 75, 171, 207, 421
BushneU, C : HI 259

605

Bussac, M.N.: II 17
Butcher, P.: I 31, 449
Butterworth, G.J.: III 293
Buttgereit, H.: I 31, 449
Butz, G.: II 263
Byers, J.A.: H 293
Cacaut, D.: I 31, 449
Caldwell-Nichols, C : I 31, 449
Callen, J.D.: I 31, 51, 75, 99, 421,
449; II 157
Callis, R.: I 159
Camacho, F.: I 139
Campbell, D.: Ill 457
Campbell, G.A.: I 31, 99, 159, 249,
433, 449, 485
Cao, Y.: 1249, 485
Caramana, E.J.: II 673
Carlstrom, T.: I 159; II 655
Carolan, P.G.: II 399
Carraro, L.: n 441
Carrera, R.: II 149
Carreras, B.A.: II 53, 157, 615, 663
Carter, M.R.: II 231
Cartwright, D.C.: III 69
Carwardine, J.: I 31, 449
Casey, J.: n 251
Casper, T.A.: II 231
Cavallo, A.: 1477, 503
Cayton, T.E.: II 423
Celata, C : III 129
Celentano, G.: I 31, 449
Celliers, P.: IE 187
Chabert, P.: I 525
Challis, C D . : I 31, 99, 449; II 573
Chance, M.: 1275, 421; II 37
Chankin, A.V.: I 237
Charlton, L.A.: II 53, 615
Cheetham, A.: I 31, 65, 449; UI 457
Chen, Jiayu: I 317
Chen, Kuoin: I 509
Chen, L.: II 119
Chen, X.: II 251
Chen, Y.J.: II 293

606

AUTHOR INDEX

Chen, Z.X.: II 413
Chenais-Popovics, C : III 55
Chernyshev, F.V.: I 607
Chiba, S.: I 11, 89, 217, 563
Chicherov, V.M.: 1237
Chikunov, V.V.: II 323
Chist'akov, V.V.: I 533
Chmyga, A.: 1477
Cho, S.: I 653
Cho, T.: I 553; II 243
Choi, P.: II 573; III 367
Chrien, R.E.: II 551, 673
Christiansen, J.P.: I 31, 99, 449
Christodoulopoulos, C : I 31, 449
Chu, C : II 655
Chu, C.K.: I 299; II 711
Chu, Mingsheng: I 373; II 45
Chu, T.K.: I 503
Chuilon, P.: I 31, 449
Citrolo, J.: Ill 259
Claesen, R.: 131, 449
Clarke, J.F.: I 3
Clauser, J.F.: II 231
Clement, M.: I 525
Coad, J.P.: I 31, 449
Coensgen, F.H.: II 231; III 333
Cohen, R.H.: II 283
Cohen, S.A.: I 31, 51, 75, 207, 421,
449, 477; III 457
Cohn, D.: Ill 259
Cole, A.J.: III 55
Coleman, J.: II 251
Colestock, P.: I 477, 503; HI 259
Colleraine, A.: I 159
Collins, G.A.: 1617
Collins, P.R.: I 575
Combis, P.: Ill 113
Combs, S.K.: I 51, 75, 171, 207, 421
Conn, R.W.: 1249, 485
Conrad, J.: II 263
Conrads, H.: 1485
Cooke, M.: I 31, 449
Cooper, W.A.: II 53, 135

Coppi, B.: 1337; III 397
Coppins, M.: II 573
Corbett, J.: Ill 55
Cordey, J.G.: I 31, 65, 99, 449;
III 457
Core, W.: I 31, 449
Correll, D.L.: II 231
Corti, S.: I 31, 99, 449; III 457
Costa, S.: II 441
Costley, A.E.: 131, 99, 449
Cottet, F.: Ill 79
Cottrell, G.A.: I 31, 99, 449
Coudeville, A.: Ill 3
Coutant, J.: III 3
Couture, P.: I 275
Cox, M.: 131, 449
Crawford, J.D.: II 337
Craxton, R.S.: III 47
Crume, E.C., Jr.: II 391
Cummings, J.: I 159
Cummins, W.F.: II 231, 283
Cunnane, J.: II 399
Czekaj, S.: II 593
D'Ippolito, D.A.: II 263
D'yachenko, V.V.: 1607
Da Silva, L.: m 187
Dagazian, R.Y.: II 639
Dahlburg, J.P.: II 639; III 47
Daido, H.: Ill 407
Dairaku, M.: I 11, 89, 217, 563
Dales, E.: II 263
Damm, C.C.: II 231
Dangor, A.E.: II 573; III 367
Datlov, J.: 1533
Dautray, R.: HI 3
David, C : 131, 449, 525
Davis, J.: I 159
Davis, S.L.: I 51, 75, 207, 421
de Mascureau, J.: Ill 113
DeSimone, P.: II 441
Dean, J.: 131, 449
DeBoo, J.: I 159
Decroisette, M.: Ill 3

AUTHOR INDEX

Deeney, C : II 573
DeGrassie, J.S.: I 187
Deksnis, E.: I 31, 449
Delettrez, J.: Ill 47
Dellis, A.N.: I 575
Delmare, C : III 3
Delvigne, T.: I 249, 485
Deniz, A.V.: I 299
Denne, G.B.: 131, 197, 449
Denton, R.E.: I 387
Denus, S.: II 593
Derfler, H.: 1461
Descamps, P.: I 249, 485
Deschamps, G.: I 31, 449
Dewar, R.L.: II 511
Diamond, P.H.: I 187, 345; II 615, 663
Dietrich, L.: 1237
Dietz, K.J.: I 31, 449
Dikij, A.G.: II 497
DiMarco, J.N.: II 423
Dimock, D.L.: I 51, 75, 207, 421
Dimonte, G.: II 231
Dippel, K.H.: 1249, 485
Dnestrovskij, Yu.N.: II 89
Dobbing, J.: I 31, 449
Dolgaev, S.I.: II 481
Dominguez, N.: II 135
Dominguez, R.R.: I 345
Dommaschk, W.: II 383
Dong, J.Q.: I 387
Donskaya, N.P.: II 481
Doolen, G.: II 639
Dorling, S.E.: 131, 449
Dorokhov, V.V.: I 645
Doublet III Physics Group: I 587
Downing, J.N.: II 423
Drake, J.F.: I 387
Drake, J.R.: E 647
Dubois, P.F.: II 283
Duborgel, B.: Ill 3
Duchs, D.F.: I 31, 325, 449
Duesing, G.: I 31, 99, 449
Dunstan, M.: I 227

607

Duperrex, P.A.: 131, 433, 449
Duquenoy, H.: I 31, 449
Durodie, F.: I 249, 485
Dutch, M.J.: I 635
Duval, B.P.: 1617
Dyabilin, K.S.: II 481
Dylla, H.F.: I , 51, 75, 171, 207, 421
Eberhagen, A.: I 125, 357, 461
Ebiswa, N.: I 11, 89, 217, 563
Eckhartt, D.: 1461, 495
ECRH Group: II 369, 603
Edlington, T.: I 575
Edwards, A.W.: I 31, 227, 433, 449;
III 457
Efremov, S.L.: 1111
Efthimion, P.C.: 151, 75, 171,
207, 421
Eggleston, D.L.: II 337
Ehrenberg, J.: I 31, 449; III 457
Eidmann, K.: Ill 25
Ejima, S.: I 159, 587
Elfimov, A.G.: I 645
Eliezer, S.: Ill 301
Emaami, M.: I 413
Emerson, L.C.: I 51, 75, 207, 421
Emery, M.: IE 47
Empson, K.L.: I 187
Engelhardt, W.: I 31, 197, 449
Engelmann, F.: Ill 223
Englade, R.: Ill 397
England, A.C.: I 51, 75, 207, 421
Eninger, J.E.: II 647
Epperlein, E.M.: III 55
Epstein, R.: IE 47
Equipe TFR: I 179, 599
Erents, S.K.: I 31, 449
Erhorn, F.: 131, 449
Erickson, R.M.: E 423
Eriksson, B.T.: 131, 449
Esch, H.P.L. de: I 599
Esipchuk, Yu.V.: I 111
Esser, H.G.: 1249, 485
Eubank, H.P.: I 51, 75, 207, 421

608

AUTHOR INDEX

Evanko, R.: I 159
Evans, D.E.: II 399
Evans, K.D.: I 227
Evans, P.M.: IE 55
Evans, T.E.: I 187
Evrard, M.: I 31, 449
Fabbro, R.: Ill 79
Fabre, E.: in 79
Fahrbach, H.-U.: I 461
Failor, B.H.: II 231
Falabella, S.: II 231
Faltens, A.: IE 129
Falter, H.: I 31, 449
Fan, Shuping: I 187
Faral, B.: Ill 79
Favre-Dominguez, M.B.: II 573
Fedorenko, S.I.: II 481
Fedyanin, O.I.: II 481
Feldbacher, R.: HI 373
Felt, J.: I 503
Fernandez, J.C.: II 519
Ferron, J.: II 263
Fessenden, T.J.: III 129
Fessey, J.: I 99
Fews, A.P.: Ill 55, 79
Field, A.R.: H 399
Fielding, S.J.: 1227
Fietz, W.A.: HI 279
Figura, E.: II 573
Finken, K.H.: I 249, 485
Finn, J.M.: 1407
Fiore, C : I 139, 509
Fishman, H.: 1275
Flanagan, C : III 259
Flor, G.: II 441
Flora, F.: II 441
Flyagin, V.A.: I 111
Foden, N.: I 31, 449
Folkierski, A.: II 573
Fomin, I.P.: II 497
Fonck, R.J.: I 51, 75, 207, 275, 421
Foord, M.: I 139
Foote, J.H.: H 231

Forman, P.R.: II 423
Forrest, M.J.: 1 3 1 , 197, 449
Forsley, L.: HI 47
Forster, J.C.: I 187
Fowler, T.K.: II 231
Francis, G.L.: II 263
Fredrickson, E.: I 51, 75, 171,
207, 421
Friedman, A.: II 283
Fritsch, R.: 1461
Froger, C : I 31, 449
Fu, G.Y.: II 135
Fuchs, G.: 1249,485
Fujii, T.: I 11, 89, 217, 563
Fujisawa, A.: II 433
Fujisawa, N.: IE 231, 267
Fujita, H.: H 343; HI 407
Fujita, J.: II 489
Fujita, M.: ffl 407
Fujita, T.: II 433
Fujiwara, E.: Ill 33
Fujiwara, M.: II 489
Fukao, M.: 1553
Fukuda, T.: I 11, 89, 217, 563
Fukuda, Y.: II 433
Fukumoto, H.: I 159
Fukutomi, M.: m 351
Fukuyama, A.: I 627
Fullard, K.: I 31, 449
Funahashi, A.: I 151, 545
Furth, H.P.: 151, 75, 207,421
Furukawa, H.: I 11, 89, 217, 563
Fussmann, G.: I 31, 197, 125, 357,
449, 461
Futch, A.H.: II 231
Gabellieri, L.: H 441
Gadeberg, M.: I 31, 99, 449
Gago, J.A.: ffl 153
Galetsas, A.: I 31, 449
Gallacher, A.: I 31, 449
Gallix, R.: I 159
Gambier, D.: I 31, 449
Gamez, L.: m 153

AUTHOR INDEX
Gammel, G.: I 51, 75, 207, 275,
421, 477
Gandy, R.: I 139
Garcia, L.: II 615, 663
Gardner, H.J.: II 511
Gardner, J.: Ill 47
Garner, H.R.: II 251, 343
Gazit, Y.: Ill 187
Gee, S.J.: II 399
Gehre, O.: I 125, 357, 461
Gentle, K.W.: I 187
Gerhardt, J.: I 125, 357, 461
Gerich, J.W.: III 333
Gerver, M.: II 251
Giannella, R.: I 31, 197, 449
Giannelli, A.: I 31, 449
Gibson, A.: I 31, 65, 99, 449;
ni479
Gierke, G. von: I 125, 357, 461
Giesen, B.: 1249, 485
Gill, R.D.: I 31, 433, 449; III 457
Gilleland, J.: I 159
Gimblett, C.G.: II 399
Girard, A.: I 525
Giubbilei, M.: II 441
Giudicotti, L.: II 441
Glad, T.: I 159
Glagolev, V.M.: HI 421
Glock, E.: I 125, 357, 461
Goebel, D.M.: 1249, 485
Goede, A.: I 31, 449
Goforth, R.R.: II 655
Gohil, P.: I 159
Gol'denberg, A.L.: 1111
Golant, V.E.: I 607
Goldston, R.J.: I 51, 75, 171,
207, 421
Goldstone, P.D.: III 69
Golovato, S.: II 251
Gomez, C : I 139
Goncharov, S.G.: I 607
Gondhalekar, A.: I 31, 65, 99, 449;
III 457

609

Gonzalez, M.C.: III 153
Goodall, D.H.J.: 1227
Goodman, D.: II 251
Goodman, R.K.: II 231
Gootgeld, A.: I 159
Gorbunov, E.P.: 1111
Gordon, C : I 31, 449
Gorelov, Yu.A.: I 111, 533
Gormezano, C : I 31, 449, 525
Goto, S.: II 529, 551
Gottardi, N.A.: I 31, 65, 99, 197,
325, 449
Goulding, R.: II 263
Gowers, C.W.: I 31, 99, 433, 449
Graffmann, E.: 1249, 485
Granetz, R.: I 31, 139, 433, 449,
509; III 457
Grassie, K.: 1357
Grebenshchikov, S.E.: II 481
Green, B.: I 31, 449
Greene, J.M.: I 373, 477; II 45
Greenfield, C M . : II 739
Greenwald, M.: I 139; III 259
Gregoli, S.: I 31, 449
Grek, B.: I 51, 75, 171, 207, 421
Griffin, D.: 1509
Griph, F.S.: I 31, 449
Grisham, L.R.: 151, 75, 171,
207, 421
Groebner, R.J.: I 159, 373
Grossman, A.A.: I 299
Grote, H.: 1237
Grubb, D.P.: II 231
Gruber, O.: I 125, 357, 461
Grun, J.: Ill 47
Gryaznevich, M.P.: I 607
Guasp, J.: II 633
Gubarev, S.P.: II 497
Guillaneux, P.: Ill 3
Gulasaryan, N.L.: I 645
Gunther, K.: I 237
Guo, S.G.: II 135
Guo, Wenkang: I 317

610

AUTHOR INDEX

Guss, W.: II 251
Gutarev, Yu.V.: II 497
Guzdar, P.N.: I 387
Gwinn, D.: I 139, 509
Haange, R.: I 31, 449
Haas, G.: I 125, 187, 357, 461
Haberstich, A.: II 423
Hahm, T.S.: I 187, 345; II 663
Hahn, K.D.: II 673
Haines, M.G.: II 573;. Ill 55, 367
Hallouin, M.: HI 79
Hamada, Y.: I 309, 495; II 27
Hamamatsu, K.: I 11, 89, 217,
563, 627
Hamamoto, M.: II 355
Hamasaki, S.: II 541
Hamberger, S.M.: II 511
Hameiri, E.: II 711
Hammett, G.: I 51, 75, 171, 207, 421
Hamnen, J.H.: I 31, 449
Han, W.: I 227
Hanai, S.: I 159
Hanatani, K.: II 355
Hancock, C.J.: I 31, 449
Hancox, R.: HI 293, 535
Hanlon, J.A.: in 69
Hara, M.: I 11, 89, 217, 563
Harada, M.: HI 161
Harafuji, K.: E 65
Haraguchi, K.: I 11, 89, 217, 563
Harbour, P.J.: 131, 65,449
Harned, D.S.: H 673, 701
Hartwig, H.: 1249,485
Harvey, R.W.: I 587; II 45
Hasegawa, A.: II 625; III 407
Hasegawa, K.: I 151, 545
Hasegawa, M.: n 687
Haskovec, J.: I 159
Hassam, A.B.: I 387
Hasselberg, G.: II 167
Hastings, D.E.: II 135
Hatori, T.: II 343
Hattori, H.: H 433

Haubenreich, P.N.: III 279
Hauer, A.: ffl 69
Hawkes, N.C.: I 31, 227, 449
Hawryluk, R.J.: I 51, 75, 171,
207, 421
Hayase, K.: H 563
Hayashi, K.: I 11, 89, 217, 563
Hayashi, T.: II 65, 529
Hayden, R.: II 399
Haynes, P.: I 31, 449, 575
Hazeltine, R.D.: H 149
He, Yexi: I 187
Heckman, E.: I 159
Hedin, E.R.: H 739
Heiberger, M.: I 159
Heidbrink, W.W.: I 51, 75, 171,
207, 275, 421
Heifetz, D.B.: 175
Heindler, M.: HI 373
Hellermann, M. von: I 31, 99,
197, 449
Hellsten, T.: I 31, 325, 449
Helton, F.J.: I 159, 373; II 45
Hemmerich, J.L.: I 31, 449
Hemsworth, R.: I 31, 449
Hendel, H.W.: 151, 75, 171,
207, 421
Hender, T.C.: I 291; II 53
Hendriks, F.: I 31, 99, 449;
IH457
Henins, I.: II 519
Henline, P.: I 159
Henshaw, M.J.: IH 55
Herrmann, P.: Ill 25
Herrnegger, F.: U 383
Hershkowitz, N.: II 263
Herzog, R.F.: I 31, 449
Hess, W.: I 525
Hesse, M.: 1525
Hickok, R.L.: I 187
Hidekuma, S.: II 343
Hill, D.: I 159; H 231
Hill, K.W.: I 51, 75, 171, 207, 421

AUTHOR INDEX

Hillis, D.L.: II 355
Hinnov, E.: I 51, 75, 207, 421
Hino, T.: H 355
Hinton, F.L.: II 221
Hintz, E.: I 249, 485
Hiraki, N.: ffl 321
Hirano, K.: II 583
Hirano, Y.: II 453
Hiratsuka, H.: I 11, 89, 217, 563
Hirayama, T.: I 11, 89, 217, 563;
HI 267
Hiroe, S.: I 51, 75, 207, 421; II 251
Hiroki, S.: I 11, 89, 217, 563
Hirokura, S.: I 309, 495
Hirose, A.: 1413
Hirota, I.: E 563
Hirshman, S.P.: II 391
Hiruta, K.: I 11, 89, 217, 563
Hively, L.: I 653
Hiyama, T.: HI 357
HL-1 Tokamak Group: I 399
Hoang, G.T.: I 525
Hodge, W.L.: I 187
Hoekzema, J.A.: 1599
Hoenen, F.: I 249, 485
Hoffman, A.L.: H 541
Hoffman, D.: I 159
Hoffmann, E.: I 159
Hofmann, J.: I 125, 357, 461
Hofmeister, F.: 1461
Hogan, J.T.: H 53
Hoida, H.W.: U 519
Hokin, S.: II 251
Holl, P.M.: HI333
Holland, A.: 1299
Hollenstein, C : I 617
Holmes, J.A.: H 53, 615, 663
Honda, A.: I 151, 545
Honda, C : II 489
Honda, M.: I 11, 89, 217, 563
Honda, Y.: H 529
Hong, R.: I 159
Honrubia, J.J.: ffl 153

611

Hooke, W.: 1503
Hooper, E.B.: II 231
Horiike, H.: I 11, 89, 217, 563;
IH 267
Horioka, K.: ffl 161
Horiuchi, R.: II 693
Hornady, R.S.: II 231
Home, S.: H 251
Horton, L.: 131, 99, 449
Horton, R.D.: II 231
Hosea, J.: 1477, 503
Hoshino, K.: I 151, 545
Hoshino, M.: ffl 357
Hosoda, R.: I 11, 89, 217, 563
Hosogane, N.: I 11, 89, 159,
217, 563
Hosokawa, M.: n 489
Hothker, K.: I 249, 485
Houlberg, W.: I 653; ffl 259, 457
How, J.: I 31, 449
Howald, A.M.: II 343
Howe, H.C.: II 355
Howell, R.B.: II 423
Howling, A.A.: I 617
Hsieh, C : I 159
Hsu, J.Y.: 1373
Hsu, W.L.: 1249, 485; II 231
Hsuan, H.: I 51, 75, 171, 207, 421
HT-6B Group: I 317
HT-6M Group: I 317
Huart, M.: 131, 449
Hubbard, A.: 131,449
Hubner, K.: I 461
Huff, B.: II 231; ffl 333
Hughes, M.: 1299; HI 259
Hugill, J.: I 31, 227, 449
Hugon, M.: I 31, 433, 449; II 187
Hugrass, W.N.: II 673
Huguet, M.: 131, 449
Hulse, R.A.: I 51, 75, 207, 421
Hunt, A.L.: II 231
Huo, Yuping: I 317
Hutchison, R.J.: HI 47

612

AUTHOR INDEX

Hwang, D.Q.: I 653
Ichiguchi, K.: II 625
Ichimura, M.: II 243
Ichtchenko, G.: I 525
Ida, K.: I 275, 309, 495
Ide, S.: I 553
Ido, S.: HI25, 33, 343
Ignat, D.: Ill 259
Iguchi, H.: II 489
Iida, M.: I 553
Iida, Y.: I 11, 89, 217, 563
Iijima, T.: I 11, 89, 217, 563
lima, M.: II 355
Iiyoshi, A.: II 355, 473
Ijiri, Y.: n 355, 473
Ikeda, K.: I 11, 89, 217, 563
Ikeda, Y.: I 11, 89, 217, 563
Ikegami, H.: II 489
Ikezi, H.: U 343
Il'in, V.I.: I 111, 533
Imai, T.: I 11, 89, 217, 563
Imasaki, K.: HI 103
Imre, K.: II 463
Ingraham, J.C.: II 423
Ingram, B.: I 31, 449
Innocente, P.: II 441
Inoue, K.: IH 357
Inoue, N.: II 433
Inoue, T.: I 11, 89, 217, 563
Intrator, T.: II 263
Inutake, M.: D 243
Irby, J.: II 251
Isaji, N.: I 11, 89, 217, 563
Isaka, M.: I 11, 89, 217, 563
Ishibori, I.: I 151, 545
Ishida, A.: II 719
Ishida, S.: I 11, 89, 217, 563; H 433
Ishii, K.: I 553; II 243
Ishimura, T.: II 551
Isler, R.C.: I 187
Itakura, A.: II 243
Itami, K.: H 433
Itatani, R.: I 553; H 109, 273

Itige, N.: I 11, 89, 217, 563
Ito, H.: H 551
Ito, T.: I 11, 89, 217, 563
Ito, Y.: I 11, 89, 217, 563; II 551
Itoh, K.: I 627; II 197, 473, 625
Itoh, S.: IU 321
Itoh, S.-L: I 627; II 197
Its, E.R.: I 607
Ivanov, A.A.: Ill 467
Ivanov, D.P.: I 533
Ivanov, N.V.: I 533
Ivers, T.: I 299
Iwata, Y.: II 433
Iyengar, S.: I 575
Izawa, Y.: Ill 33
Izvozchikov, A.: I 125, 357,
461, 607
Izzo, R.: Ill 259
Jaanimagi, P.: Ill 47
Jackel, H.: I 31, 99, 197, 449;
III 457
Jackson, G.L.: I 159, 187
Jacobs, S.: HI 47
Jacquinot, J.: I 31, 99, 449
Jadoul, M.: I 249, 485
Jaeckel, H.: I 65
Jaeger, E.F.: H 463
Jaehnig, K.P.: I 51, 75, 207,
275, 421
Jahns, G.: I 159, 275
Jahoda, F.C.: U 423
Jain, K.: II 441
James, R.A.: n 231
Janeschitz, G.: I 125, 159, 357, 461
Jankowicz, Z.: I 31, 449
Jarboe, T.R.: II 399, 519
Jardin, S.: IH 259
Jarvis, O.N.: I 31, 99, 449
Jassby, D.: I 51, 75, 207, 421
Javadi, M.: Ill 367
JET Team: I 31, 449
Ji, H.: II 433
Jitsuno, T.: IH 25, 33

AUTHOR INDEX
Jobes, F.C.: I 51, 75, 207, 421, 503
Johnson, D.W.: 151, 75, 171,
207, 421
Johnson, E.: I 159
Johnson, I.: I 575
Johnson, L.C.: I 51, 75, 171,
207, 421
Johnson, P.C.: I 227
Jones, E.M.: I 31, 449
Jones, P.: I 31, 449
Jones, T.T.C.: 1 3 1 , 99, 449;
III 457
Jotaki, E.: Ill 321
Joye, B.: 1617
JT-60 Team: I 11, 89, 217, 563
Judd, D.L.: III 129
Jupen, C : 131, 197, 449
Kadota, K.: II 489
Kadoya, Y.: II 109
Kaiser, T.B.: II 293
Kaita, R.: 1 5 1 , 75, 207, 275,
421, 477
Kako, E.: I 309, 495
Kako, M.: II 551
Kakurin, A.M.: I 533
Kaleck, A.: I 249, 485
Kallne, E.: I 31, 99, 197,449
Kallne, J.: I 31, 99, 449
Kamada, Y.: U 413, 433
Kamimura, T.: II 489
Kaminaga, A.: I 11, 89, 217, 563
Kamperschroer, R.: I 51, 75, 207,
421

Kandaurov, I.V.: II 323
Kaneko, H.: II 355, 473
Kaneko, O.: I 31, 309, 449, 495
Kapitonov, V.A.: II 323
Karger, F.: I 125, 187, 357, 461
Karpenko, V.P.: III 333
Kasai, S.: I 151, 545
Kasai, T.: Ill 167
Kashimura, T.: I 151, 545
Kashiwa, Y.: I 151, 545

613

Kasperczuk, A.: II 593
Kasuya, K.: Ill 161
Katanuma, I.: II 243
Kato, S.: II 355; III 161
Kato, T.: Ill 357
Kato, Y.: n 529; III 33
Kaufmann, M.: I 125, 357
Kaw, P.K.: II 101
Kawabata, T.: II 355, 473
Kawahata, K.: I 309, 495
Kawai, M.: I 11, 89, 217, 563
Kawakami, T.: I 151, 545
Kawamata, Y.: I 11, 89, 217, 563
Kawamoto, T.: II 343
Kawano, K.: El 357
Kawasaki, K.: I 11, 89, 217, 563
Kawasaki, S.: Ill 321
Kawashima, H.: I 151, 545
Kawasumi, Y.: I 309, 495
Kawata, S.: Ill 177
Kaye, A.: I 31, 494
Kaye, S.M.: I 51, 75, 207, 275,
337, 421
Kazawa, M.: I 151, 545
Keck, R.L.: III 47
Keefe, D.: HI 129
Keen, B.E.: 131, 449
Keilhacker, M.: I 31, 65, 99, 125,
357, 449, 461; III 457
Kellman, A.: I 159
Kerbel, G.: Ill 259
Kernbichler, W.: Ill 373
Kesner, J.: II 251
Kessler, T.: Ill 47
Kever, H.: I 249, 485
Khalilov, M.A.: 1607
Khil'chenko, A.D.: H 323
Kho, T.H.: in 55
Kholnov, Yu.V.: II 481
Khvostenko, P.P.: I 533
Kikuchi, K.: I 11, 51, 89, 217,
545, 563
Kikuchi, M.: I 11, 89, 217, 563

614

AUTHOR INDEX

Kilpatrick, S.J.: I 51, 75, 207, 421
Kim, C.H.: III 129
Kim, H.: HI 47
Kim, J.S.: I 159, 373
Kim, S.K.: I 187
Kimura, H.: I 11, 89, 217, 563
Kimura, T.: I 11, 89, 217, 563
Kinahan, G.: I 31, 449
Kirby, P.: II 3, 399
Kirov, A.G.: 1645
Kishimoto, H.: I 11, 89, 217, 563
Kishimoto, Y.: I 627; II 27
Kislov, A.Ya.: I 111, 533
Kissel, S.: 131, 99, 449
Kitagawa, S.: I 309, 495
Kitagawa, Y.: II 583; III 407
Kitahara, K.: I 11, 89, 217, 563
Kitamura, S.: I 11, 89, 217, 563
Kito, M.: II 563
Kitsunezaki, A.: I 11, 89, 217, 563
Kiwamoto, Y.: II 243
Kiyama, H.: II 563
Kiyama, S.: II 563
Kiyono, K.: I 11, 89, 217, 563
Klepper, C.C.: I 187
Klesov, V.V.: ffl 467
Kleva, R.G.: I 387
Klieb, L.: I 659
Kliiber, O.: I 125, 357, 461
Knize, R.J.: I 51, 75, 207, 421
Knowlton, S.: I 139, 509
Knox, S.O.: H 519
Knyazev, B.A.: II 323
Kobayashi, N.: I 11, 89, 217, 563
Kobayashi, S.: II 355
Kobayashi, T.: Ill 267
Koch, R.: 1249, 485
Kochin, V.A.: 1533
Kock, L. de: I 31, 99, 449
Kodama, K.: I 11, 89, 217, 563
Koepke, M.E.: H 739
Koga, T.: II 355
Kohli, J.: I 159

Koide, Y.: I 11, 89,217, 563
Koike, T.: I 11, 89, 217, 563
Koizumi, K.: in 357
Kojdan, V.S.: H 323 •Kolik, L.V.: H 481
Komata, M.: I 11, 89, 217, 563
Komeiji, S.: Ill 167
Komori, A.: H 489
Komoshvili, K.G.: I 645
Kondo, I.: I 11, 89, 217, 563
Kondo, K.: H 355, 473
Kondoh, Y.: H 655
Konen, L.: 1249, 485
Kononenko, V.I.: II 497
Konoshima, S.: I 11, 89, 217, 563
Konovalov, V.K.: H 497
Konstantellos, A.: I 31, 449
Konyukhov, V.V.: II 323
Kooijman, W.: I 599
Kopecky, V.: I 533
Kornherr, M.: I 125, 357, 461
Korten, M.: 1249,485
Kostomarov, D.P.: II 89
Kostsov, Yu.A.: I 607
Kotel'nikov, I.A.: H 305; III 467
Kotschenreuther, M.: II 149
Kotsubanov, V.D.: II 497
Kovanen, M.: I 31, 449
Kovrizhnykh, L.M.: II 481
Kovrov, P.E.: I 111, 533
Koyama, K.: H 563; in 167
Koz'minykh, Yu.L.: HI 467
Kozhevin, V.M.: HI 309
Kozman, T.A.: ffl 333
Krall, N.A.: H 541
Krasheninnikov, S.I.: U 177; in 387
Krasnikov, Yu.I.: HI 467
Krause, K.H.: ffl 333
Kravchin, B.V.: n 497
Kremens, R.: ffl 47
Krikunov, S.V.: 1607
Kruer, W.L.: m 137
Kruglyakov, Eh.P.: II 323

AUTHOR INDEX

Kryska, L.: 1533
Ku, L.P.: in 259
Kubo, H.: I 11, 89, 217, 563; II489
Kubo, S.: I 553; n 687
Kugel, H.: I 51, 75, 207, 275, 421
Kukushkin, A.S.: n 177
Kulaga, A.E.: II 497
Kumazawa, R.: II 343
Kunieda, S.: I 11, 89, 217, 563
Kuo-Petravic, G.: ffl 259
Kuphnapfel, U.: 131, 449
Kupschus, P.: 131, 449
Kurakata, S.: I 11, 89, 217, 563
Kurihara, K.: I 11, 89, 217, 563
Kuriyama, M.: I 11, 89, 217, 563
Kuroda, T.: I 11, 89, 217, 309, 495,
563;IE 357
Kusaka, M.: I 11, 89, 217, 563
Kusama, Y.: I 89, 217, 563
Kusano, K.: U 433, 693
Kuwahara, H.: I 413
Kuznetsov, Yu.K.: n 497
Kwok, Y.-C: II 711
La Haye, R.J.: H 655
Labaune, C : EH 79
Labiak, W.C.: HI 333
LaBombard, B.: I 139, 249
Lackner, K.: I 125, 357, 461;
HI 517
Lallia, P.: I 31, 99, 449; H 187
LaMarche, P.H.: I 51, 75, 171,
207, 421
Lamb, M.J.: HI 55
Lane, B.: II 251
Lang, D.D.: HI 333
Lang, R.: I 125
Langhorn, A.: I 159
Lao, L.: I 159, 373
Larionova, N.F.: II 481
Larsen, J.T.: IH 145
Laslett, L.J.: HI 129
Lasnier, C.J.: U 231
Last, J.R.: I 31, 449

615

Lathrop, G.H.: HI 333
Launspach, J.: Ill 3, 113
Lauro-Taroni, L.: I 31, 449
Lawson, K.D.: I 31, 197, 449
Lazarus, E.A.: n 53
Lazzaro, E.: I 31, 65, 99,449
Leavitt, C.A.: II 231; III 333
Lebeau, D.: I 249, 485
Lebedev, S.V.: I 607; II 323
LeBlanc, B.: I 51, 75, 207, 275, 421
Lee, E.P.: HI 129
Lee, G.S.: I 345; II 663
Lee, J.K.: I 357, 373; H 45
Lee, P.: I 159; II 655
Lee, T.: HI 47
Lee, Y.C.: 1387
Lehnert, B.P.: II 647
Lehrman, I.: I 477
Leikind, B.J.: II 343
Lenoci, M.: I 125, 357, 461
Lerbinger, K.: II 17
Lesniyakov, G.G.: II 497
Letzring, S.A.: III 47
Leung, W.K.: 1249, 485
Leuterer, F.: I 461
Lewis, C.L.S.: III 55
Lewis, H.R.: H 673
Li, L.: 1249, 485
Li, X.L.: 1299
Lie, Y.T.: I 249, 485
Lieber, A.J.: 1587
Liewer, P.C.: 1407
Lightner, S.: I 159
Likin, K.I.: I 533
Likin, K.M.: U 481
Lindquist, W.B.: IH 333
Linford, R.K.: II 673
Ling, K.M.: H 673
Lingertat, J.: I 237
Lipin, B.M.: I 607
Lipschultz, B.: I 139, 509; IH 259
Lisitano, G.: I 125, 357, 461
Lister, J.B.: 1617

616
Little, R.: I 51, 75, 217, 421
Litunovskij, R.N.: I 607
Litvinov, A.P.: II 497
Litwin, C : II 263
Liu, C.S.: I 387
Llobet, X.: II 135
Lloyd, B.: I 575
Lobel, R.C.: I 31, 449
LoDestro, L.L.: II 283
Lohr, J.: I 159
Lomas, P.J.: 131, 99, 433, 449
Longinov, A.V.: II 205
Lopes-Cardozo, N.: I 31, 99,433, 449
Lorenz-Gottardi, M.: I 31, 449
Lotz, D.: II 673
Lotz, W.: II 383
Lowry, C : 131, 449
Lozovskij, S.N.: I 645
Lubell, M.S.: Ill 279
Luciani, J.F.: III 79
Luckhardt, S.C.: I 509
Lue, J.W.: III 279
Luhmann, N.C.: I 187
Luhmann, N.C., Jr.: Ill 137
Luk'yanov, S.Yu.: 1111
Luton, J.N.: III 279
Luxon, J.: I 159
Lynch, V.E.: II 615, 663
Lysenko, S.E.: I 237
Ma, C.H.: 151, 75, 207, 421
Maassberg, H.: II 369
Machida, M.: II 551
Macmahon, A.: I 187
Maeda, M.: II 355, 489
Maehara, S.: I 11, 89, 217, 563
Maejima, Y.: II 563
Maekawa, T.: I 553
Maeno, K.: I 11, 89, 217, 563
Magyar, G.: 131, 197, 449;
III 457
Man, R.: Ill 69
Mahdavi, M.: I 159

AUTHOR INDEX

Maisonnier, D.: I 31, 449
Majeski, R.: H 263
Makhankov, A.N.: III 309
Maksimov, Yu.S.: I 111, 533
Malacarne, M.: I 31, 433, 449
Malmberg, J.H.: II 337
Malykh, L.Ya.: 1645
Manickam, J.: I 263, 275, 421;n 37
Manka, C : III 47
Manley, A.: II 399
Manning, H.: I 139
Manos, D.M.: I 51, 75, 207, 421
Mansfield, D.K.: I 51, 75, 171,
207, 421
Mantica, P.: III 457
Marchese, V.: I 31, 449
Marjoribanks, R.S.: III 47
Markelov, V.A.: I 111
Marklin, G.J.: II 519, 639
Marmar, E.: I 139, 509; II 355
Marshall, F.: Ill 47
Marshall, T.C.: I 299
Martens, V.: 165
Martin, P.: II 441
Martinez-Val, J.M.: III 153
Martini, S.: II 441
Masai, K.: 1309,495
Masamune, S.: II 413
Mase, A.: II 243
Mase, S.: I 11, 89, 217, 563
Massey, R.S.: II 423
Massmann, P.: I 31, 449
Mast, F.: I 125, 357, 461
Masugata, K.: Ill 177
Masumoto, H.: II 343
Matoba, T..: I 151, 545
Matsuda, K.: I 587
Matsuda, S.: I 11, 89, 217, 563;
III 267
Matsuda, T.: I 151, 545
Matsuda, Y.: II 283
Matsui, M.: II 687

AUTHOR INDEX

Matsukawa, M.: I 11, 89, 217, 563
Matsukawa, T.: I 11, 89, 217, 563
Matsumoto, H.: I 151, 545
Matsumoto, M.: II 109
Matsumoto, Y.: Ill 167
Matsuoka, A.: II 413
Matsuoka, K.: I 309, 495; II 27
Matsuoka, M.: I 11, 89, 217, 563
Matsushima, I.: Ill 167
Matsuzaki, Y.: I 151, 545
Matsuzuka, S.: II 433
Matthews, G.F.: I 227
Mattor, N.: 1345
Mauel, M.E.: I 299
Mayberry, M.: I 159, 509
Mayer, H.M.: I 125, 357, 461
Mayo, R.M.: II 519
Mazzitelli, G.: I 477
Mazzucato, E.: 1477, 503
McCann, R.T.: I 51, 75, 207, 421
McCarthy, A.L.: I 635
McCarthy, M.P.: I 51, 75, 207, 421
McCool, S.C.: I 139, 187
McCormick, K.: I 125, 357, 461
McCracken, G.M.: I 31, 197, 227, 449
McCrory, R.L.: III 47
McCullen, P.: I 31, 449
McCune, D.C.: I 51, 75, 171, 207, 421
McDermott, F.S.: I 509
McGuire, K.: I 51, 75, 171, 207
McHarg, B.: I 159
McKelvey, T.: I 159
McKenna, K.F.: H 673
McKenty, P.: Ill 47
McLean, E.: Ill 47
McLeod, J.: HI 69
McNeill, D.H.: I 51, 75, 207, 421, 477
Mead, M.J.: 1 3 1 , 449
Mead, W.C.: III 69
Meade, D.M.: I 51, 75, 207, 421
Meassick, S.: II 263
Medley, S.S.: I 51, 75, 171, 207, 421
Medvedev, A.A.: I 111

617

Medvedev, V.A.: 1237
Meisel, D.: I 125, 357, 461
Meiss, J.D.: H 149
Mekler, K.I.: II 323
Mens, A.: Ill 113
Meriguet, P.: I 31, 449
Merkel, P.: II 383
Merlin, D.: II 441
Merlo, V.: I 31, 449
Mertens, V.: I 31, 125, 357, 449, 461;
III 457
Meservey, E.: I 503
Meshcheryakov, A.I.: II 481
Meshkov, O.I.: II 323
Messiaen, A.M.: I 249, 485
Mexmain, J.M.: III 113
Meyer, W.H.: II 231
Michard, A.: Ill 79
Michishita, T.: II 489
Midzuno, Y.: II 27
Migliuolo, S.: HI 397
Mikic, Z.: II 701
Mikkelsen, D.R.: I 51, 75, 207, 421
Miklaszewski, R.: II 593
Miley, G.H.: II 701
Millar, W.: 1227
Miller, G.: II 423
Miller, R.: I 159
Mills, S.: I 31, 449
Millward, P.: I 31, 449
Milora, S.L.: I 51, 75, 171, 207, 421
Mima, K.: Ill 25, 33, 407
Minagawa, H.: II 355
Miner, W.H.: I 187
Minguez, E.: Ill 153
Minyaev, O.A.: I 607
Mioduszewski, P.K.: II 355
Mirin, A.A.: II 639
Mironov, Yu.K.: II 497
Mitarai, O.: 1413
Mitchell, B.: Ill 293
Mitobe, K.: Ill 161
Miura, Y.: I 151, 454

618

AUTHOR INDEX

Miya, N.: I 11, 89, 217, 563
Miyamoto, K.: II 433
Miyamoto, S.: Ill 103
Miyanaga, M.: Ill 33
Miyanaga, N.: Ill 25
Miyati, K.: I 11, 89, 217, 563
Miyo, Y.: I 11, 89, 217, 563
Miyoshi, S.: II 243
Mizuhashi, K.: I 11, 89, 217, 563
Mizuno, M.: I 11, 89, 217, 563
Mizuuchi, T.: II 355, 473
Mochizuki, T.: ffl 25, 33
Moeller, C.P.: I 159, 587
Mohri, A.: I 309, 495, 553; II 687
Moissonnier, A.: I 31, 449
Moller, J.M.: U 231
Molvik, A.W.: II 231
Momota, H.: II 719
Mondino, P.L.: I 31, 449
Mondt, J.P.: n 639
Montgomery, D.: II 639
Montgomery, D.B.: HI 259
Monticello, D.: 1421
Moody, J.: I 509
Moore, D.: I 159
Mora, P.: Ill 79
Moran, T.: II 251
Moreau, D.: I 31, 449
Moret, J.-M.: I 617
Morgan, P.D.: I 31, 99, 197, 449;
III457Mori, M.: I 151,447, 545Mori, S.: Ill 195
Morikawa, J.: II 433
Morimoto, S.: n 355, 473; HI 161
Morioka, M.: Ill 161
Morishita, T.: I 627
Morita, S.: I 495; H 355
Moriyama, S.: Ill 321
Moroz, P.E.: II 481
Morozov, D.Kh.: HI 435
Morris, W.: 151, 75, 171, 207,
275, 421

Morrison, P.J.: II 149
Moses, R.W., Jr.: II 423
Motley, R.: I 503
Motojima, O.: II 355, 473
Mourou, G.A.: III 47
Mueller, D.: I 51, 75, 171, 207, 421
Mugishima, M.: I 309, 495
Mukhovatov, V.: I 51, 75, 207
Muller, E.R.: I 31, 125, 197, 357,
449, 461
Munich, E.: I 461
Munson, C.P.: II 423, 433
Murai, R.: I 11, 89, 217, 563
Murakami, M.: II 355
Murakami, Y.: I 11, 89, 217, 563;
II 433
Muraoka, K.: II 355, 489
Murata, H.: II 413
Murayama, M.: Ill 177
Murmann, D.: I 31, 449
Murmann, H.: I 125, 357, 461
Murphy, G.: I 31, 449
Murphy, J.A.: I 503
Muto, M.: I 11, 89, 217, 563
Mutoh, T.: n 355, 473
Myra, J.R.: II 263
Nagami, M.: I 11, 89, 217, 563
Nagao, A.: Ill 321
Nagashima, A.: I 11, 89, 217, 563
Nagashima, K.: I 11, 89, 217, 563
Nagashima, T.: I 11, 89, 217, 563
Nagata, A.: II 413
Nagata, M.: II 529
Nagaya, S.: I 11, 89, 217, 563
Nagayama, Y.: II 433
Naito, H.: II 135
Naitou, H.: II 27, 625
Nakai, M.: Ill 33
Nakai, S.: Ill 25, 33, 103, 407
Nakajima, H.: IE 357
Nakajima, S.: II 109
Nakamura, H.: I 11, 89, 217, 563
Nakamura, K.: HI 161, 321

AUTHOR INDEX

Nakamura, M.: I 553
Nakamura, Y.: I 11, 89, 217, 563;
III321Nakashima, H.: Ill 343
Nakashima, Y.: II 243
Nakasuga, M.: II 355
Nakatsuka, M.: m 25, 33
Nakayama, Y.: II 27, 65
Narihara, K.: II 687
Nassri, A.: Ill 373
Nave, M.F.: I 31, 433, 449; II 3
Navratil, G.A.: I 275, 299
Nazarov, N.I.: II 497
Nebel, R.A.: II 423, 701
Neilson, G.H.: II 355
Nelson, B.: II 263
Nelson, P.: IH 3
Nelson, R.L.: III 333
Nemoto, F.: Ill 167
Nemoto, M.: I 11, 89, 217, 563
Nemoto, T.: II 27
Nerem, A.: I 159
NET Team: III 249
Nevins, W.M.: II 283
Newton, A.A.: H 399, 453
Nexsen, W.E.: II 231
Neyatani, Y.: I 11, 89, 217, 563
Ng, A.: IH 187
NI Group: II 369, 603
Nickesson, L.: I 31, 449
Niedermeyer, H.: I 125, 357, 461
Nielsen, P.: I 31, 99, 449;
III 457
Nieschmidt, E.B.: I 51, 75, 171,
207, 421
Nihei, H.: II 433
Niikura, S.: I 11, 89, 217, 563
Niki, H.: Ill 33
Nilles, E.J.: II 655
Ninomiya, H.: I 11, 89, 217, 563
Nishi, M.: Ill 357
Nishiguchi, A.: Ill 33
Nishihara, K.: HI 25, 33, 103, 407

619

Nishikawa, M.: II 529; III 343
Nishimura, H.: Ill 25, 33
Nishimura, K.: II 343
Nishio, S.: Ill 267
Nishitani, T.: I 11, 89, 217, 563
Noda, N.: I 309, 459; II 355, 473
Nogi, Y.: II 551
Noll, P.: I 31, 65, 159, 449
Nomata, H.: I 11, 89, 217, 563
Nomura, Y.: II 719
Noonan, P.G.: II 399
Norimatsu, T.: Ill 25, 33
Noterdaeme, J.-M.: I 461
Notkin, G.E.: I 111
Nowak, S.: 131, 65, 449
Niihrenberg, J.: n 383
O'Brien, M.: I 575
O'Neil, T.M.: II 337
O'Rourke, J.: 131, 51, 75, 99,
171, 207, 421, 449; III 457
Obama, T.: IH 161
Obara, K.: I 11, 89, 217, 563
Obenschain, S.: Ill 47
Obert, W.: I 31, 449
Obiki, T.: II 355
Ocaiia, J.L.: III 153
Oda, T.: II 355
Odajima, K.: I 151, 545
Ogawa, H.: I 151, 545
Ogawa, I.: I 309, 495
Ogawa, K.: II 453
Ogawa, T.: I 151, 545
Ogawa, Y.: I 151, 309
Ogiwara, N.: I 11, 89, 217, 563
Ogura, K.: 1553
Ohara, M.: n 551
Ohara, Y.: I 11, 89, 217, 563; III 267
Ohasa, K.: I 11, 89, 217, 563
Ohga, T.: I 11, 89, 217, 563
Ohgo, T.: II 355.
Ohhara, H.: I 11, 89, 217, 563
Ohi, S.: II 551, 719
Ohkawa, T.: I 159; II 45, 221

620

AUTHOR INDEX

Ohkubo, K.: 1309, 495
Ohkubo, M.: I 11, 89, 217, 563
Ohnishi, M.: H 719
Oho, K.: I 553
Ohshima, T.: I 11, 89, 217, 563
Ohta, K.: I 11, 89, 217, 563
Ohta, M.: I 11, 89, 217, 563
Ohtaka, M.: I 11, 89, 217, 563
Ohtake, I.: II 355
Ohtsuka, H.: I 151, 545
Ohuchi, K.: I 151, 545
Ohuchi, Y.: I 11, 89, 217, 563
Ohyabu, N.: I 159, 187, 275
Oikawa, A.: I 11, 89, 217, 563
Oka, Y.: I 309
Okabayashi, M.: I 275
Okabe, Y.: U 583
Okada, H.: II 109, 355, 473
Okada, M.: ffl 351
Okada, S.: H 551, 673
Okamoto, M.: II 719
Okamura, S.: II 343
Okano, H.: I 151, 545
Okano, K.: IH 267
Okazaki, K.: H 355
Okuda, I.: HI 167
Okumura, H.: I 11, 89, 217, 563
Okumura, Y.: I 11, 89, 217, 563;
III 267
Okuno, K.: i n 279, 357
Olsson, M.: I 31, 449
Omori, K.: I 11, 89, 217, 563
Omori, S.: I 11, 89, 217, 563
Omori, Y.: I 11, 89, 217, 563
Onishchenko, V.V.: I 645
Ono, M.: 1477, 503
Onodera, S.: II 433
Ortolani, S.: II 441
Osborne, T.: I 159
Oshikiri, M.: m 357
Oshiyama, H.: II 413
Otero, R.: Ill 153
Overskei, D.: I 159

Owadano, Y.: Ill 167
Owens, D.K.: I 51, 75, 171, 207, 421
Oyama, H.: II 355
Ozaki, A.: H 529
Ozaki, S.: I 553
Ozeki, T.: I 11, 89, 217, 563; II 27
Paccagnella, R.: II 441
Pacco, M.G.: I 31, 449
Paduch, M.: II 593
Paillere, J.: I 31, 449
Pakula, R.: HI 25
Papastergiou, S.: I 31, 449
Parfeniuk, D.: LQ 187
Parham, B.J.: I 575
Park, H.: I 51, 75, 171, 207, 421
Park, W.: 1421
Parker, J.: I 139
Parker, R.: I 139, 509; HI 259
Parks, P.B.: II 343
Parlange, F.: 1525
Partridge, J.: 131, 449
Pashnev, V.K.: H 497
Pasini, D.: 131, 449
Patel, A.: II 399
Patou, C : m 3
Patterson, D.M.: I 187
Paul, S.: 1275
Paulson, C : HI 259
Pavlichenko, O.S.: II 497
Pavlov, I.P.: 1607
Pavlov, S.S.: II 205
Pawley, C.J.: IE 137
Peacock, N.J.: I 31, 197, 227, 449
Pearlstein, L.D.: II 283
Pearson, D.I.C.: I 249, 485
Pease, R.S.: II 399
Pedorenko, A.V.: II 89
Peebles, W.A.: I 187
Peerless, S.J.: HI 367
Pellat, R.: H 17; III 79
Peng, H.S.: m 123
Peng, Y.-K.M.: IH 259
Peranich, L.: II 263

AUTHOR INDEXPerepelkin, N.F.: II 497
Perkins, D.E.: H 231
Perkins, F.W.: I 337
Perlado, J.M.: III 153
Persing, H.: II 263
Pert, G.J.: III 55
Pescatore, M.: I 31, 449
Pesme, D.: HI 79
Petersen, P.: I 159
Peterson, R.L.: HI 333
Petrasso, R.: I 139
Petravic, M.: HI 259
Petrie, T.: I 159
Petrov, A.E.: II 481
Pfirsch, D.: m 447
Phillips, C.K.: I 653; ffl 259
Phillips, J.: I 159
Phillips, J.A.: II 423
Phillips, M.: 1299; II 37, 283;
ffl 259
Phillips, P.E.: I 187
Pickles, W.L.: H 231
Pickrell, M.M.: H 423
Pietrzyk, Z.A.: I 125, 461; H 673
Pigarov, A.Yu.: I 237; III 387
Pimenov, A.B.: I 111
Pincosy, P.A.; II 231
Pinsker, R.: I 503
Pisarczyk, T.: II 593
Pistunovich, V.I.: I 237
Pitcher, C.S.: I 227
Plancoulaine, J.: 131, 449
Platts, D.A.: H 519
Plyusnin, V.V.: II 497
P6cs, L.: E 251
Podesta, D.L.: IE 333
Podyminogin, A.A.: Ill 467
Poffe", J.-P.: I 31, 449
Pogutse, O.P.: H 127; HI 435
Pohl, F.: HI 447
Polakova, G.N.: II 497
Politzer, P.A.: R 45
Polman, R.W.: I 599, 659

621

Pomphrey, N.: I 51, 75, 207, 421;
II 37; III 259
Ponomarenko, N.P.: II 497
Popov, A.M.: II 89
Popov, I.A.: 1111
Popov, S.N.: II 481
Popovichev, S.V.: I 111, 237
Porcelli, F.: m 397
Porkolab, M.: I 509
Porter, G.D.: n 231, 283
Porter, J.: I 187
Poschenrieder, W.: I 125, 357, 461
Pospieszczyk, A.: I 485
Post, D.E.: ffl 213, 259
Post, R.S.: II 251
Postupaev, V.V.: II 323
Poulsen, P.: H 231
Powell, E.: I 275
Power, A.: Ill 367
Pozharov, V.A.: II 177
Poznyak, V.I.: I 111
Prater, R.: I 159, 587
Prentice, R.: I 31, 99, 449
Pribyl, P.: I 139
Price, H.D.: II 343
Prichard, B.: I 51, 75, 207, 421
Prinja, A.K.: 1249
Pritchard, J.: I 227
Puhn, F.: HI 259
Puiatti, M.E.: II 441
Puri, S.: 1461
Qu, W.X.: II 157
Radtke, J.: II 263
Raimondi, T.: I 31, 449
Ramen, R.: II 673
Ramette, J.: 131, 197, 449
Ramos, J.J.: H 135
Ramsey, A.T.: I 51, 75, 171, 207, 421
Ranyuk, T.Yu.: II 497
Rapp, H.: I 125, 357, 461
Rawls, J.: I 159
Rax, J.M.: I 525
Raymond, C : I 31, 449

622
Razdobarin, G.T.: I 607
Razumova, K.A.: I 111
Rebut, P.H.: 131, 449; II 187
Redi, M.H.: 151, 75, 171, 207,
337, 421
Reed, J.R.: III 333
Reis, E.: I 159
Reisse, J.-M.: Ill 3
Reiter, D.: 1485
Reiter, R.: I 249
Rej, D.J.: II 673
Remez, L.N.: II 497
Removille, J.: I 31, 449
Remsen, D.: I 159
Renner, H.: II 369
Rensink, M.E.: II 283
Ress, D.B.: II 231
Reusch, M.: I 275
Rewoldt, G.: 1337
Rhodes, T.L.: I 187
Ribe, F.L.: II 739
Rice, J.: I 139, 509; II 355
Richards, B.: I 187
Richardson, M.C.: III 47
Riediker, W.: I 31, 449
Riedler, H.: I 125, 357, 461
Riedy, P.: I 159
Riley, D.: Ill 55
Ringler, H.: II 603
Ripin, B.: Ill 47
Ritz, C.P.: I 187
Riviere, A.C.: I 575
Roberts, D.: II 263
Roberts, R.: 131, 449
Robertson, E.W.: III 55
Robinson, D.C.: I 31, 291, 433,
449, 575
Robson, A.E.: III 367
Rock, P.: I 159
Rogister, A.: I 249, 485; II 167
Rognlien, T.D.: II 283
Rogozin, A.I.: III 467
Rohr, H.: I 125, 357, 461

AUTHOR INDEX

Rolfe, A.: I 31, 449
Romain, J.P.: III 79
Roquemore, A.L.: I 51, 75, 207, 421
Rose, S.J.: III 55
Rosenbluth, M.N.: II 135; HI 301
Roshchin, V.I.: H 481
Roslyakov, G.V.: E 305; III 467
Rosocha, L.A.: III 69
Ross, D.W.: I 187; II 135
Ross, G.: 1249, 485
Ross, R.T.: I 31, 99, 449
Roth, J.: I 125, 357, 461
Rowan, W.L.: I 187
Rozhdestvenskij, V.V.: I 607
Rubtsov, K.S.: H 497
Ruchko, L.F.: I 645
Rumyantsev, E.N.: I 607
Rupert, P.R.: III 333
Rupprecht, G.: I 31, 449
Rusbridge, M.G.: II 399
Rusbuldt, D.: I 249, 485
Rushton, R.: 131, 449
Rutherford, P.H.: I 51, 75, 207, 421
Ryter, F.: I 125, 357, 461, 525, 617
Ryutov, D.D.: II 305, 323; III 467
Sabbagh, S.: 1299
Sachsenmaier, P.: HI 25
Sack, H.C.: I 31, 449
Sadler, G.: I 31, 99, 449; III 457
Saffert, J.: 131, 449
Saigusa, M.: I 11, 89, 217, 563
Saito, K.: II 413
Saito, M.T.: II 433, 655
Saito, R.: HI 267
Saito, T.: II 243
Sakabe, S.: Ill 25
Sakai, K.: II 109
Sakamoto, K.: I 11, 89, 217, 563
Sakamoto, M.: I 495
Sakamoto, Y.: II 355
Sakasai, A.: I 11, 89, 217, 563
Sakata, S.: I 11, 89, 217, 563
Sakharov, N.V.: I 607

AUTHOR INDEX

Sakurai, K.: I 309, 495
Saleres, A.: in 113
Salmon, N.: I 31, 449
Samm, U.: I 249, 485
Samushia, V.D.: I 645
Sand, F.: I 31, 449
Sandmann, W.: I 125
Sanin, A.L.: II 323
Sano, F.: 11355,473
Santagiustina, A.: I 31, 449
Santolaya, J.M.: III 153
Sanuki, H.: II 489
Sanz, J.: Ill 153
Saoutic, B.: I 197
Sapozhnikov, A.V.: II 481
Sardei, F.: II 603
Sarkissian, A.H.: I 413
Sarksyan, K.A.: II 481
Sasaki, T.: m 33
Sasao, M.: I 309, 495
Sass, B.: E 441
Sato, K.: I 309, 477, 495; II 413, 719
Sato, K.N.: I 309, 495
Sato, M.: H 355, 473
Sato, T.: II 27, 65, 343, 433, 529, 693
Sato, Y.: II 563
Satomi, N.: II 529
Satou, M.: I 11, 89, 217, 563
Saunders, R.: I 31, 449
Sauneuf, R.: Ill 113
Sauthoff, N.R.: I 51, 75, 207, 421
Savrukhin, P.V.: 1111
Sawada, K.: II 243
Sawahata, M.: I 11, 89, 217, 563
Sawai, K.: Ill 407
Sbitnikova, I.S.: H 481
Scardovelli, R.A.: II 701
Scarin, P.: II 441

Schaffer, M.J.: H 655
Schaubel, K.: I 159
Schep, T.J.: I 599, 659
Schilling, G.: I 51, 75, 207, 421
Schirmann, D.: HI 3

623

Schissel, D.: I 159
Schivell, J.: I 51, 75, 171, 207, 421
Schliiter, A.: II 383
Schliiter, J.: 1249, 263, 485
Schmid, M.: I 31, 449
Schmidt, G.L.: I 51, 75, 171, 207, 421
Schmidt, J.: IH 259
Schmitter, K.H.: 1461
Schnack, D.D.: II 701
Schneider, F.: I 125, 357, 461
Schneider, K.P.: II 399
Schoch, P.M.: I 187
Schoenberg, K.F.: II 423
Schofield, A.E.: II 423
Schopf, K.: in 373
Schrader, W.J.: I 599
Schiiller, F.C.: I 31, 65, 99, 433, 449;
III 457
Schuresko, D.: I 171
Schwarzmeier, J.L.: II 673
Schweer, B.: I 249, 485
Schwenn, U.: II 383
Scott, B.D.: II 663
Scott, S.D.: 151, 75, 207, 421
Scoville, J.: I 159
Seggern, J. von: I 31, 449
Seimiya, M.: I 11, 89, 217, 563
Seka, W.: in 47
Seki, M.: I 11, 89, 217, 563
Seki, S.: I 11, 89, 217, 563; II 27
Sekiguchi, T.: HI 497
Selin, K.: I 31, 449
Semenov, I.B.: I 533
Sen, A.: II 101
Sen, A.K.: I 299
Sengoku, S.: I 151, 545
Senju, T.: U 355
Senko, T.: I 171
Seraydarian, R.: I 159
Serrano, J.F.: III 153
Sesnic, S.: I 51, 75, 171, 207, 275, 421
Sethian, J.D.: U 573
Setzensack, C : I 125, 357, 461

624

AUTHOR INDEX

Severn, G.: II 263
Sevillano, E.: II 251
Seyler, C.E.: II 673
Sgro, A.G.: II 639
Shaing, K.C.: II 157, 391
Shakhovets, K.G.: 1607
Shannon, T.: Ill 259
Shapoval, A.N.: II 497
Sharp, L.E.: U 511
Shats, M.G.: II 481
Shaw, R.: I 31, 449
Shcheglov, M.A.: II 323
Shcherbinin, O.N.: I 607
Sheffield, J.: HI 259
Shen, H.Z.: III 123
Shen, J.W.: II 135
Shen, S.S.: III 279
Shi, X.H.: II 511
Shiba, T.: Ill 343
Shibanuma, K.: I 11, 89, 217, 563
Shibata, T.: I 151, 545; IH 267
Shibuya, T.: I 151, 545
Shigaev, A.F.: I 533
Shiho, M.: II 231
Shiina, S.: II 413
Shiina, T.: I 151, 545
Shikama, T.: Ill 351
Shimada, M.: I 159, 207
Shimada, R.: I 11, 89, 217, 563
Shimada, T.: II 453
Shimamoto, S.: Ill 357
Shimamura, S.: II 551
Shimer, D.W.: III 333
Shimizu, K.: I 11, 89, 217, 563
Shimizu, M.: I 11, 89; 217, 563
Shimoda, K.: II 583
Shimomura, Y.: I 11, 89, 217, 563
Shimotori, Y.: Ill 177
Shindo, H.: II 413
Shinohara, S.: II 433
Shinozaki, S.: I 11, 89, 217, 563
Shiraga, H.: HI 33
Shirai, H.: I 11, 89, 217, 563

Shirakata, H.: I 11, 89, 217, 563
Shishkin, B.V.: I 111
Shitomi, M.: I 11, 89, 217, 563
Shoji, T.: I 151, 159, 545; II 489
Short, R.W.: III 47
Shpigel', I.S.: II 481
Shtan', A.F.: II 497
Shumaker, D.E.: II 673
Shvets, O.M.: H 497
Sibley, A.: I 31, 449
Sidorov, V.P.: I 645
Siebert, K.D.: II 157
Siemens, P.D.: III 333
Siemon, R.E.: II 673
Sigel, R.: Ill 25
Sigmar, D.: I 139; II 135; III 259
Sigournay, D.: I 31, 449
Sillen, R.: 131, 449
Siller, G.: I 125, 357, 461
Silver, E.H.: II 231
Simm, C.W.: I 617
Simon, A.: Ill 47
Simonen, T.C.: II 231
Simonet, F.: I 31, 449
Simonini, R.: I 31, 449
Sing, D.: II 263
Sinitskij, S.L.: II 323
Sinman, A.: II 731
Sinman, S.: II 731
Sinnis, J.C.: I 51, 75, 207, 421
Sitar, E.V.: 1111
Sitt, B.: Ill 3
Skarsgard, H.M.: I 413
Skibenko, A.I.: II 497
Skupsky, S.: HI 47
Skvortsova, N.N.: II 481
Slavnij, A.S.: II 497
Sleaford, B.: I 159
Sledziriski, S.: H 593
Slough, J.T.: II 541
Smatlak, D.: II 251
Smeulders, P.: I 31, 99, 125, 357,
433, 449, 461

AUTHOR INDEX

Smirnova, A.D.: II 481
Smith, D.K.: II 251
Smith, G.R.: II 283
Smith, J., Jr.: I 159
Smith, L.: Ill 129
Smith, P.: I 159
Smith, T.: I 159
Snider, R.T.: I 159, 373
Snipes, J.A.: I 31, 291, 433, 449
Sokolov, Yu.A.: I 533
Soldner, F.X.: I 125, 357, 461
Solntsev, A.M.: 1111
Solodovchenko, S.I.: II 497
Soltwisch, H.: I 249, 263, 485
Sonnenberg, K.: I 31, 449
Sonnerup, L.: I 31, 449
Sorensen, H.: II 441
Soures, J.M.: III 47
Speth, E.: I 125, 357, 461
Spong, D.A.: II 135, 157
St. John, H.: I 159, 373
Stabler, A.: I 99
Stacey, W.M., Jr.: Ill 199
Stallard, B.W.: II 231
Stambaugh, R.D.: I 159, 373
Stamp, M.F.: I 31, 99, 197, 449;
HI 457
Stamper, J.: Ill 47
Start, D.F.H.: I 575
Stauffer, F.J.: I 51, 75, 171, 207,
387, 421
Stav, R.: I 159
Steed, C.A.: 131,449
Steinhauer, L.C.: II 541
Steinmetz, K.: I 125, 461
Stepanenko, M.M.: I 111
Stepanov, K.N.: II 205
Sterbentz, W.H.: IH 333
Steuer, K.-H.: I 125, 357, 461
Stevens, J.: 1477, 503
Stewart, J.J.: II 283
Stockdale, R.: I 159
Stodiek, W.: 1263

625

Stone, R.R.: III 333
Storey, D.P.: E 399
Stork, D.: 131, 99, 449
Storm, E.: Ill 15
Stotler, D.P.: II 135
Stott, P.E.: I 31, 99, 433, 449
Strachan, J.: I 207
Strait, E.J.: I 159, 373
Stratton, B.C.: I 51, 75, 171, 207, 421
Strauss, H.R.: II 701, 711
Street, R.: I 159
Strelkov, V.S.: I 111
Strickler, D.: Ill 259
Stringer, T.E.: 131, 325, 449
Stupakov, G.V.: II 305
Subbotin, A.A.: Ill 421
Suckewer, S.: I 477
Sudo, S.: n 355, 473
Sugai, H.: II 355
Sugama, H.: II 625
Suganuma, K.: I 11, 89, 217, 563
Sugie, T.: I 11, 89, 217, 563
Sugihara, M.: Ill 267
Sugihara, R.: II 109
Sugiyama, L.: III 397
Sugiyama, T.: I 11, 89, 217, 563
Sukachev, A.V.: I 645
Sukhodol'skij, V.N.: II 481
Sullivan, J.: II 251
Sullivan, J.S.: in 333
Summers, D.R.: I 31, 99, 449
Summers, H.P.: I 31, 197, 449
Sunaoshi, H.: I 11, 89, 217, 563
Sundaram, A.K.: II 101
Sushko, I.S.: II 497
Suzuki, K.: 111, 89, 151, 217, 545, 563
Suzuki, M.: I 11, 89, 217, 563
Suzuki, N.: 111, 89, 151, 217, 545, 563
Suzuki, S.: I 11, 89, 217, 563
Suzuki, Y.: I 11, 89, 217, 563
Swain, D.: Ill 259
Swanson, D.G.: I 653; III 259
Sykes, A.: I 227

626

AUTHOR INDEX

Szydlowski, A.: II 593
Tabatabaei, S.D.: III 55
Tachikawa, K.: Ill 357
Tada, E.: Ill 357
Tagger, M.: II 17
Tagle, A.J.: I 31, 65, 449
Taira, T.: Ill 161
Tait, G.D.: I 51, 75, 171, 207
Tajima, T.: II 719; HI 301
Takabe, H.: IE 33, 103
Takada, S.: I 151, 545
Takagi, M.: IE 33
Takahashi, H.: I 11, 89, 217,
275, 563
Takahashi, M.: I 11, 89, 217, 563
Takahashi, R.: II 355
Takahashi, S.: I 11, 89, 217, 563
Takahashi, T.: I 11, 89, 217, 563;
II 551
Takahashi, Y.: HI 357
Takaishi, T.: II 529
Takasaki, M.: I 11, 89, 217, 563
Takase, Y.: I 139, 509
Takasugi, K.: n 489
Takatsu, M.: I 11, 89, 217, 563
Takayama, K.: II 343
Takeda, S.: H 563
Takeda, T.: II 27
Takeiri, Y.: n 355, 473
Takeno, H.: II 273
Takeshita, A.: I 11, 89, 217, 563
Takeuchi, H.: I 11, 89, 217, 563
Takeuchi, S.: II 109
Takeuchi, T.: II 489
Tallents, G.: I 31, 197, 449
Tamai, H.: I 151, 545
Tamano, T.: II 655
Tamaru, T.: II 413
Tamura, S.: I 11, 89, 217, 563
Tanahashi, S.: 1309, 495
Tanaka, H.: I 553
Tanaka, K.: HI 25, 33
Tanaka, M.: II 65

Tanaka, S.: I 11, 89, 217, 553, 563;
III 267
Tang, W.M.: I 337
Tanga, A.: 131, 65, 99,449;
III 457
Tani, K.: I 11, 89, 217, 563;
III 267
Tani, T.: I 151, 545
Taniguchi, Y.: I 309, 495
Tanimoto, M.: in 167
Tarakanov, A.V.: I 111
Taran, V.S.: H 497
Tarasyan, K.N.: 1111
Taroni, A.: I 31, 65, 325, 449
Taylor, G.: I 51, 75, 171, 207, 421
Taylor, P.L.: II 655
Taylor, T.S.: I 159, 373
Tazima, T.: Ill 343
Teplov, P.P.: I 607
Terakado, M.: I 11, 89, 217, 563
Terakado, T.: I 11, 89, 217, 563
Ternovskaya, M.F.: II 481
Terrington, A.: I 31, 449
Terry, J.: I 139, 509; II 355
Terry, P.W.: I 187
Terumichi, Y.: I 553
Tesini, A.: I 31, 449
Tetsuka, T.: I 495
Texter, S.: I 509
Thomas, G.J.: I 485
Thomas, P.R.: I 31, 65, 99, 249, 449;
III 457
Thome, R.: HI 259
Thompson, E.: I 31, 99, 449
Thomsen, K.: I 31, 99, 449
Throop, A.L.: HI 333
Tibone, F.: I 31, 65, 325, 449
Tiefenback, M.G.: HI 129
Timberlake, J.R.: I 51, 75, 171, 207,
421, 477
Titishov, K.B.: 1111
Tivey, R.: I 31, 449
Tobita, K.: I 11, 89, 217, 563

AUTHOR INDEX

Todd, A.M.M.: I 299; II 37; III 259
Todd, T.N.: I 31, 449, 575
Todoroki, J.: II 625
Todoroki, T.: II 135
Toi, K.: 1309, 495
Tokuda, S.: II 27
Tokutake, T.: I 11, 89, 217, 563
Tolok, V.T.: II 497
Tomie, T.: m 167
Tomita, T.: II 687
Tone, T.: Ill 267
Toner, W.T.: III 55
Tonetti, G.: 199, 433
Tooker, J.: I 159
Torti, R.P.: n 251
Toryak, T.O.: II 497
Totsuka, T.: I 11, 89, 217, 563
Towner, H.H.: I 51, 75, 171, 207, 421
Toyama, H.: H 433
Toyoshima, N.: I 11, 89, 217, 563
Trevalion, P.: I 31, 449
Trubnikov, B.A.: HI 421
Trukhin, V.M.: 1111
Tsai, S T . : II 135
Tsakiris, G.D.: III 25
Tschudin, M.: I 31, 449
Tsuboi, F.: II 489
Tsubouchi, D.: II 243
Tsuda, H.: I 11, 89, 217, 563
Tsugita, T.: I 11, 89, 217, 563
Tsui, H.Y.W.: II 399
Tsuji, H.: Ill 357
Tsuji, S.: I 11, 89, 217, 563
Tsukahara, Y.: I 11, 89, 217, 563
Tsunematsu, T.: II 27
Tsuneoka, M.: I 11, 89, 217, 563
Tsuzuki, T.: II 687
Tubbing, B.J.: I 31, 99, 433, 449
Tuczek, H.: II 673
Tuda, T.: II 27
Tupper, M.: I 159
Turner, L.: II 639
Turner, M.F.: I 227

627

Turner, W.C.: II 231
Tuszewski, M.M.: II 673
Twynam, P.: 131, 449
Uchino, K.: II 355
Uckan, N.: Ill 259
Ueda, Y.: II 433
Uede, T.: Ill 267
Uehara, K.: I 11, 89, 217, 563
Uesugi, Y.: I 151
Ulbricht, A.R.: III 279
Ulrickson, M.: I 51, 75, 171, 207, 421
Umehara, M.: I 11, 89, 217, 563
Uo, K.: II 355, 473
Uramoto, Y.: I 11, 89, 217, 563
Urano, S.: II 355
Usami, H.: I 11, 89, 217, 563
Ushigusa, K.: I 11, 89, 217, 563
Usselmann, E.: I 31, 449
Usui, K.: I 11, 89, 217, 563
Utumi, M.: II 433
Uyama, T.: II 529
Valisa, M.: II 441
Vallet, J.C.: I 525
Van Dam, J.: 1299; II 135
VanderBeken, H.: I 31, 449
Van Heesch, E.J.M.: 1413
Van Houtte, D.: I 525
Van Mark, E.: 1461
Van Montfoort, J.E.: I 31, 449
Van Nieuwenhove, R.: I 249, 485
Van Oost, G.: I 249, 485
Van Rij, W.I.: H 391
Van Wassenhove, G.: I 249, 485
Vance, C.F.: II 511
Vandenplas, P.E.: I 249, 485
VanDevender, J.P.: IE 93
VanSant, J.H.: III 333
Vasil'ev, M.P.: II 497
Vasin, N.L.: I 111, 533
Velarde, G.: HI 153
Velarde, P.M.: Ill 153
Verdon, C : IH 47
Vershkov, V.: I 51, 75, 207, 237, 477

628

AUTHOR INDEX

Vertiporokh, A.N.I.: I 111
Vien, T.: I 461
Virmont, J.: Ill 79
Vlases, G.: I 125; II 673
Vlasov, S.N.: 1111
Vojtenko, D.A.: I 645
Vojtsenya, V.S.: II 497
Volkov, E.D.: II 497
Volkov, V.V.: I 111, 533
Volkov, Ya.F.: II 497
Vollmer, O.: I 125, 357, 461
Von Goeler, S.: I 51, 75, 207, 421, 503
Voronov, G.S.: II 481
Voropaev, S.G.: II 323
Vyacheslavov, L.N.: II 323
W VII-A Team: II 369, 603
Wada, H.: Ill 357
Wade, T.: 131, 449
Waelbroeck, F.: 1249, 485
Wagner, F.: I 51, 75, 125, 207,
357, 421, 461
Waidmann, G.: I 249, 485
Wakabayashi, K.: II 583
Wakatani, M.: II 625
Walker, C : 131, 449
Walker, S.P.: III 367
Wallander, B.A.: I 31, 449
Walravens, M.: 1 3 1 , 449
Walter, K.: I 31, 449
Waltz, R.E.: I 345
Wang, Enyao: II 263
Wang, S.J.: IE 123
Wang, X.-H.: I 299
Wark, J.: Ill 47
Warwick, A.I.: III 129
Watanabe, K.: I 11, 89, 217, 563;
II 529Watari, T.: I 309, 495
Watkins, M.L.: I 31, 99, 449;
III 457
Watson, M.: I 31, 449
Watt, R.G.: II 423
Watteau, J.-P.: Ill 3

Watterson, R.: I 139
Webber, J.P.: 1227
Webberley, D.: I 31, 449
Weber, P.G.: H 423
Wedler, H.: I 461
Weiland, J.: II 639
Weisberg, K.: II 441
Weisen, H.: I 617
Weitzner, H.: II 463
Weldon, D.M.: II 423
Weller, A.: I 31, 197, 433, 449;
II 603; III 457
Welter, H.: II 11
Wereszczyriski, Z.: II 593
Werley, K.A.: II 423, 673, 701
Wesner, F.: 1461
Wesson, J.A.: I 31, 433, 449; II 3
Westerhof, E.: I 599, 659
Weynants, R.R.: I 249, 485
White, R.B.: H 119; in 259
Whitfield, S.: II 399
Wieland, R.: I 51, 75, 171, 207, 421
Wienhold, P.: I 249, 485
Wight, J.: II 673
Wilcock, P.D.: II 399
Wiley, J.C.: I 187
Wilgen, J.B.: I 51, 75, 171, 207, 421
Wilks, J.: I 31, 449
Willi, O.: ffl 55
Williams, M.: I 51, 75, 171, 207, 421
Wilson, D.C.: UI 69
Wilson, J.R.: I 477, 503
Wingate, C.A.: IU 69
Winkel, T.: I 31, 449
Winter, J.: 1249, 485
Witkowski, S.: HI 25
Wobig, H.: II 369
Wolf, G.H.: I 249, 485
Wolfe, S.: I 139, 509; ffl 259
Wolfe, S.W.: 1413
Wong, K.L.: I 51, 75, 207, 421
Wong, S.K.: I 159, 345
Wood, R.D.: II 231

AUTHOR INDEX

Woodward, C : I 31, 449
Wootton, A.J.: I 187
Wouters, A.: 1477
Wright, B.L.: II 519
Wurden, G.A.: II 423
Wykes, M.: I 31, 449
Xie, Jikang: I 317
Yaakobi, B.: Ill 47
Yabe, T.: Ill 25, 33, 343
Yagi, Y.: II 453
Yaguchi, K.: H 355
Yagyu, J.: I 11, 89, 217, 563
Yahagi, E.: II 563
Yamada, H.: H 433
Yamada, K.: I 11, 89, 217, 563
Yamada, S.: II 413
Yamada, Y.: II 583
Yamagishi, K.: II 433
Yamaguchi, N.: II 243
Yamaguchi, S.: I 159
Yamami, Y.: 1545
Yamamoto, M.: I 11, 89, 217, 563
Yamamoto, S.: Ill 267
Yamamoto, T.: I 151, 545; II 583
Yamamoto, Y.: II 583
Yamanaka, C : III 25, 33, 103, 407
Yamanaka, M.: II 583; III 25, 33
Yamanaka, T.: Ill 33
Yamashina, T.: II 355
Yamashita, O.: I 11, 89, 217, 563
Yamashita, Y.: I 11, 89, 217, 563
Yamauchi, R.: 1249
Yamauchi, T.: I 151, 545
Yamauchi, Y.: Ill 351
Yamazaki, K.: I 309, 495; II 27
Yang, J.G.: in 123
Yano, K.: I 11, 89, 217, 563
Yano, M.: Ill 167
Yao, X.Z.: II 251
Yaoita, A.: Ill 167
Yaramyshev, G.S.: I 111
Yasaka, Y.: I 553; II 273
Yasukawa, T.: I 11, 89, 217, 563

629

Yatsu, K.: II 243
Yatsui, K.: Ill 177
Yokokura, K.: I 11, 89, 217, 563
Yokomizo, H.: I 11, 89, 217, 563
Yokoo, M.: II 489
Yokoyama, K.: I 151, 545; II 413
Yokoyama, M.: II 583
Yoneda, H.: Ill 161
Yonekawa, I.: I 11, 89, 217, 563
Yoshida, H.: I 11, 89, 217, 563
Yoshida, K.: ffl 33
Yoshida, K.: HI 357
Yoshida, Z.: E 433
Yoshikawa, K.: II 719
Yoshikawa, M.: I 11, 89, 217, 563
Yoshikawa, S.: I 51, 75, 207, 421
Yoshinari, Y.: I 11, 89, 217, 563
Yoshino, R.: I 11, 75, 89, 217, 563
Young, D.: I 31, 449
Young, K.M.: I 51, 75, 207, 421; III 259
Yu, T.L.: II 231
Yu, Xuehua: I 187
Yuyama, T.: II 489
Zago, S.: II 441
Zaikind, V.M.: II 497
Zakharov, L.E.: II 75
Zannelli, L.: I 31, 449
Zarnstorff, M.C.: I 51, 75, 171, 207, 421
Zasche, D.: I 125, 357, 461
Zaveryaev, V.S.: I 51, 75, 111, 207
Zbasnik, J.P.: III 333
Zehrfeld, H.P.: I 357
Zelenin, G.V.: II 497
Zheng, Shaobai: I 187
Zheng, Z.J.: III 123
Zhou, D.F.: II 511
Zichy, Z.A.: HI 279
Zille, R.: II 383
Zilli, E.: II 441
Zouhar, M.: 1461
Zushi, H.: II 355, 473
Zwart, J.W.: I 31, 449
Zweben, S.J.: I 51, 75, 171, 207, 421

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