Research Using Small Tokamaks

IAEA-TECDOC-969

Research using small tokamaks

Proceedings of a Technical Committee meeting held Ahmedabad,in India, December6-7 1995

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RESEARCH USING SMALL TOKAMAKS IAEA, VIENNA, 1997 IAEA-TECDOC-969 ISSN 1011-4289 ©IAEA, 1997 Printed by the IAEA in Austria September 1997 FOREWORD

The technical reports in these proceedings were presented at the IAEA Technical Committee Meeting on Research Using Small Tokamaks, held in Ahmedabad, India, 6-7 December 1995 purpose .Th thif eo s annual meetin provido t s gi exchang e foruea th r mfo f eo informatio varioun no s smal mediud an l m sized plasma experiments t onlno ,y tokamakse Th . potential benefit thesf so e research programme: to e sar

test theories, suc effects ha plasmf o s a rotation; check empirical scalings, such as density limits; develop fusion technology hardware; develop plasma diagnostics, suc tomographys ha ; train scientists, engineers, technicians d studentsan , , particularl n developini y g Member States.

e meetinTh s Institute hosteth gwa y r Plasmdb fo e a Research, Bhat, Gandhinagar, India. The participants are indebted to P.K. Kaw, A. Sen and P. Ranjan for the excellent meeting arrangements meetine Th . attendes gwa abouy db participant0 5 t s fro countries2 m1 . This TECDOC includes the individual research papers (or in a few cases the abstracts) and a summary that was published in Nuclear Fusion 36 1425-1429 (1996). The session chairmen were J. Fujita, R. Amrollahi, J. Stoeckel, X.W. Deng, E. Azizov and P.K. Kaw.

A broad spectrum of papers was presented, both theoretical and experimental. Some interestin resultw ne g s (suc s froa h m START d techniquean ) s (suc s neuraa h l network analysis of tomographic data) were reported. The main problem for most participants was trying to do state-of-the-art research on low budgets.

e centreTh f excellenco s e thadevelopine ar t severan gi l countries will form strong technical bases for future research projects. International collaboration may be useful to facilitate the exchange of ideas, joint funding of large experiments, and increased public awarenes importance th f so fusiof eo n research. EDITORIAL NOTE

In preparing this publication press,for IAEAthe staffof have pages madethe up from the original manuscripts as submitted by the authors. The views expressed do not necessarily reflect IAEA, thosethe governments the of nominatingthe of Member nominatingStatesthe or organizations. Throughout the text names of Member States are retained as they were when the text was compiled. The use of particular designations of countries or territories does not imply any judgement by publisher,the legalthe IAEA,to the status as suchof countries territories,or theirof 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 imply intentionnot any infringeto proprietary rights, should construednor be it as an endorsement or recommendation on the part of the IAEA. The authors responsibleare havingfor obtained necessarythe permission IAEAthe to for reproduce, translate materialuse or from sources already protected copyrights.by CONTENTS

Energy confinement in ohmic H-mode in TUMAN-3M...... 7 M. V. Andrejko, L.G. Askinazi, V.E. Golant, V.A. Kornev, S.V. Lebedev, L.S. Levin, A.S. Tukachinsky Magnetic fluctuations in CASTOR tokamak and studies of their relationship with the electrostatic fluctuations (Abstract)...... 15 V. Dhyani, K. Jakubka, J. Stockel, F. Zdcek Correlation analysi edgf so e fluctuation CASTOe th n so R tokamak (Abstract)...... 7 1 J. Stockel, Petrzilka,J. Svoboda,V. Horn,M. Kryska,L. EndlerM. Observations on the peculiarities of sawtooth relaxation in SINP tokamak (Abstract)...... 9 1 A.K. Hui, R.K. Paul, Banik,D. Mukhopadhyay,S. RanjanP. Study on plasma biasing, poloidal flow and fluctuation suppression in a toroidal plasma...... 1 .2 K.K. Jain Conceptual desig steadf no y state superconducting tokamak SST1...... 9 .2 SST Team Phenomenolog STARf yo T operating regimes ...... 9 .3 C. Ribeiro, Gryaznevich,M. Hugill,J. Jenkins,I. Martin,R. Robinson, C. D. A. Sykes, T.N. Todd, M.J. Walsh Ergodic magnetic limiter experiment at the small TBR tokamak ...... 45 M.S.T. Araujo, A. Vannucci, I.L. Caldas, I.C. Nascimento Present TCA/Bstatue th f so R tokamak...... 3 .5 I.C. Nascimento, R.M.O. Galvdo, A.G. Tuszel, F.T. Degasperi, Ruchko,L. R.P. Silva,da A.N. Fagundes, A.C.P. Mendes, E.K. Sanada, Taran,V. R.M.O. Pauletti, W.P. de Sd, A. Vannucci, J.H. Vuolo, H. Franco, J.I. Elizondo, I.L. Caldas, M.C.R. Andrade, A.C.A. Ferreira, I.H. Tan, V.S.W. Vuolo, M.V.A.P. Heller, M.Y. Kucinsky, E. Ozono, M.S.T. Araujo, E.A. Lerche HL-1M modificatioe th , HL-e th f 1no tokamak...... 1 .6 HL-1M Team Recent development ADITYn so A operation...... 3 .7 P.K. Atrey, V. Balakrishnan, S.B. Bhatt, D. Bora, B.N. Buch, C. Chavda, C.N. Gupta, C.J. Hansalia, K.K. Jain, Jha,R. P.I. John, P.K. Kaw, Kumar,A. Kumar,V. S.K. Mattoo, C.V.S. Rao, H.A. Pathak, H.R. Prabhakara, H.D. Pujara, P. Ranjan, K. Sathyanarayana, Y.C. Saxena, G.C. Sethia, A. Vardarajalu, P. Vasu Improved confinemen currend an t t driv biasey eb d electrod vern ei y low-ga discharges of SINP tokamak...... 79 J. Ghosh, Chattopadhyay,P. Pal,R. A.N.S. lyengar Recent HL-1M results...... 7 .8 HL-1M Team ICRF plasma productio heatind URAGAN-3ne an th n gi M torsatron...... 5 .9 V.V. Plyusnin, V.E. Moiseenko, A.I. Lysoivan, S.V. Kasilov, A.I. Zhukov, N.I. Nazarov, E.D. Volkov, K.N. Stepanov, O.S. Pavlichenko DAMAVAND — An Iranian tokamak with a highly elongated plasma cross-section...... 107 R. Amrollahi Present status of research activities at the National Institute for Fusion Science and its role in international collaboration...... 127 J. Fujita Status of plasma physics research activities in Egypt...... 135 MasoudMM. Slow bank system of SINP tokamak: A short report...... 143 Ray, R. Ranjan,P. Chowdhury,S. BaseS. JUST: Joint Upgraded Spherical Tokamak...... 149 E.A. Azizov, N.Ya. Dvorkin, O.G. Filatov, G.P. Gardymov, V.E. Golant, V.A. Glukhikh, R.R. Khayrutdinov, V.A. Krylov, Leykin,IN. V.E. Lykash, A.B. Mineev, G.E. Notkin, A.R. Polevoy, K.G. Shakhovets, S.V. Tsaun, E.P. Velikhov, N.I. Vinogradov nonlocaA l mode electror fo l n heat transpor tokamaks...... n i t 3 17 . A. Das, P. Kaw, S. Dastgeer Nonlinear saturation of the Rayleigh Taylor instability...... 181 Das,A. Mahajan,S. Kaw,P. Sen,A. Benkadda, S. VergaA. Shear reversal, improved confinemen temperaturn io d an t e gradient mode ...... 9 18 . S. Sen, A. Sen Ponderomotive modification of drift tearing modes...... 197 G. Urquijo, R. Singh, A. Sen Equilibrium and fluctuations in a currentless toroidal plasma...... 203 Singh,R. Mahajan,S. AvinashK. A transport bifurcation model for enhanced confinement with negative shear...... 211 K. Avinash, P.K. Kaw, SinghR. Tomography using neural networks ...... 9 .21 G. Demeter, . ZoletnikS

Research Using Small Tokamaks: Summary publishe Nuclearn di Fusion...... 7 22

List of Participants...... 233 ENERGY CONFINEMENT IN OHMIC H-MODE IN TUMAN-3M

M.V. ANDREJKO, L.G. ASKINAZI, V.E. GOLANT, V.A. KORNEV, S.V. LEBEDEV, L.S. LEVIN, A.S. TUKACHINSKY

A.F. loffe Physico-Technical Institute, Russian Academ Sciencesf yo , St. Petersburg, Russian Federation

Abstract The spontaneous transition from Ohmically heated lirniter discharges inte oth regime with improved confinement termed as "Ohmic H-mode" has been investigate "TUMAN-3"n di typicae Th . l signature f H-modo s tokamakn ei s

with powerful auxiliary heating have been observed: sharp drop of Da radiation with simultaneous increas electroe th n ei n densit stored yan d energy, suppression of the density fluctuations and establishing the steep gradient near the periphery 199n I . vacuuw 4ne m vesse beed ha l n installe TUMAN-n di 3 tokamak. The vessel has the same sizes as old one (Ro=0.55 m, ar=0.24 m). New vessel was designed to reduce mechanical stresses in the walls during BT ramp phas shota f eo . Therefore modified devic e- TUMAN-3 Mabls i o t e respectivelyA M 2 0. d an . DurinT 2 o t g produc p firsu , tp I e d highean T B r experimenta devicn operateru s l ewa Ohmin di c Regime thesn I . e experiments the possibility to achieve Ohmic H-mode was studied. The study of the parametric dependencie e energth f o sy confinementd timan botn i eH O h Ohmic H-mode was performed. In Ohmic H-mode strong dependencies of IE on plasma current and on input power and weak dependence on density were found. Energy confinement tim TUMAN-3/TUMAN-3n ei M Ohmic H-mode reveales ha d good agreement with JET/DIII-D/ASDEX scalin ELM-frer gfo e

H-mode, resulting in very long TE at the high plasma current discharges.

1. Introduction H-mode was discovered in 1982 in the experiments with powerful heatin Neutray gb l Beam Injectio ASDEn i n X [1]. Since that time H-mode was founa numbe n i d f experimento r s with different auxiliary heating schemes. In a majority of the experiments L-H transition takes place in a divertor configuratio somn i t ebu n case e H-modth s s observewa ee th n i d limiter bounded plasmas also. In ordinary Ohmic regime the transition into the improved confinement regime similar to H-mode was found in the TUMAN-3 tokamak in circular limiter configuration [2]. 2. Energy confinement time parametric studie TUMAN-3n si M device TUMAN- smala s 3i l tokamak with circular cross sectio metallid nan c walls and limiters [3]. After modification including replacement of the vacuum vessel devic names ewa d TUMAN-3M vessew s designeNe . wa l reduco dt e mechanical stresses in the walls during By-ramp phase of a shot. Therefore modified device - TUMAN-3 M is able to produce higher BT and Ip up to 2 T and 0.2 MA respectively. Some machine and plasma parameters are listed in Tabl. e1 Energy confinement time parametric dependencies were studied using diamagnetic measurements of a stored energy. TE dependency on density for OH with two different values of plasma current are given on Fig. 1. The data show weak density dependenc t giveea n curren substantiad an t l increas Tn ei E wit h. I Simila r dependencies were tokamae founth n di k Alcator-CMODt A . lowep r densities experimental energy confinement time exceeds Neo-Alcator

Table 1. TUMAN-3/TUMAN-3M parameters. Range Typical OH Ro, m 44-0 4 06 0 53-0 55 ai, m 0 11-024 0 22-0 24 BT,T 0 34-1 2 (2 0) 0 40-0 8 IP, MA 004-0 16(02) 009-0 15 19 5 nlv, 10 rn 04-52 1 0-25 Tco, keV 03-08 05-06 V T,oke , 009-017 12-0 5 01

u.uu* • < * IP=117 kA, q ^=3.2 d kA, qey<=2.7 E 4 IP=145 0.003 • c » * » E 0.002 • ^.-*>

C tt * o *^9V o 0.001 ET o^ OJ c- n onn '' LL) 0.0 0.5 1.0 1.5 2.0 2.5 Average density, 1019m~3 Fig. 1. Energy confinement time as function of density for OH regimes with

Ip-117&145kA. 10.0 "10.15 u,,v 0.10' 5.0 0.05

V 4.0 U**t, au. 10V 2.0 • i

a.u.

SxR a.u.

10 20 30 40 50t,ms Fig. 2. Temporal evolution of some plasma parameters in 115 kA Ohmic H-mode.

o> IT o o 04 - 01 o: 01 0) Q O o 1 Q, li 02 - "OE

0 0044nBTS 00 0 0 1 0 5 0 0 00 1 50 nB, 10A19mA-3T Fig. 3. H-mode operational region. Symbols show input power before transition into Ohmic H-mode. prediction recyclin[4] w o factoa tw Lo .y resulsf b y o r gdiminisma n i t e th f ho atomic processes influence on confinement and corresponding changes in parametric dependencies. Ohmic H-mode appears spontaneousl r coulyo triggeree db somy db e increas deuteriun ei m puffing TUMAN-3raten i s a , M [2]. Typical waveforms of some plasma parameters in the 115 kA Ohmic H-mode are shown in Fig. 2.

Transitio . Durin ms ne decrease 7 ar startg2 p t thiU a s sd d perioan a D d indicating gradual reduction of the particle flux in periphery and widening of the current density profile The Ohmic H-mode is ELM-free and is Table 2. Examples of the scalings describing energy confinement in Ohmic L- and H- modes.

Scaling, IEs m , Expression (1»1019 m'3, m, MA, MW, T, keV) Neo-Alcator 1.9 Merezhkin-Mukhovatov 025 275 05 05 2.3 l.lna R qk°125A1 / Goldston (L-mode) 10.7

ITER89-P (L-mode) 9.9 85 2 3 5 2 5 5 a s a q 48IP° R* a° k° n° V A,° F° F[fs - fq - ] DIII-D/JET (H-mode) 15.6 110p-046lD103R148

in E o before boronizotion v* • after boronizotion £15

c D lp=112 kA i" "c 0 u

ti c u

0.0 1.0 2.0 3.0 4.0 Average density, 1.E19 m-3 Fig Energ. 4 . y confinement timfunctioa s ea densitf regimne o th r yfo e with current 1 12 kA before and after boronization.

characterized by continuous density increase indicating improvement in particle confinement. After transition SXR emission from the plasma center increases due to electron temperature rise. Fig. 3 presents H-mode operational region in the (Pinput, coordinates. Points show plasma parameters just before transition into the Ohmic H-mode. Discharges enter H-mode operational space from the low density margin (shown by vertical lines). It should be mentioned that all points lie substantially higher threshold power [5].

10 • good boronization n poor boronization 0.03 -

002 -

o o nD E? 001 O) UJ

ooo 0 16 0 12 0 8 40 Plasma current, kA Fig Energ. 5 . y confinement timfunctioa s ea plasmf no a currenn i t TUMAN-3 and TUMAN-3M Ohmic H-mode.

1.0

JET ' om-D 0.1

d>, ASDEX f 0.01 TUMAN-3

i o. o.oi 1.0 0.106 P^L l.4S Fig . Experimenta6 . l energy confinement tim Ohmin ei c H-mod functioa s ea n of JET/DIII-D H-mode scaling.

To describ n energa e y confinemen - modeH Ohmicn d f i to s an - L , tokamak operation different expression usualle sar y used Table se , . Typicae2 l Ohmic scaling predicts linear dependencies of energy confinement time on average density and safety factor. Size dependence is cubic. Examples of this kind of scalings are "Neo-Alcator" [6] and "Merezhkin-Mukhovatov" [7]. Energy confinement in the auxiliary heated L-mode plasmas is characterized y ratheb r different dependencies appear E e proportionaT b . o t s o plasmt l a curren inverseld an t y proportiona squaro t l e roo f inpuo t t power. Frequently cited expressions are "Goldston" [8], "ITER89-P" [9]. Parametric

11 dependencies of IE in H-mode plasma are only slightly different from that ones in L-mode. Exampl expressioe th f eo H-modr nfo e plasm "DIII-D/JEs ai - TH mode" scaling [10]. Tabl showe2 s prediction differene th f so t scaling5 15 r sfo kA Ohmic regime in TUMAN-3 tokamak. Therefore the question is how to describe energy confinemen Ohmin i t c H-mode. Eithe somy rb e enhancement factor over Neo-Alcator scaling or using different expressions for instance JET/DIII-D H-mode scaling. The study of the IE dependencies on input power and density was performed in the regime with a plasma current of 112 kA [11].(Fig. 4) After boronization the input Ohmic power drops by a factor of two whereas energy confinement time increases substantially This figure also indicates that the IE dependenc densite th n eo weas yi r negligibleko Neo-Alcatoe Th . r scalins gi also shown for comparison. Before recent experiment s beeha s n started TUMAN-3s Mwa boronized using Carborane (C2B10H12) deposition in He glow [12]. As a result relatively low recycling was observed in the experiments. Ip scan was performe boronizen di vesselw vessed ne t qualit dol coatine d bu , th an l f yo ga vessew s worsne wa l d evessel ol tha n i n, because onle sourcon y f o e Carborane was used instead of two. Spectroscopic data shows decrease of Oxygen concentration by a factor of 2-3. Under this conditions we have found becamthaE T t e lowea facto y f b 1.5ro r . ReductioundeE I e rth poof o n r vacuum conditions means direct influence of plasma purity on confinement. This coul seee d b Fig n no . .5 Comparison of the results from Ohmic H-mode confinement studies on TUMAN-3M wite DIII-D/JEth h T scalin r ELM-fregfo e H-mode plasma shows good agreement [10]. Fig. 6 Shows that majority of the TUMAN-3M point e neah e lineas th r r approximatio r threfo n e tokamaks (JET, DIII-D, ASDEX).

3. Summary Studies of the energy confinement time in OH in TUMAN-3 M showed weak dependence on density contradicting to Neo-Alcator scaling. The transition into H-mode in ohmically heated plasma was found in a simple circular limiter configuration. Studies of the energy confinement time in Ohmic H-mode in TUMAN- showeM 3 d strong dependenc plasmn eo a inpucurrenn o d t poweran t . Confinement in Ohmic H-mode corresponds well to JET/DIII-D H- mode scaling. These results hel supporo pt e conjecturth t e that H-mode physics ha s common natur tokamakn ei s with different geometr heatind yan g method.

12 ACKNOWLEDGEMENTS This work was partly supported by Russian Foundation for Fundamental Studies (Gran 93-02-334N N t 93-02-169092& Internationay b d )an l Science Foundation (Gran R2T300)N t .

REFERENCES [I] Wagner F et al 1982 Phys. Rev. Lett. 49 1408 ] , ProcArbuzo[2 al t e . 17tL vA h Eur. Conf Contrn o . . Fusio Plasmd nan a Heating (Amsterdam (19909 parB 29 14 )tI ) [3] Vorob'ev GM et al 1983 Sov. J. Plasma Phys. 9 65 [4] Hutchinson ffl et al, Plasma Phys. Contr. Fusion 36 B143 (1994) [5] Ryter F et al, Proc. 21st Eur. Conf. on Contr. Fusion and Plasma Phys. (Montpellier) 18B part I 334 (1994) ] Blackwel[6 l 198a t e 2 B lProc Inth 9t . . Conf Plasmn o . a Physicd san Contr. Nuclear Fusion Research, Baltimore vol II (Vienna: IAEA) p 27 [7] Merezhkin VG and Mukhovatov VS 1981 Sov. JETP Lett. 33 446 [8] Goldston RJ 1984 Plasma Physics and Controlled Fusion 26 87 [9] Yushmano 199N vP 0 Nuclear Fusio 1990 n3 9 [10] Schissel DP et al 1991 Nuclear Fusion 31 73. Simonel 199a t e 1 C nPreprinT t General Atomics GA-A20711 [II] MV Andrejko et al, Plasma Phys. Control. Fusion 36 (1994) A165-A170 [12] Askinazi LG et al, Proc. 20th Eur. Conf. on Contr. Fusion and Plasma Phys. (Lisbon) 17C part IV 1517 (1993)

13 XA9745735 MAGNETIC FLUCTUATIONS IN CASTOR TOKAMAK AND STUDIE THEIF SO R RELATIONSHIP WITE HTH ELECTROSTATIC FLUCTUATIONS (Abstract)

V. DHYANI1, K. JAKUBKA, J. STOCKEL, F. ZACEK Institute of Plasma Physics, Czech Academy of Sciences, Prague, Czech Republic

magnetie Th c fluctuation monitoree sar radian di poloidad lan l directions deep (upt0.4= a )o r/ int tokamae oth theid kan r relationship f anyi ,s ha , been traced out in the tokamak plasma. For carrying out these experiments different sets of movable probes, consisting of magnetic and Langmuir probes movable from r/a= r/a=0.4o t l usede ar , . eighf o t tA se magneti c probes probee , witth l hal s registerin radiae gth l component of magnetic field and radially spaced from each other, is constructed to study the radial magnetic field fluctuations. With this set the whole radial profile from r/a= r/a=0.o t l 4 coul achievee db r eacdfo h shot. The radial correlation lengt propagatiod han n velocit radiae th f yo l magnetic field fluctuations (DT) are calculated and radial profile of the propagation velocity (BT) is obtained over a substantial cross-section of the plasma column. The advantage of this set of probes is that the behaviour of magnetic fluctuations could be studied at different radial positions (from r/a = 1 to = 0.4 a )r/ durin e shotgon . Moreove coule w r d observ e fluctuationth e s with different plasma current lood relationshir san kfo p betwee plasme nth a

current (/p) and magnetic fluctuations (DT). The root-inean-square (RMS) values correlatioe th , n value radiad san l wavenumbe frequencd an r y spectra are obtained for magnetic fluctuations at all available radial positions of the probes. Two movable sets, each of two probes, are installed for observing the DT in toroidal and poloidal directions. In one set the probes are toroidally space d eac an ds facinhi g radial magneti e tokamakc th fiel f o de dat Th .a from these toroidally spaced radial probes tell the behaviour of radial magnetic field fluctuation e toroidath n i se radia le th othespace th t lse n rI . magnetic probes are poloidally spaced from each other and both are look- ing at the radial direction of the magnetic field. The toroidal and poloidal wavemimbers and frequency spectra are computed and RMS and correlation values are obtained at different radial positions and compared particularly for two cases, r/a=l and r/a=0.4. For studying the relationship between the magnetic field fluctiations and the electrostatic fluctuations two other sets of probes are used. In one set the two magnetic probes are radially separated from each other, each observing the radial component of magnetic field, and a Langmuir probe is situated parallel to them. In the other set two magnetic probes are poloidally spaced from each other, each registering the poloidal component of the magnetic field, and a Langmuir probe is installed in their axis.

'Cultural exchange programme between India and the Czech Republic.

15 diameten i magnetie m m Th lengtn r4 i whild m chan m probe e4 e ar s the Langmuir probes are 2mm long and 0.5 mm in diameter. For performing these experiments the CASTOR tokamak was operated with the following parameters: major radius R = 0.4 m, minor radius a 0.08= , toroida , plasm5cm IT l= fielaT currenB d - 4-6-8-1 p t I ta A 2k 19 3 electron density ne < 10 m~ and central electron temperature Te ~ 100 - 150eV. The position of plasma column is feedback controlled. The results indicate that the level of magnetic fluctuations in both, radial 5 4 (Dr/BT) and poloidal (BP/BT), directions is in the range of 10~ - 510~ continuousls i d an 4 0. betwee y= a radiae r/ th nd lan position1 = a r/ s increasing as probes are taken towards the core of the CASTOR. The propagation velocity of the radial magnetic field fluctuations is found to be hi the range of 5-14 km/sec and is increasing while coming towards the edge from the core. The direction of propagation of BT is from core towards the edgte. The correlation length of BT is observed to be in the range of 5-6 cm whic comparabln hi e range wit minoe hth r radiudevice th f casso n ei e of CASTOR tokamak. We had previously reported[lj that radial magnetic fluctuations were broadenin synchronisind gan toroidale th n gi poloidad an l spac therd formatios ean edimensionawa o tw f no l structures because ar r eB movin poloidae th n gi l frame also. Her fine T havew d B tha e higth t h correlation value l radiaal t sa l positions between r/a= r/a=0.o t l eve d correlatioe 4an nth n betwee innere nth - mos 0.4 = outermost d a prob)an r/ t e(a t prob equits i (r/) e1 asubstan= - tial (about 0.5) whic obviousls hi explanation ya higr nfo h radial correlation wavenumbee Th lengt . r B hf o (5- r) frequenc6cm y spectra sho increasn wa - ing shift in the k-space corresponding to the increasing propagation velocity s observei t i d dan thaT e frequencieoB fth t e rangth f 50-15n eo i s z 0kH are quite dominant in the fluctuation spectrum. At the edge of the toka- k (r/a=lma spectrue th ) m becomes increasingly turbulent evaluation A . n of wavenumber frequencied san s wil presentee b l detain e meetingdi th n i l . correlatiow lo e Th n betwee magnetie nth electrostatid can fore th m n c(i saturation oio f n curren floatind an t g potential) fluctuations seem indio t s - cate that there might be some marginal connection (not necessarily through the same origin) between the magnetic and electrostatic fluctuations. How- ever, the relationship between the two fluctuations appears to be more subtle d complean x becaus e intrinsith f o e c complexit e observeth f yo d phenom- t advocata whicno en o hd readilr efo y available interpretation basie th sn so of a linear MHD model where magnetic and electrostatic fluctuations are simply correlated. Further efforts are going on through computations of the frequencie d wavenumberan s s involved wit magnetie hth electrostatid can c fluctuations for the proper evaluation of the relationship between them. The latest results will be presented in the meeting.

Reference 1. V. Dhyani et al., Transactions of Fusion Technology, 473, Vol. 27 April, 1995.

16 XA9745736 CORRELATION ANALYSIS OF EDGE FLUCTUATIONS CASTOE ONTH R TOKAMAK (Abstract)

J. STOCKEL . PETRZILKAJ , . SVOBODAV , , M. HORN, L. KRYSKA Institut Plasmf eo a Physics, Prague, Czech Republic M. ENDLER Max-Planck-Institut fur Plasmaphysik, Garching, Germany

Electrostatic fluctuation generalle sar y assume responsible b o dt e th r efo anomalous particle/energy losses in tokamaks. However, their nature is not still well understood here Langmuie us eth e W . r probe technique providing spatiall temporalld yan y resolved informatio frequencnw aboulo < e th / t ( y 0.2 MHz) fluctuation plasme th f o s a densit potentiad yan CASTOe th n lo R tokamak (R = 0.4 m, a = 0.085 m, B = 1 T, I = 12 kA). We use the multiple tip Lang nuir probes (a row of 16 tips) oriented in the poloidal or radial direction . The even tips monitor the floating potential which is related to the local plasma potential tipd , whilod s e measursaturation eth io e eth n current r ;lated to the plasma density. The probe signals are digitized and analyze calculatiny db e spatial-temporagth l correlation function. The r 'suiting form of the correlation function corresponds to blobs ("single events") which appear randoml d decaan y y again. They propagate poloidall.', having a certain poloidal periodicity, but they do not appear pe- riodicallj in time. Typically, the density and potential fluctuations on CAS- poloidae TOth n Ri l directio e characterizenar e correlatioth y db n length , correlatiocm 2 ~ n time 10-15 /zs, poloidal wavelengte th d han 8-1m c 0 poloidal -elocity 0.7 - 1.2 km/s. r < xperimentaOu l arrangement (the adjacent tips operat n differentei modes) allows to derive the crosscorrelation between the floating potential ansaturation dio n current foune W . quita d e strong positiv negativd ean e correlatio 0.5> n( ) between bot. Thi signale distance cm hth sth 5 t 1. sa e+ can be understood as formation of "eddies" in the edge plasma.

The (dge turbulence is also analyzed in the radial direction. In this experiment, an additional limiter (covered by the insulating layer) with a = 60 mm was installed into the tokamak chamber. Comparison with the poloidal direction data shows:

radiae th • l correlation lengt shortes . hi 2 factoa - y 5 b r1. r • neither spatial periodicity nor radial propagation of the blobs is ob- servec • relat levee r densitf o l y fluctuations wit insulatine hth g limite lowes ri r standare thath n n d limiter configuration.

Comparison of our results with those from the ASDEX tokamak (in particula poloidae th i i r l direction) reveal similaa s r characte e edgth ef o r

17 fluctuations despite of the quite different sizes of both the experiments. This indicates ai universal nature of edge turbulence in tokarnaks.

Acknowledgement

The wo:k was performed under GACR grants 202/0711 and 202/1502 and suppor ed by the IAEA contract 6072/R2/RB.

18 OBSERVATION PECULIARITIEE TH N SO F SO XA9745737 SAWTOOTH RELAXATIO SINN I P TOKAMAK (Abstract)

A.K. HUI, R.K. PAUL, D. BANIK, S. MUKHOPADHYAY, P. RANJAN1 Saha Institute of Nuclear Physics, Calcutta, India

SINP (max) T tokama2 t B smal, a s ki cm l5 machin7. a , ecm wit0 3 hR lp 75 kA (max). The sawtooth phenomena is being studied in SINP in connection with investigations of mhd instabilities leading to internal, minor and major disruptions. Main diagnostics used in these studies are external magnetic sofa loop d t x-raan s y imaging system. Presently, ther s onlei y one array in the imaging system though more are being installed. Con- sequently, solid rotatio f plasmno s beeaha n assume e tomographith n di c reconstruction presentee b o st thin di s report. Though many characteristic SINf so P sawtoot sees ha sofn i t x-ray agree with e classicathosth f eo l (Kadomstev) model, ther mane ear y differences. Here we list some of them which are indicative of a mechanism additional to the mhd m=l tearing mode leading to sawtooth relaxation.

• Many reconstructions of the soft x-ray emissivity contours during a swatooth period show a growing cresent-like island structures. How- ever, the widths of the islands never grow enough to engulf the cores befor e crashth e thin I . s respec reconstructione th t s show similarity with those of TFR wherein a small scale magnetic turbulent phenom- beed ha n a invokeden . • SINP doe t havcontroy no s ean currenr fo l t profile. Consequentle yth inversion radius of sawtooth oscillation, r,, changes widely from 0.8 cm (r./ i 10%ac minoe , th wher (r,/s i m rc a radiuse4 a~ 2. o t ) 30%e e time widtislan th th f crasTh eo f .t hda o h doe t increasno s e significantly in magnitude when one goes from the lowest to highest . Thir, s clearly indicate l s tearintham= td gthoug mh mod e th he may play a significant role in causing this relaxation, this is aided by some other phenomena t seemI . d s thamh e t th whe o islane t nth e ddu mode attains sufficient width, the level of some kind of turbulence grows enoug resulo ht rapin i t d transpor energf o t y fro coree md th an , hence the sawtooth crash. • The above observation is substantiated by the study of the sawtooth period with the inversion radius. Though the sawtooth period does increase wit e inversiohth n 0 radiu23 o t s s (fro n averagn ma 0 17 f eo // s) the increase is much less than what one should expect from three

fold increase in rt.

'Present address: Institute for Plasma Research, Bhat, Gandhinagar 382424, India.

19 • The sawtooth oscillation in SINP is stabilized by a strong gas puffing One of the essential features of the preliminary analysis of the data on this stabilization process is that the mhd m=l tearing modes still persists on stabilization at a location close to inversion radius when sawtooth occured Its amplitude saturates at a value not less than the amplitud thif eo s oscillatio sawtoote th t na h crash. This givea s strong indicatio t beeno nt ne i turbulen aideth thad y ha tdb t phenomena postulated, this mode would saturate well before causine gth relaxation The stabilization of sawtooth is being effected by lowering e turbulancoth f epuffins levega n o lg

The possible candidate for the triggering of the sawtooth crash is looked upo magnetis na c turbulence Amphtxid broadbanf eo d magnetic fluctuation is negligibl e edgth en ei regio nr studied d beewherfa ha o ns t i e , however, more recent studies indicate its rapid increase as one goes into the core again Consequently, ther beed eha n man) theoretical activitie magnetin o s c turbulenc speculationd an e role f thith eo n s so phenomen overale th n ai l confinement Als recenn oi t years some experimental works indicate significant magnetic fluctuation during strong tearing mode activity. In absence of any parametric study in these works it is difficult to do any correlation, particular lase termn i hth t f observatioo s n However, experiment e plannear s d for observation of small scale magnetic fluctuation on SINP during sawtooth actu it\ In this report we shall present the experimental data as well as the re- e pnmarsultth f o s > analysis

20 XA9745738 STUDY ON PLASMA BIASING, POLOIDAL FLOW AND FLUCTUATION SUPPRESSIO TOROIDAA NN I L PLASMA

K.K. JAIN Institut Plasmr efo a Research, Bhat, Gandhinagar, India

Abstract

We observe poloidal plasma rotation (Ve) and concomitantly, reduction in fluctuation levepura n i le toroidal plasma with positive electrode bias- radiae ingTh . l structur poloidaf eo l flows show regiosa sheae V f no r where suppressio fluctuatiof no significants ni . Dependenc biae th sf V n eo gvolto - toroidae th agd ean l magnetic fiel presenteds di .

e nee Th understano t d e physicadth l mechanism responsibl- ob r efo served improved H-mode plasma confinemen tokamakn i t s urgeni s - be t cause this appears to be an important step towards success of controlled thermonuclear fusion. In recent years [1-5], various theoretical models have been propose explaio dt H-modee nth . Mos f theo t m predice t thath n i t H-mode, a radial electric field ET is generated (can also be externally imposed) at tokamak edge resulting in poloidal (Ve) and toroidal (V^) plasma rotation theoreticae Th . l calculations show tha sheae th t r (Vg) and/or curvature (Ve poloidan )i l plasma rotatio capabls ni suppressinf eo plasme gth a fluctuations and thus reducing the plasma transport. The models based on linear stability [3], scaling approac turbulencd han predic] e[4 threshola t d for poloidal plasma velocity shea causo rt e fluctuation suppression.

Experimentally, the H-mode is observed with auxiliary heating [6-8] and has also been induced by bias electrode [9-10] in many tokamaks. Prelim-

inary measurement plasmd an T E a f rotatioso auxiliarw fe t edgna a n yei heated tokamaks have been reported [11-12], however, whethe poloidae rth l plasma rotation lead suppressioo st fluctuationf no biat e clearth no s n s si I . electrode induced H-mode, detailed structure of ET at edge is determined, but the poloidal plasma rotation profile is mainly inferred from ET X B+. Does the plasma in the bias-induced H-mode also rotate poloidally with biasing doe?w Wharadiat e affecho si e th d s th i tlan e structurV e th f eo plasma fluctuation? What happen fluctuatioo st n power spectru mpresn i - ence of the poloidal rotation? These are some of the important questions need to be answered experimentally to get insight of the H-mode physics. In brief, although theoretical models predict strong influence of plasma rotatio fluctuationn no , experiment t demonstratesye hav t eno ddefinita e correlation betwee twoe nth .

In this paper presene w , t results showin ga connectio n betweee th n poloidal plasma rotation and fluctuation suppression with positive electrode biasing in a pure toroidal plasma having no poloidal magnetic field.

21 Poloidal rotation of the plasma with biasing and its radial profile at equi- torial plan measureds ei . Dependenc f Vbian eo go s voltagtoroidad an 4 e1 l

magnetic field Bv is determined.

The poloidal plasma rotation and fluctuation measurement experiments with electrode biasing were carried out in a toroidal device called BETA. It quasi-steadd an whic, s m 2 c h1. 5 1 decide~ y( = a , s plasmcm 5 4 ha= a sR duratio toroida) ncurrentlesT e Th . l kG magneti 0 s1. toroida o t cp u fiel v l dB plasm produceais dischargdby e betwee dia filamenmmm hot n 2 t tungsten wire and grounded limiter (i.d. = 18 cm, o.d. = 24 cm), similar to the ACT I device [14]. Typically, the discharge voltage Vd ~ -150 V, the discharge curren whilA 2 e~ t plasma densit electrod yan n temperature ear e rang10 X —i1 nth 1 3 0f eo cm" d 1-1 respectively3an V 5e . Sinca n ei pure toroidal device like BETA magneti,o thern e ear c flux surfaces rina , g electrode (i.d. = 10 cm, o.d. = 12 cm) is used to bias the plasma which also simulates the flux surfaces of a tokamak as far as biasing is concerned. Pulse positive bias voltage up to 450 volts having risetime ~ 1 ms and exponential RC decay time ~ 60 ms is applied to the ring electrode using capacitoa r bank. Detail toroidae th f so l plasma production, maintainance s equilibriumoit f , schemati BETe th f Aco devic biad ean s - circuide e ar t scribed in Ref.[13].

The temporal behavior of the applied bias voltage V& and the extracted bias electrode current Ir (or cross-field radial current) is monitored by a resistive voltage divider and a current transformer respectively. Single as well as arrays of Langmuir probes are used to measure the floating potential and ion saturation current, from which plasma density and its fluctuations are obtained. The poloidal plasma rotation is monitored using a Mach probe s depictea , Fign di . l(a)radialle Th . y movable Mach probe con- sista cylindrica f o s l insulator housindiam m m 5 gdiscs7. wite o on ,htw lookin t upstreaga casn m (i poloidaf eo l rotation othed an ) r downstream flow, mounted in it back to back. The assymetry in ion saturation current collecte thesy Mac e ddiscb o th ef tw h so probe will represen e poloidath t l flo f plasmwo a [15].

e positivTh e bias voltag e rineth gVfo t electrod, s applieewa d during the discharge (discharge duration ~ 1.2 s). Figs. l(b)-(c) show the time evolution of ion saturation current collected by upstream and downstream discs of the Mach probe for argon and hydrogen plasmas with Vf, = +50 V,5y, == 300 G. On application of the bias (depicted by arrow), we observe large assymmetr saturation io e th n nyi currents measuree th y db upstream and downstream discs for both hydrogen and argon plasma, and concomitantly significant reduction in fluctuations. The ratio R of measured ion saturation currents by the upstream to downstream discs at peak is ~ 6 for hydrogen and ~ 1.8 for argon plasma. This large current ratio implies significant poloidal plasma flow on biasing. The observed change in

direction of flow with Bv direction confirms poloidal plasma rotation. For plasma flow along magnetic field lines, from theoretical calculations [15],

22 S S DISC

a mo

TIME (!0ms/div) Fig. 1 (a) Schematic of the Mach probe, showing also its orientation used for poloidal flow measurement n saturatioIo ) (b , n current measurer dfo argon plasma with upstream disdownstread can m Mace disth f hco probe, saturation (cIo ) n current measure r hydrogedfo n plasma with upstream disd downstreacan m discarrowe figure Th th . n esi show tim t whicea h rine th g o biaelectrodt V s voltag0 5 e+ f appliedeo verticae Th . l scalr efo upstrea downstread man m discs signa samee ar l .

• eHYDROGE N • O ARGON

-12 -8 -4 12 MINOR RADIUS (cm) Fig. 2 Radial variation at equatorial midplane of peak value of ratio R of upstream to downstream discs measured ion saturation current with bias . V 0 5 o+ f

23 the ratio R of ion collection currents to the upstream and downstream discs kv d V 0.2T ~ lup/Ido^ d = - \ J i an ser R give e= | fo " y 9 nb wher1. ~ ek is the flow velocity normalized to ion acoustic velocity. Unfortunately, at presen theoro n t ycollection exisio r fo t Macy nb h prob flor efo w acrose sth magnetic field. Under this circumstance, we assume that expression for ratio R for parallel flow is also valid for perpendicular or poloidal flow to get leasat t approximate valu Vg.eof With this assumption measure,the d ratio

argor fo 8 n 1. implie = r poloidaR o s3 V<0. t— l rotatioX n 5 velocit1. = 8 yV 5 10 cm/sec (Te = 10 eV). For hydrogen plasma, from Fig. l(c), the ratio 6 R ~ 6 gives Ve ~ 2.8 X 10 cm/sec. At Mach probe position of r = +6.0 cm,.E V/c2 hydroger ~ mrfo n plasma implyin v X velocit B 7 X ~ r f gyE o 105 cm/sec. It is interesting to note from Fig. 1 that fluctuations reappear after rotation (or biasing) is over.

The measured peak value of ratio R as a function of minor radius r at equatorial midplan s plotteei hydroger fo Fig n di 2 .argod nan n plasma. For argon plasma, R is ~ 1 (implies no flow) up to ~ 5 cm and increased negative Th innee . th ecm n r i valuregio3 R 1 f eo + n = tor t aboua 5 4. t (i.enegativr fo . e valu show) r f eo s that relative amplitud saturation io f eo n current collected by discs of the Mach probe has reversed. In otherwords, the disc which was collecting more current in the outer region collects less in the inner region and vice-versa. This is expected for poloidal rotation of plasma. For hydrogen plasma too, there exists a region of large variation of R. Thus, for argon and hydrogen, for | r |> 5 cm, poloidal velocity Vg has shear. It is to be noted that before bias (i) for argon plasma, ratio R and5 1. ,s i r (ii t i fo ) , cm 5 3. = whil m r c t ea 5 3. > r r fo ise closon o et hydrogen, finite plasma rotation (R up to ~ 3.0) in opposite direction (rel- biae ativth s o ecaset ) along with large leve fluctuatiof o l observeds ni e Th .

dependence of ratio R at r = +6.0 cm with Bv and V^ is depicted in Fig. 3. The ratio R increases with Bv. On the otherhand, R increases up to a

- 5

cr o

• WITH Vb

oWITH Bf

0 30 0 20 0 10 0 BIAS VOLTAG VOLTSI E • ) oi .1.1.1__o.z 0.4 0.6 _ asi i TOROIDAL MAGNETIC FIELD (KG)

Fig. 3 Dependence of ratio R with toroidal magnetic field and bias voltage.

24 (o) • • • • 1.0 o o

0.6 o o

• BEFORE BIAS O DURING BIAS 02 0 4 8 12 MINOR RADIUS (cm) I

• **atea»-

TIME Il0m*/div)

Fig. 4 (a) Sn/nav Vs. minor radius r for argon plasma before and during applied bias voltage ) Oscillogramme(b , saturation io f so n current measure t threda e different radial location hydroger sfo n plasma e arrowTh . s show time at which bias voltage is applied

certain value with thed V&an n tends saturateo t r electrodFo . e biasin- gex

periment poloidae ,th l plasma rotatiov expectes nB i X drivee r b J o dy t n b force [16 radiae ] th where s i l T currenJ t density r earlieou n I r. experiment [13], we observed increase of Ir with Vb initially and then saturation (simila Langmuio rt r probe characteristics). Thu observee sth d dependencf eo agreemenn i s i v B d Vn tR bo wit an h poloida v forceB X l. T rotatioJ o t e ndu

fluctuatioe Th n leve (normalizen 8 l stationere no dth t av n )i y phasf eo plasma before biasin durind gan g biasin argor gfo s showni Fign ni . 4(a). oscillogramme Th Langmuif eo r probe measure saturation dio n current fluc- tuatio t threna e different radial locatio hydroger nfo n plasm depictes ai n di Fig. 4(b). Fig. l(c) shows time profile of density at r = + 8.0 cm. These figures clearly indicate that suppressio fluctuatiof no significans ni botr fo t h hydrogen and argon plasma, at locations where Vg as well as its shear exists (compar Sn/no s , cm e 0 wit5. hr argoo t FigFo p u . n .2) 0 plasma 1. ~ R , reductio t significant no decrease n s n 8 11.i t > 8n/n. , r no r 5t cm t A Fo . sbu hydrogen reduction r |<3. 8 | , r 5fo ,cm negligibls ni r appreciableo e only for very short duration e apparoacheon s A . s toward e largesth r V& region (or large r), the durations for which fluctuation are suppressed increases.

In the BETA device, at Bv & 200 G, the coherent peaks in power spectru fluctuationmf o s (without biased t frequencie)a 6,1~ s appeaz 2kH r rfo

25 hydrogen plasma. As Bv increases, additional peaks at various frequencies emerge d finall an se powe th y r spectrum becomes broadband turbulence.

The power spectrum of the density fluctuation at Bv w 200 G and ~ 500 G is depicted in Fig. 6(a)-(b). Observation of coherent peaks at low Bv and development of power spectrum towards turbulent at large Bv for density fluctuations without biasing has also been observed for argon plasma [17]. observnot Sincdo e appreciablewe e poloidal plasma flow wit biahout sat Bv ~ 200 G in particular for argon plasma, and measured phase difference between tin and fnp is much less than 180° which suggest that the coherent modes may not be a result of Kelvin-Helmholtz (KH) instability. In the BETA magnetie th , c field line toroidae sar Rayleighe th curvedr o (o ls d -)an Taylor (RT) instabilit mosa s yi t likely observee causth r efo d fluctuations. However, similar fluctuations are observed in both bad and good curvature

/nregionn S av d ~ ef

casey foune an w , n I d tha fluctuatione higd th tv an i.eh B w . bothlo t a s , coherent mode broadband san d turbulen suppressee b n ca t d with Vf,.

Theoretical work linean o s r stability analysi f variouo ] [3 s s moden i s presenc f poloidaeo l flow predic a criticat l valu givee r sheaV efo y f nb o r Vg = i/KyAx above which the mode will be suppressed. Here 7 is the growth rate of the mode, Ky is the poloidal wave number and Ax is the mode width in radial direction. The theory of turbulence with shear flow

by Biglari, Diamond and Terry (BDT) [4] also found minimum Ve' required for quelling the turbulence. Another model [5] on turbulence shows that e sheath r flo s morwi e effectiv suppressinn i e g long wavelength modes a s

compare shoro dt t wavelength t hig A v .whe hB n BETA plasm initialls ai y in turbulent state, BDT model condition is satisfied in the Vg shear region observethe for d G, coheren 200 ~ t v modeB « 0.2 [13]At , Ky ,takin . g

shea~ z Vg/L= e A rV w lengt ,d TTd 2 an ,an X hL z Re(ii>)kH ~ 6 7 ~ e v

v find agreement with theory.

In conclusion positive th , e biasin pura f go e toroidal plasma resultn si significant poloidal rotation of plasma and concurrently suppression of fluctuations measuree Th . d V largs gi hydroger efo n compare argoo dt n plasma. There exist a region of Vg shear and there reduction in fluctuation is significant. These results strongly suggest that the poloidal plasma rotation is capable of suppressing the fluctuations and is perhaps the reason for observed H-mod tokamaksn ei .

26 REFERENCES

. K.C1 . Shain E.C.Crumed gan , Phys. Rev. Lett , 23663 .9 (1989). 2. S.I Itoh.K. Ito, Physhand . Rev. Lett 22760, .6 (1988). . A.B3 . Hassam, Comments Plasma Phys. Controlled Nucl. 5 Fusio27 , n14 (1992). 4. H. Biglari, P.H. Diamond and P.W. Terry, Phys. Fluids B2, 1 1990. . Y.Z5 . Zhan S.Md gan . Mahajan, Phys. Fluid , 200sB5 0 (1993). 6. F. Wagner et al., Phys. Rev. Lett. 49,1408 (1982). 7. K.H. Burel t al.e l , Phys. Fluid , 140sB2 5 (1990). 8. K. Toi et al., Phys. Rev. Lett. 64, 1895 (1990). 9. R.J. Taylo t al.e r , Phys. Rev. Lett , 23663 . 5 (1989). . R.R10 . Weynant t al.se , Nucl . (1992)7 Fusio83 , n.32 11. R.J. Groebner, K.H. Burell and R.P. Seraydarian, Phys. Rev. Lett. 64, 3015 (1990). t al.e a , Id Phys . K . . Fluids12 255, B4 .2 (1992). 13. K.K. Jain, Phys. Rev. Lett. 70, 806 (1993). . K.L14 . Won t al.ge , Rev. Sci. Instrum (1982)9 40 , .53 . . K.S15 . Chun I.Hd gan . Hutchinson, Phys. Rev4728 3 1A . (1988). 16. R.R. Weynants and R.J. Taylor, Nucl. Fusion 30, 945 (1990). 17. G.Prasad et al., Geophys. Res. Lett. 19, 241 and 245 (1992). . H.W18 . Hende T.Kd an l . Chu Methon i , Experimentaf do l Physics edited . Lovberg. GriebR yH d man , Vol , par. 9 , p.34A t 5 (Academic Pressw Ne , York, 1970).

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27 XA9745739 CONCEPTUAL DESIG STEADF NO Y STATE SUPERCONDUCTING TOKAMAK SST1

TEAMT SS 1 (Presented Saxena)C. Y. by

Institute for Plasma Research, Bhat, Gandhinagar, India

Abstract

A steady state superconducting tokamak SSTl \B being designed at, the Institute for Plasma Research. The main objectives of SSTl are to establish scientific basis of steady-state operation of tokakamks, study the physics of the divcrtors. radiative layers and gas targets, steady state heat and particle exhaust and control, plasma confinement improvement and advanced tokamak configurations. SSTl tokamak envisages elongated plasma in a double null configuration, superconducting toroidal and poloidal coils, steady-state heat removal using actively cooled first-wall components, particle exhaust using external puuips, steady-state current driv plasmd ean a hca.iiug. The major parameter SSTf so l are: major radius RO 1.0= , mino5m - rra dius a = 0.20 m, elongation K. = 1.7-2.0, triangularity S = 0.4—0.7, toroidal field B.- = 3 T, average density 2 x 10l9m~3 and average temperature of 1.5 currene e healinkeVth Th .d tgan drive f wilICRHo l 5 W includ0. ,M 0 1. e MW of LIIOD. 0.2 MW of ECRH and 0.8 MW of neutral beams. Hydrogen plasm it will be produced. The conceptual desig SSTf no l wil presentede b l .

INTRODUCTION

A steady state tokamak. "SSTl" beins i , g designe Institute th t da Plasmr efo a Research. Bhat, Gandhinagar, India. SST larga s i l e aspect ratio tokamak configure doubla n ru eo t d null, elongated, triangular plasma, with a pulse duration ~ 1000 sees to obtain fully steady state plasma. The typical parameters of the machine are given in table 1. Fuelling will be by gas puffing. Introduction of Pellet injection method will be examined at a later stage. The TF &; PF systems and vacuum vessel are designed to allow later upgradatio high-currenta o nt 1.1,= large Q . 60.2= R m , a r9, m plasm kA 0 40 a wit= p hI with lower elongations. SST will pass through different phases of operation. Each phase will give way to the next when reliable operation and the physics objectives have been achieved. Phase IA will have

. Amardas'A Bahl. R , . BandopadhyayI , . BandopadhyayR , . BedakihaleV , . BhiseM , . Biju K ,. BisaiN , , D. Bora . ChakrabortyA , . ChandraD , . ChaturvediS , . ChennD , a Reddy . DeshpandeS , . GangradeyR , , J.P. Gaur, J. Govindrajan . JacobS , . JaishankarS , , M.R. Jana . KaurR , , P.P. Kaw . Khirwadkar,S , S.K. Mattoo . MukherjeeA , , H.B. Pandya, H.A. Pathak, P. Pathak, S. Pradhan, E. Rajendrakumar, R. Rajesh, D. Raju, A.K. Sahu, K. Sathyanarayan, Y.C. Saxena, P.K. Sharma, P. Sinha, R. Srinivasan, V. Tanna, K.J. Thomas, A. Varadharajulu, S. Varshney, M. Warner.

29 Tabl 1 eMACHIN E PARAMETER S( SSTl ) Major Radius RO 1.05 m Minor Radius a = 0.20m Elongation K 1.7-2.0 Triangularity 8 0.4-0.7 Toroidal field B0 3T / Plasma current IP = 220 KA Aspect ratio A — 5.2 Safety factor q 5.0-7.0 Average density < n > = 2 x 1013 cm-3 Average Temperature V -ke 5 1.

• Plasma species: Hydrogen • Pulse length: 100~ 0s • Configuration: Double-null Poloidal Divertor • Heatin Currend gan t Drive: Y LoweM\ 0 r1. Hybrid Neutral Beam 0.8 MW ICRH 1.0 MW Total power input to plasma < 1 MW. circular limiter plasmas. Phase IB will have shaped plasmas with LHCD in both limiter and divertor configurations. Phase II will involve full-power operation with a divertor configuration. Phase III will involve Advanced Tokamak operation. Table 2 shows typical plasma parameters in Phases I & II. In the table. "H-factor" refers to Te/TIT£R£9P, Paux includes lower hybrid power, T, n are in units of keV and 1018 m~3 respectively.

Table 2: TYPICAL OPERATING POINTS Parameter IA IA IB II II (Ohmic) (LHCD) #o(T) 1.5 1.5 1.5 2.9 1.6 IP (kA) 110 110 150 220 100 30 4 5 20 20 P«u: (MW) - 0.12 0.19 1.0 1.0

Zsff 1.7 5.0 5.0 1.7 1.7 *95 1.0 / 1.0 1.6 1.6 1.6 <595 0.0 0.0 0.44 0.44 0.44 TE (ms) 12 11 16 12 11 0.24 0.84 0.94 0.98 0.89 0.09 0.33 0.38 0.39 0.36 3(%] 0.18 0.08 0.12 0.13 0.39

3P 0.33 0.16 0.25 0.49 2.1 Bootstrap O.OS 0.03 0.05 0.12 0.53

fracs b f .

30 10

POUO H T - 1 16 2*if - SHALL peers 1 NBI PORT - STANDARS I o t 1 D PORTS

FIG lo - GENERAL SCHEMATIC OF SST1 (PLAN)

WC 16 - CEHtRAL SCHEMATIC Of SST1 (tlfTATIOH)

31 VACUUM SYSTEM

Fig. 1 shows a cross-sectional elevation of the vessel and cryostat. The vacuum vessel is •+J * all-welden a d continuous structure mad f SS-304Leo . wedge-shape6 Ther1 e ear d sectors1 . interconnectin 6 thick1 m c d an , g rings. Four interconnecting rings have bellows permio T . t long pulse operation up to 1000 sec, all plasma facing components are actively cooled with pressurised water flow. The vessel is capable of being baked at 525 K during initial pump down and 425 K during steady state operation vessee radiad Th . an l l ports permit human entry insid vessee eth r fo l the assembly and maintenance of in-vessel components. Fig. 1 shows the inboard toroidal limiter. which outboaralso tw oinnee e actth th s rdsa f stabilizero poloida e on d lan , limiters. A liquid nitrogen (LX2 reduceK ) shiel0 8 e radiativ t th sa d maseK hea4 s e t th loa n o d fro t surfacesmho e cryostaTh . t enclose e entirth s e assembly, excludin e centragth l borr efo the ohmic transformer. A 10~° torr vacuum will be maintained inside the crvostat. usine; O ^ turbomolecular pumps o reduct , e residuath e s conductioga l n heat load. Several vertical ports will be used for pumping. The maximum pumping requirement is 38 torr-liters/sec 19 3 (T-L/s) for operation with < ne > = 4 x 10 m~ . assuming TP = 4-£ = 24 msec. XBI will anothed ad r loaT-L/s4 f do . Henc pumpine eth g syste mdesignes i handlo dt steady-statea e T-L/s5 loa4 f do . This estimate ignores fuelling efficienc s puffingga y b y , which will push 19 3 e loadth p s u significantly. However x 10 2 < r Phases-m~ fo , > . wher ,e II n thi & Ie< s system wil e sufficientb l . Pumping wil turbomolecula6 e base1 b l n do r pumps, each witha capacit f 5.00yo 0 liters/s s puffinGa . maie g th wil ne b lmod f fuellingeo . COIL SYSTEM

There are 16 superconducting (SC) TF coils spaced uniformly around the torus. The coils have a modified D-shape. These produce a 2.9 T field at R with a ripple < 1% over the

plasma cross-section. Ther foue ear r pair up-dowf so n symmetri poloidaC cS l fieldQ (PF) coils. coil F producn P se ca th , eT 9 2. reference th = r Q B Fo . e m parameters2 0. = a 1-0= . ^ 5m Ft ,

a rang f plasmeo a shapes 1.7-2.0— X 0.4-0.75= K .x S . , wit rangha f pressureo d currenean t density profiles. These profiles yield /, = 0.65 to 1.25. j3p — 0.68 to 1.4. and

Lower 3P values are accessible only for higher KX values - this has implications for startup, where plasma presure would be lower. For the accessible shapes, the outer divertor has a slot geometry o permit , higa t h recycling regim impuritd ean y injection experimentse Th . inner divertor plat mountes ei d inboare closth o et d vesse attempo l n wall d an mad,s i t o et produc sloea t configuration. Fig show2 . e positioninth s f thrego e reference equilibria with respect to the first wall. Sinc coilC eS s mad f XbTeo i cannot tolerate high dB/dt. they canno usee b r breakt dfo - down and fast current ramp-up. Hence the ohmic transformer (OT) and correction coil syste mmads i F coif coppeindependens ei P o l d C systemS an r e th f .o t Ther alss ei opaia r of up-down symmetric vertical field (YF) coils made of copper, for use during fast ramp-up to 50-100'kA. DIVERTOR

divertoT SS e beins i r Th g designe alloo dt w high recyclin d radiativgan e divertor operation. The reference design of the divertor assumes 1 MW of total input power into the

32 9£>0

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200

100

000 7 05 ld.9. 7.0 5 | 18.9| 5 5 lll.8. B 11.2 51 .0 12.5 li0 0 10090 00 11080 0 0 70 1200 1300 1400

FIG. 2: Flu x surfac ethree plotth r es fo referenc e equilibria, with (/cx=1.9, <5x=0.64), (/cz=1.8, <5r=0.66d an ) («r=1.7, (5^=0.67), showinL e mappinSO th ge th f o g from the midplane to the divertor. The plot shows the separatrix and the surfaces separated from it by 1, 2 e outboarth n o m danc midplaned3 .

33 machine witradiatio% h20 nreference fro th e core r mth Fo . e equilibrium = wit1.8 X X hK 8 . = 0.66. the power input per unit plasma volume is ~O.S MW/m3, the power flow across the separatri s ~0.0xi 7 valu w MW/me energlo 8 keVf th ~ o e e d s i yTh . an 2 n outflo io r wpe

wil> e ln yiel edg< w dlo e densities. wilT lSS hav edouble-nula l (DN) diverto reduco rt e hea t platese load th divertoe n so Th . r

accomodatn ca reference3 e equilibria wit e followin0.64= hth - £ X K :& g9 1. shapes = X K : = 1.8 & &r = 0.66: KX = 1.7 & 5X = 0.67, as shown in Fig. 2. Because of high c55 values, toroidal connection lengths lie in the range 21-27 m. Give e smalnth l poloidal separation betwee e X-poinnth d outean t r strike point, both high recyclin radiativd gan e divertor operation requir esloa t geometr reduco yt e backflow of neutral impuritied an s e coreth o .t s Henc s beeesloa ha t n formeoutee th t ra ddivertor . whic s wheri h e pumping taking place e outeTh . r divertor strike poin s bee:ha t : p.aces da fas possibla r e fro e X-poinmth thao s t t field line e slofocusee narrowers th i sar t d an de Th . precise geometry of the slot will be computed later using neutral transport calculations. For double null (DN) operation, assuming 1:4 in-out asymmetry and 1.2:1 up-covm asymmetry in heat flow, we get a heat flow of 0.35 MW to each outer plate. The inner plates are designed for the worst case of SN operation, with 1:2 in-out asymmetry. Taking account of the inclination of the plates, this leads to peak and average fluxes of 1.5 and 1 MW/'m2 respectively. Allowing a safety factor of two due to various uncertainties, the plate must be designed for a peak flux of 3 MW/m2. The portion of the baffles bounding :he slot is designed for the same load. Steady-state cooling will keep tile surface temperatures below 1000° K. Divertor wilT platelSS havn si e demanding tile alignment requirements becaus f neareo - grazing incidence of field lines; the angle between the total magnetic field and the normal to the plate is 89° for some reference equilibria. Isostatically pressed graphite will be used as the basic armor material for firs: ^all components compositeC C- . s wil usee b l d onl locationn yi s where graphit t suitableno s ei . CRYOGENIC SYSTEM

e cryogeniTh c syste s mdesignei o handlt d e loads during three situations: cooldown. quenc normad han l operation e steady-statTh . e cryogeni0 c coil12 C hea S s i se t th loa n do h liqui1/ 5 d5 heliud an mK (LHe5 r curren4. fo t ) a t W leads. Closed-cycle operatios i n systemK 5 4. planne e . th r dfo magnen to 5 3 t e wil coolee Th b l forcey db d flow usin phaso gtw e helium (qualit >T 5' y~ wit hmasa conductorC s CI s flothroug e produceg/ e w th 0 rate e 3 cryoLH . f :h hth e o y -db genic system will first be stored in a Main Control Dewar and then circulated through the magne pume naturay b tLH r po eithen a l y convectionb r . Cooldown from room temperature wil K done b l5 day4 e 4. o reduco slowlt st o t 3 n yei thermal stresses. Cooldown requires more than four outlets for He from the refrigeration system at different temperatures between 300 K and 4.5 K. The heat load on the LX2 shield is ~25 kW at 80 K. The SO K shield requires an LX2 flowrate of 200 g/s in the panels.

POSITION AND SHAPE CONTROL

The vertical position control system must allow satisfactory operation over :he entire

range of plasma shapes and 3P, /, values expected in SST. The DX configuration in SST is up-down symmetric.

34 T vacuuSS r fro e e mplasmfa m th Th o vesse to o providt a s i l e effective stabilization. Hence a passive stabilizing structure in a saddle configuration is included on both outboard and inboard sides, as shown in Fig. 1. The stabilizer plates, made of DS copper, are 2.5 cm thic covered e plasma-facinkan th n do thicm c g k1 sid graphity eb e tiles. Ther singla s ei e toroidal break filled with high resistivity material to permit ohmic field penetration. e activTh e feedback coil consist paia f single-turf so o r n toroidal loops placed just behind the outer stabilizer, connected in a saddle configuration. Plasma position will be fed back

through a PID controller. The worst case equilibrium expected in SST. with K,X = 2. /, — l 0.96. has an open loop growth rate ~<0i — 36 s~ . For this equilibrium, the system is designed to control step perturbations up to 1 cm, and random disturbances characterized by (AZ)rms = 1 cm and a bandwidth of 70;. The corresponding power supply is rated at 10 kA. 20 V. and permits sle kV/s5 w e maximuo ratet Th .p u s m processing delay permitted between motioe e plasmth th f d activationo aan powee th f no r suppl msec2 s yi . F coilpair4 P e f so s havTh e independen controllerD PI t r plasmfo s a shap d radiaean l position control e extenTh f .curren o t t variatio eacn ni h coil e ,tim th A/PF ed scalan , e

of regulation. Ai5^ape. are being determined. Ipp is particularly sensitive to changes in £z. especially for KX = 1.7 and 1.8 reference equilibria. These will therefore govern the design of the shape control system. Lower bound e imposear s reasonso n A^apdo tw r fo e. Firstly, high voltages wile b l required because of the large PF coil inductances, a consequence of using superconducting coil r feedbackfo s . Secondly e dB/dth , t arising F coilfro P e self-fiel mth e s th mus f e o db t - thi belos s T/ impose w2 lowesa r boun ~0.f do 4 secuppen A . rimposee bounb y dn ma d o e couplinth Aijfcapo t e g du betweee n vertica radiad lAn l motion. HEATING AND CURRENT DRIVE

CYCLOTRONION HEATING

ICRF power of 1.5 MW will be coupled to the plasma. The number of antennas, and their dimensions, are chosen based on the following considerations. The maximum power density from each antenna shoul t exceedno kW/cm5 1. d e poloidaTh 2. l lengt limites hi d maximue th . cm 0 m4 openino t g permissibl passive th n ei e stabilizer. Antenna optimization require certaia s n rati f poloidaoo toroidao t l l length, yielding length, widt d thicknesan h s of 40. 8 and 1 cm respectively. Hence each antenna can radiate 375 kW. necessitating the

Table 3: ICRH SCENARIOS

Phase FREQ (MHz) PHASE V IA 45.6 0,0 88.7% IA 45.6 7T,0 75.7% IB 45.6 0,0 85.0% IB 45.6 7T,0 55.4% II 22.8 0,0 86.5% II 22.8 7T,0 70.0% II 24.4 0,0 87.0% II 24.4 7T,0 71.7% II 91.2 0,0 93.0% II 91.2 7T,0 87.0%

35 use of four antennas for SST. Two antenna boxes, each with two antennas, will be placed at adjacen fielw lo d t e sideportantennae th Th n . so s wil place e e shadob l th poloidaa n d i f wo l limiter to reduce the conducted heat flux. The separation between antenna strips is S cm. Different plasma heating scenarios have been examined for different operating conditions. harmonid viz2n . c heatin a singl r plasmn gfo e io minoritd aan y heatin ion-iod gan n hybrid resonance heating for a two-ion plasma. Optimization of the system includes maximization of damping near the plasma center. Damping lengths for the fast wave for different heating case e quitar s e smal densitw lo l e excepyth operatinr fo t g regime (phase IB) fine .W disa d - tinct advantag drivinn ei g adjacent element f phaso t eou s (dipole) relativ e in-phasth o et e (monopole) casee d reaimaginarTh an l . y e impedancpartth f o s e vary from 6 0.35. o 6t ohm3 6 ohm o st frod respectively1 an sm1 . Tabl showe3 e resultth s r differensfo t phasef o s operation. TJ denotes the coupling efficiency. ELECTRON CYCLOTRON HEATING

Penetration of the ohmic electric field thrrough the vacuum vessel requires a short L/R time of the vessel. On the other hand, a longer L/R time is desirable to shield the TF coils from dB/dt produced during plasma disruptions. Due to the latter, it may become necessary breakdowo t n witloow hlo p voltages, hence ECRH-assisted breakdow necessarys ni . ECRH wil usee b l d firsr assistinfo t g breakdow lated r currennan fo r t profile modification and localized heating l casesal X-mode n I th ,. e wil injectee b l d fro hige mth h field sider Fo .

breakdown-assist in Phase-I. with BQ = 1.5 T on axis. 200 kW will be injected at ^ve = 42 GHz. For BO = 3 T operation in Phase-II. an 84 GHz source is required with the same power. LOWER HYBRID CURRENT DRIVE

LHCD will be the main source of steady-state current drive in SST in Phases I & II. klystro50V 0k\ n amplifier wilz usede b lGH s6 . operatin4. Ove r o rz thi t eithega GH s 7 3. r

rang f parameterseo . A'||iOCC varies from 1.1 2.16o t . Henc e waveguideth e gril designes i l o dt e rangprovidth n ei y 1.5-2.5eA' . Tabl showe4 s result f waveguidso e optimizatioo tw e th r nfo frequencies. In both cases, the edge density is assumed to be 4 x 101' m~3. The efficiency of LHCD depends on target plasma temperature, which in turn depends e totath n l o power inpu e poweth te rPL ar LHCinputo t HC PIC, + e PI Ddu s & wher H ePL and ICRF/NBI respectively. Hence various combinations of PLH a^d PIC can drive the same 3 19 n eitheca e 0.6rw — W havx 10 5 M C ! 2 eP m~ — > e n < t currenta A k drivo T 0 . 22 e . Henc e MW LHC eth 1 1. D = syster onlH o . 0.8yPL — m s bein i 6 witH MW gPL h designed for coupling a maximum of 1 MW to the plasma.

Tabl : OPTIONe4 LHCR SFO D GRILL Parameter z GH 7 3. 4.6 GHz Periodicity (cm) 0.95 0.70 • Septa thickness (cm) 0.15 0.15 No. of wgds per row 16 16 ! Power flux (kW/cm2) 1.3 2.5 1 90°= N\6 )( \ 2.1 2.2 ! AA-H 0.47 0.43 j Reflection coeff. (%) 12 32 |

36 NEUTRAL BEAM INJECTION

SST will have a neutral beam injection system capable of delivering 1.7 MW into the plasma for heating and current drive experiments. Beam energies will lie in the range 30-50 keV/nucleon firs a intY s tA oa ke . step 0 systee 3 th ,t a m W wil usee M b l7 injeco dt 0. ~ t 19 3 target plasma with < ne > — 2 x 10 m~ . This will result in a temperature increase of ~ 0.7 keV and current drive of ~ 28 kA. The main system parameter followss a e sar : beam specie ° 'D°:sH multipole bucket type plasma source; ~10 0extractioA n current; 0.36-0.5 transmission. 10-100 s 0puls e length: horizonta d verticaan l respectivelym l 8 foca d lan lengthm :6 neutralizef o s r thicknesn i s h 1/ range 3 7 th —10x 5 consumptiod 1 e2 0. 1 6an molecules/cm LX h d 1/ an 0 5 e f no LH d 2an : respectively. DIAGNOSTICS

SSs fouTha r special features fro diagnostie mth c design poin viewf o t . Firstly shapea , d plasma means that 2-D measurements are necessary. Secondly, due to the different phases of operation wida , e rang f plasmeo machind aan e parameters rmis e measuredb : . Thirdly, long pulse operation has implications for diagnostic techniques, data aquisition. analysis and storage. Fourthly, ther accese ear s problemsloa o tt divertoe sdu r geometr passive th d eyan stabilizing structure. Quantitie measuree b o st d include plasma cxirrent. positio shaped nan .

3\ n, T and T in the core, edge and divertor regions, impurity concentration, radiated e t l

powerP , q-profile, divertor plate temperatur fluctuationd ean s ove widra e rang frequenciesf eo . Appropriate diagnostic systems have been designed. XA9745740 PHENOMENOLOGY OF START OPERATING REGIMES

C. REBEIRO1, M. GRYAZNEVICH, J. HUGILL2, I. JENKINS, R. MARTIN2, D.C. ROBINSON, A. SYKES, T.N. TODD, M.J. WALSH UKAEA Government Division, Fusion, Culham, Euratom-UKAEA Fusion Association, Abingdon, Oxfordshire, United Kingdom

Abstract START (Small Tight Aspect Ratio Tokamak) has demonstrated several advantage low-aspecf so t ratio plasmas with high plasma curren< p [I t SOOfcA] at low toroidal field [B+ > O.IT, Ip/Irod < 0.8, where B^ = n0Irod/2*R0 and R0 is variable], compact volumes [R0 < 0.34m, a < 0.27m, aspect ratio triangularit d - 3.0 Ran 2 1.2> 0,/a 1. , ~ elongatio0.8] < a y.6 b/ = nK With Ohmic heating alone, good stabilit current-terminatino y(n g disruptions), high energy confinement times [e.g., TE/TtfeoAicator ~ 2 at ne ~ 1 x 1020m~3, and more recently TE ~ 4ms(T£ ~ 2 x T/TER93-Hm«fc) at 20 3 nc ~ 0.53 x 10 7tt- with Ip ~ 120 kA)], hot plasmas [Te(0) ~ IkeV], and high #[/30(0~ 3.9%> ^ 0 ]1-9< 20% ) < havN , fl , e been observed. The vessel and the plasma-facing components are regularly boronised using deuterated tri-methyl boron in helium glow discharges. Naturally exhausted, single or double-null X-point plasmas are currently produced. In this study, values of the safety factor down to q^i ~ l.Q[R0/a ~ 90(95%d 1.2an ] 2.3[fl~ ) 0 /1.6]~ a , both usin ITEe gth R definition, have been found, so far, without the conventional plasma current terminating disruption. Further curren r densito t y increase insteae ar s d limitee th y db internal reconnection events (IREs). Thes valuew agreq elo f o s e with free boundar stabilitD yMH y calculations, carrieT.Cy b .t dRendeou t al.n e ri , which the external n=l kink instability dominates. 14 Values of IP/N down to 0.7 x 10~ Am, close to the Greenwald limit, 2 have been achieved [wher 7ra= eN Kn witd Murakame can ] hth i parameter upto 8 x lO19™-2!1-1. ^ Py& l3 Values of IP/N upto 1.5 x lQ~ Am are observed but with little evidence of runaways which is unlike the situation in large-aspect ratio tokamaks where such values of IP/N would create slide-away or runaway behaviour. Sawtooth oscillations have been observed both at low-q and close to the 19 3 density limit [ne = n™" ~ 8 x 10 m~ at Ip= ISOkA, B+ = 0.37T] until the IRE appears. Increase of radiated power at n™x supports the conventional explanation of the high density limit, namely the onset of MHD instabilities due to the current density peaking, associated with high radiative losses from the plasma edge. We also observe centrally peaked visible radiatio somn i e discharges, in which nc and CV line radiation increases. In these discharges a very peaked density profile is measured by Thomson scattering [ne(0) ~ 1 x 20 3 19 10 m~ , ne ~ 310 cm-3 which is associated with a decrease in Te(0) [from flattened/hollowea 200-40 d 100-15o t an V ] 0e 0eV e profileT d . Somf o e these "radiating core" discharges move fro mstabla e regio operatine th f no g diagram towards the Hugill Limit, but without the usual IRE. This investigation shows that the operational regime of START is broader than that in modern tokamaks, and moreover is bounded by IREs rather

Universidad Pauloo PauloSa o e Sa e,d , Brazil. 2UMIST, Manchester, United Kingdom. 39 than current terminating disruptions. Still broader operational regimes are expecte keV0 4 , I species(0.H ,d NB 5witMW e h,th Co-injection) whics hi now being commissioned. START results provid esignificana t extensiof no the usual tokamak data bas low-aspeco et t ratio.

This worjointls Departmenkwa K U y e fundeth - y Tradf b In do t d ean dustry and EURATOM and Conselho Nacional de Pesquisas Tecnologicas (CNPq), Brazil.

Introduction

START (Small Tight Aspect Ratio Tokamak) has demonstrated several advantages of low-

aspect ratio plasmas with high plasma curren toroidaSQQkA]w < lo p t [I t a > O.IT l fiel^ [B d,

Ip/Irod < 0.8, where B+ = fL0Irod/2'KR0 and R<, is variable], compact volumes [R0 < 0.34m,

0.27m< a , aspect ratio - R 3.0 triangularitd 0 2 /a> 1.2,1. an ,~ elongatio a < b/ 6 y= K n 0.8]. With Ohmic heating alone, good stability (no current-terminating disruptions), high energy 2 3

confinemen m~ 10 mord ,x an s te1 time recentl4m ~ T£/7x g ~ e [e s n E eot yT a Aicato 2 ~ r

TITER93-Hx 2 ~ (T~ 0.5 n mode) 310t x m-a 20 wit ~ 120fcA] I h t plasmaho , s [T(0~ ) E e p e

IfceV], and high /? [&(0) < 20%, /3 ~ 1.9, < 0 >~ 3.9%], have been observed[l). The vessel N and the plasma-facing components are regularly boronised using deuterated tri-methyl boron in helium glow discharges. Naturally exhausted, singl r double-nuleo l X-point plasma e currentlsar y produced.

g-Limi STARn o t T

In this study, values of edge safety factor q down to q^i ~ 1.0; q^, ~ 5.5 [R/a ~ 1.2] and 0

q cy~ 1.2, ; 9^(95% [R<,/a3 2. ~ 1.65]~ ) , both usin ITEe gth R definition, have been foundo s , far withou e conventionath t l plasma current terminating disruption. Further curren r densito t y increases are instead limited by internal reconnection events (IREs). An attemp studo t t aspece yth t ratie origith limiof , thid no influencq^ san t e limith s i n t eo presented. In figure 1 our results are compared to those predicted from the ERATO MHD instability code (T.C.Bender et al.). A reasonable agreement between the results from the experiment

d cod an t fla e 0/o=1.65 (2.2 against 2.0 s showni ) , wherea t .Ro/a=1.2a s 3 different values (5.5 against 3-4 e observed)ar s difficulti t I . clarifo t y y quantitativel e associate th e limit yth d an s d physics from this comparison threo t e e du ,reason t leasta s : 1-The external n=l kink instability mode predicted to cause the q limit from the ERATO code e onlb ye fuly th parl ma f causo t r thaefo t limit n additioI . e presenth n t calculations assume g(0)=1.15 which is also incompatible with the necessary (but not sufficient) condition for the appearanc f sawteeteo h oscillations , g(0 , commonlie ,1 )< y observe edgw lo e t da value . q f o s 2Perhap- inappropriatn a s e qchoic= ITERq f eo . This paramete s definee i equilibriur th t a d m whic obviousls hi physicae th t yno l plasma state immediately befor IREse th e , whe internae nth l inductanc increasins ei g wit peakine hth plasme th f go a current profile whic expectes hi occuo dt r just befor appearance th e IREe th f eo .

3- We have observed sawteeth oscillations before the IREs in the case of g=5.5, R0/a = 1.23. Experimentally, this shot was produced by reducing B^ during the discharge, and the maximum Ip/Irod (0-8) was obtained. In this scenario, the plasma current density (J) redistribution around possible th o c flattenint eT e du 1 q~ g sawtoote causeth y db he activitth n increasi y yma J V e| vicinity of q = m/n = 2/1 and the destabilization of the m/n=2/l mode could lead to the IREs.

40 • ERATO MHD INSTABILITY CODE O——— EXPERIMENTAL

q FROM ITER DEFINITION r

1.0 1.2 1.4 1.6 1.8 2.0 2.2 2.4

R0/a

Figure 1-Dependence of the edge safety factor q^, with the aspect ratio Ro/a from the experiment ancalculatione dth s (externa l kinn= l k instability) fro ERATe mth O Code.

1.4

1.2 1/2 •FTU-1993 1.0 I START-1995 1/q 0.8 cyl JT-60-199r-60-199lTCV1 TCV-199- 5 ASDEX-1993 ASDEX-U-1994 0.6

0.4

0.2

0.0 0.0 0.2 0.4 0.6 0.8 1.0 1.2 1.4

20 2 neR0/B^(10 /m T)

Figure 2-Operating plot from START and some modern tokamaks for Ohmic operation in L-mode.

41 1.2

G=1 (Greenwald Limit) 1.0

0.8 oio

0.6 8o O O O O o 1C 0.4 o o

0.2

UKAEA, C.RIBEIRO, NOV-95 I 0.0 0123456 2 lp(MA)/7ia

Figure 3-Greenwald plo r STARfo t T

Therefore sawtoote th , h activity perhaps precipitatee th e IREe b t sth sno appearancy ma t i t ebu main physical mechanism behind this phenomena. Indeed, there are frequent observations of IREs without any sawtooth activity precursor, and sawtooth activity with almost unaltered period and amplitude until immediatel hign i yhE befordensitIR e th e y regime.

Density Limits In figure 2, our operating plot is presented and compared with the results from some modern tokamak Ohmir sfo c operatio L-modn ni e (corresponding result relativele sar y scarc H-modr efo e obtained with ohmic heating only). The lines shown define the achievement in every operating regime [B^/UfRaqcyi, 1/Qcyl and ngRo/B^] to enable easy comparison of different machines. The operatio t ultrn a density w alo ,expecte e closth o et d runaway limi shows i t n onl STARr yfo T (little information about this possibl eothee limith n ri t device bees sha n explicitly reported). 19 2 1 A Murakami parameter up to 8 x 10 m~ T~ is obtained, whereas a very low-gcy/ (~1.0) distinguishes our device from the others. Here we treat qcvi as a technical parameter (q^, should be used for MHD stability studies) in the sense that it reflects the highly effective use of B^ in

START. For example, in JET Ip/Irod ^ 0.1, with 0.4 (the closest value to our 0.8) being obtained in CDX-U (R0/a ~ 1.5) at Princeton[2]. The TS-3 device[3] in Tokyo is able to obtain Ip/Irod ~3 at R0/a ~ 1.05 but, so far, only at low temperature (20eV) and in short (perhaps transient) discharges of order O.lms. 13 2 At very low density operation, values of /P/N up to 1.5 x 10~ Am [where TV = Tra Knc} are observe t witdbu h little evidenc f runawayeo s whic s unlikhi situatioe eth large-aspecn ni t ratio tokamaks where such value f /Nso would create slide-awa r runawayo y behaviour.

P l4 Abou densite th t y limit, value If Pso /N 10~x 7 dow0. Am,o nt Greenwale closth o et d limit, have been achieve 10 x 15QkA,= 8 19 0.37T = d„ m~~ / ^ andt * B a 3far o ,s ,usin ,n™ puffings gga . The Greenwald plot is shown in figure 3.

Sawteeth oscillations have been observed both at low-g and close t.o Increasn e of radiated e power at n™az supports the conventional explanation of the high densitmai y limit, namely the onset of MHD instabilities due to J peaking, associated with high radiative losses from the plasma edge.

We also observe centrally peaked visible radiatio somn ni e discharges linV eC d whichn i , an c n radiation increases thesn I . e discharge vera s y peaked density profil s measurei e Thomsoy db n 20 3 19 3 scattering [ne(0] ~ 1 x 10 m~ , ne ~ 3 x 10 m~ ] which is associated with a decrease in Te(0)

[from 200-400eV to 100-150eV] and a flattened/hollowed Te profile. Some of these "radiating core" discharges move fro mstabla e regiooperatine th f no g diagram toward e Hugilth s l Limit, but without the usual IRE.

Summary This investigation shows thae operationath t l regim f STARo e s relativelTi y broader than that in modern tokamaks, and moreover is bounded by IREs rather than current terminating disruptions. Still broader operational regimes are expected with the NBI [0.5MW, 40keV, H species, Co-injection] beinwhicw no g s hi commissioned . e possibilitTh connectioe th m/n=2/e f yo th d nan betweelE tearin externae IR th e nr th g o l n=l kink instability modes has been discussed; more results at q limit with different Ro/a are necessar clarifo yt y thi physicas sit limid an t l mechanisms. START results provid significanea t extensio usuae th f lno tokamak data bas low-aspeco et t ratio.

Acknowledgments

This wor jointls kDepartmenwa K yU fundee th y db Tradf o t Industrd ean EURATOd yan d Man Conselho Naciona Pesquisae d l s Tecnologicas (CNPq), Brazil.

References

[1] A.Sykes et al., in Symposium on Fusion Engineering(SOFE 1995) (Proc. 16"1 IEEE/NPSS, Illinois, USA) (to be published). [2] Y.S.Hwang et al/A.Morita et al., in Plasma Physics and Controlled Fusion Research(1994) (Proc. 15 Int. Conf. Seville, published)Spaine b o (t ) . th [3] Y.Ono et al., Phys. Fluids B5, 3691 (1993).

43 XA9745741 ERGODIC MAGNETIC LEVflTER EXPERIMENT AT THE SMALL TBR TOKAMAK

M.S.T. ARAUJO, V.A. VANNUCCI, I.L. CALDAS, I.C. NASCIMENTO Institute of Physics, University of Sao Paulo, Sao Paulo, Brazil

Abstract

A syste f foumo r Ergodic Magnetic Limiter (ELM) rings wa s projected and installed inside the TBR tokamak vacuum chamber, yielding a dominant m/n = 7/2 magnetic perturbations, within the plasma edge. werA 0 e applie50 e rings- Currentth A o e rang dt , 0 th whic 10 n ei s h correspond a rati r somo t oFo s . eIEML/I plasm1% = P a dischargesa , significant attenuation of the MHD activity, measured by the Mirnov coil system observes wa , d durin applicatioe gth externae th f no l perturbation. In other cases, however, an increase in amplitude of the MHD activity occurred leading, sometimes majoa o ,t r disruption.

I - Introduction

Ergodic Magnetic Limiters (EML) and Resonant Helical Windings (RHW) have been use createo dt t soma , e particular region plasme th f so a confine tokamaksy db , small resonant magnetic fields perturbations, usually refereed as Resonant Helical Fields (RHF). These imposed perturbations somn i , e circumstances, were observe improvo dt e the magnetic confinement, reduce the particle and energy losses and help the control of the inward impurity flux from the vacuum chamber walls [1-4]. However, maybe the most important result obtained, with this method of interaction with the plasma, is related to the possibility of controlling the onset of major disruptions.

As it is already well known, the disruptive instabilities are usually preceded by a fast amplitud emodeD growtMH sf o hand particularn ,i e majoth , r disruptione ar s commonly observed to be preceded by a sequence of minor disruptions. Among the different proposed mechanisms of controlling and possibly suppressing the onset of major disruptions [5], the use of resonant helical windings has produced interesting results e firsTh .t pioneer studies, showin w disruptivho g e instabilities are affected by resonant magnetic fields, were carried on the Pulsator

45 tokamak discharges, which showed thae occurrencth t majoa f eo r disruption coule db either delaye r activatedo d n responsi , a wea o t ek externally imposed magnetic perturbation created by a m/n = 2/1 RHW [6,7]. After this work, other investigations involving the application of resonant helical magnetic fields in different tokamaks have been performed, extendin e previougth s result o different s t discharges conditiond an s other perturbing helical modes [8 - 14]. Basically, these experiences showed that the application of RHF could strongly attenuate the Mirnov oscillations amplitude. This could mean that the intrinsic nature and the way the disruptive instabilities are triggered, could the investigatee nb thie d us swit windingse hth similar o , r apparatus.

2 - Experimental Set-Up

The TBR is a small ohmic heated tokamak, with small radius R0 = 0,30 m, limiter

radius ra = 0,08 m, toroidal field BT = 0.4 T, plasma current 7p < 12 kA and discharge duration up to 10 ms (fig. 1). In this toroidal device, resonant external perturbations can be created within the plasma magnetic surfaces by two different systems. One of them consist helicao tw f so l(fig1 , thawindings2/ 2) .t = runn sm/ ,d witan 1 h 4/ helicit = n ym/ acros entire sth e vacuum chamber toroidae th n i , l direction. Both winding externalle ar s y located. e otheTh r system correspond f fouo t r se poloida a o t s l rings (fig , toroidall.3) y distributed, that are connected to each other in a way to yield a dominant m/n = 7/2 magnetic perturbation, withi plasme nth a edge. The response of the plasma to these perturbations is observed by measuring the total perturbed poloidal magnetic field with an array of 16 Mirnov pick-up coils. These coils are poloidally distributed inside the vacuum chamber, in a given toroidal position, and they are not equally spaced but arranged in such a way that the toroidal geometry of the system is directly taken into account in our experimental measurements.

3. Result discussiod san n

max Fig show4 . typicasa dischargeR TB l , with plasmt aA . currenkA 5 t 8. Ip = t = 2 ms, an EML current pulse of IEML = 100 A was applied, which lasted for 1 ms,

46 GAS ANALYSER

O.H.T.

HELICAL WINDING PANEL

TOROIDAL FIELD COIL

VESSEL LOOP WINDINGS

Fig. 1 - Schematic picture of the TBR tokamak.

1 m/2/ n= m/n = 4/1

eef

vtf 27CT

•wo' 180* 180"

Fig. 2 - Location of the Resonant Helical Windings around the TBR tokamak. a

foue th rf o ring e s On tha - Fig t3 . mak Ergodie eth c Magnetic Limiter system, showin differene gth t spacing betwee wire nth e cellsn i , order to take into account the toroidal geometry.

47 Fig. 4 - Plasma discharge in which a low intensity RHF was applied, causing a visible attenuation of the MHD activity. The experimental signals showed corresponds to: (a) plasma current, (b) loop voltage, (c) horizontal position and (d) Mirnov oscillations.

25 3 Time (ms)

MirnoDetai- e Fig 5 th . f o l v oscillation externasL attenuateEM e lth y db perturbation. Note the small disruptive instability (a) at the same tim spikea e occur looe th pn si voltag e signal (b).

48 2.5 3 Time (ms)

externae th activit Attenuatio- D o Figt 6 e .l MH perturbatioydu e th f no n resonan1 4/ = n t m/ helica currenA e th 0 ln create30 i t a y db windings.

2.S 3 Time (ms) - Stron Fig7 . g activatio e Mimoth f no v oscillation observew no s i s d 1 RHW4/ = . n Therm/ e ar whee e th nn i applyin A 0 30 g indications that activitthin D i , s caseMH ye o increast th , e du s ei a possible plasma-wall interaction.

approximately (fig. 4a). During this period of time that the EML perturbation was being applied observee w , d thacorrespondine th t g loop voltage signa slightls wa l y smoothed (fig. 4b), whil plasme eth a column moved outwards (fig. 4c). At the same time, however, a significant attenuation of the Mirnov oscillation is detected in the magnetic coil signals (fig. 4d). This attenuation, and also the exact time a minor disruption occurred during the application of the EML, 20 jus before the termination of the imposed perturbation, can be better observed in fig. 5. The minor

49 disruption can be also identified by the negative loop voltage spike, followed by an increas plasme th f eo a resistivity (fig. 5b). The effect of a magnetic perturbation created by the m/n = 4/1, over the TBR discharges s presentei , fign di . Onc6 . e again a significan, t attenuatio e Mirnoth f o nv oscillations is observed, although there are some perceptible differences when comparing these results with the ones obtained by using the EML system. For example, when the magnetic coil fluctuating signals in figs. 5 and 6 are compared to each other, it is clear fluctuationD thae attenuatioth tMH e th f so n occur smoothea n i s r way, when external L systemperturbationEM e . th Anothee applief ar o s e drus interestinwite th h g characteristics of the RHW, that should be noted, is the 150 - 200 fJ.s time delay that are often observe thesn di e signals. Finally, when stronger currents (!RHW = 300 - 500 A) is applied to the resonant helical windings, the amplitude of the MHD activity sometimes do not decrease but, on contrarily, it is even enhanced (fig. 7). On the other hand, the results obtained with the application of strong current on the EML coils system did not alter significantly the overall results already presented, indicating that this latter system would be more convenient to work with, when external magnetic perturbations are to be applied to plasma confine tokamaksy db .

4. Conclusions

This work reports the results that were obtained when resonant external magnetic perturbations were created within some specific regions of the plasma confined by the ringsL f fou o smal tokamakR EM t r, se TB recentll e Th . y installed, showed thaa t significan t activitD attenuatio plasme obtainee b th MH f n ye o aca th df n o eve rather nfo r low perturbing current ringse th ) Also n % ( IEML/IP ,so 1 t shouli ,= stressee db d thao n t major disruptive instabilities were observed during the application of external magnetic perturbation wit ergodie hth c magnetic limiter system t trusame relationo n ei Th . s ei o nt the RHW system, that causes the plasma to disrupt, sometimes, even when the MHD fluctuations are under attenuation.

References [1] - S. C. McCool et al; Nuclear Fusion 29 (1989) 547. . LippmanS - ] ; Nuclea[2 al t ne r Fusio (1991)21311 n3 . [3] - Y. Shen et al; J. of Nuc. Materials 168 (1989) 295.

50 Takamur. S - Phys- ] . [4 al t .a e Fluid (19870 s3 ) 144. [5 T.N]- . Tod dNucl- . Fusio (1992)10712 n3 . . KargerF - ] . Wobig[6 H , . Corti . S 197, al Plasmn t 5i e , a Physic Controlled san d Fusion Research (Proc Inth 5t .. Conf., Tokyo, 1974) IAEA 1207. . KargerF - ] . Lackner[7 K , . FussmanG , . 197 al Plasm n t i 7e , a Physic Controlled an s d Fusion Research (Proc. 6th Int. Conf., Berchtesgaden,1976) IAEA 267. [8] - A. Vannucci, I.C. Nascimento and I.L. Caldas - Plasma Phys. Control. Fusion 31 (1989) 147. Itoh. [9K - ], S.I. Itoh Fukuyam. ,A Nucl- . al t .a e Fusio (1992)18512 n3 . [10] - A. Vannucci, O.W. Bender, I.L. Caldas et al. - II Nuovo Cimento 10D (1988) 1193. [11 D.E- ] . Roberts Sherwell. ,D , J.D. Fletche Nucl- . al t . re Fusio (19911 n3 ) 319. [12] - Y. Huo - Nucl. Fusion 25 (1985) 1023. [13] - J. Chen, J. Xie, Y. Huo et al. - Nucl. Fusion 30 (1990) 2271. [14] - M. Mori, H. Aikawa, Y. Asahi, T. Fujita et al. - Proc. 14th IAEA Conf. Plasma Phys Contrd an . . Nucl. Fus. Res., Wurzbeerg, Vol (19922 . ) 567.

mm PAG£(S) left BIAMK

51 XA9745742 PRESENT STATUS OF TCA/BR TOKAMAK

I.C. NASCIMENTO, R.M.O. GALVAO, A.G. TUSZEL, F.T. DEGASPERI . RUCHKOL , , R.P SILVAA D . , A.N. FAGUNDES, A.C.P. MENDES, E.K. SANADA . TARANV , , R.M.O. PAULETTI, W.P. DE SA, A. VANNUCCI, J.H. VUOLO, H. FRANCO, J.I. ELIZONDO, I.L. CALDAS, M.C.R. ANDRADE, A.C.A. FERREIRA, I.H. TAN, V.S.W. VUOLp, M.V.A.P. HELLER, M.Y. KUCINSKY, E. OZONO, M.S.T. ARAUJO, E.A. LERCHE Institute of Physics, University of Sao Paulo, Sao Paulo, Brazil

Abstract

tokamaA TheTC beins ki g partially reconstructe reassembled dan d Plasme inth a Laborator Universite Th Paulo o f y o afterwardSa d f yan ,o t si will be named TCA/BR. The first discharges are expected by June/July of next year. The main scientific objectives envisaged for the machine are: AlfVen wave heating and current drive, confinement improvement, disruption turbulenced san thin I . s pape alse rw o describe presene th ) (i : t projectstatue th f so , (ii diagnostie )th c systems, (iiicontroe )th datd an la acquisition system, (iv) the RF system for the excitation of AlfVen waves, thae beinar t g developed d alse resultan th o, f predictivo s e transport simulations of its performance.

1. Introduction

The TCA/BR Project has as main objective to put into operation, at the Laboratori Universidade Fisice od th f ao Pauloo Sa tokamae a ,ed k adequat perforo et m research at a reasonable level of competitiveness [1]. In this paper we report the present statuprojece th f so t transferres tokamae wa Th A kTC d from Lausanne, Switzerlan sincd dan e March 1994 its being partially reconstructed and reassembled. TCA/Be Th tokamaa Rs i k very appropriate magnetir dfo c confinement researcn hi problem f interese controlleo s th o t t d fusion s maiIt . n parameters are: major radius

0.6= ImR , minor radiu 0.1= , magneti s8plasm, a m T 2 1. 0 ac curren- 12 fiel T = dB p I t 13 3 , plasmkA a duratio , maximums 0 n 10 mT p= averag 10x e8 densit 7. cm" = ^ yn , electron

. temperaturn eV io 0 d 40 an temperatur= V t e e T 0 75 = e eT

53 presente Ath t projecstatue e folloth th ,s a f so s i t w The annex building with an area of 260m as well as the high voltage transmission 2 line are scheduled for completion in February/March 1995 The installation of a 2 Ton crane and the air cooling system have been recently completed The tokamak assembling has already started and the parts corresponding to the supports, the bottom base of the tokamak and the OH transformer, are in place The toroidal field power supply is ready for e ohmitestinth d c gan healin verticad gan l fiel% 0 d9 powed an r% supplie0 7 e ar s completed, respectivel discharge Th y e cleanin preionizatiod gan n system alse sar o about 90 % completed The vacuum vessel pumping-down is expected for March/April Accordin schedulr ou o g t expece ew begio t teste nth energizatiof so coile th sf no in May/June and to fire the first shots in July/August The experimental research program coul probable db y initiate secone th n di d semeste f nexro t year e followinTh g line f researco s projecr w thihfo ne s t were chosen taking into consideration competitiveness, feasibility and the requirement of international programs Alfven wave heating and current drive, confinement improvement, disruptions and turbulence

2. Diagnostics

e diagnosticTh e installe b e firs th o t s t n i dphas f tokamao e k operatioe ar n interferometer, ECE, visible optical spectrometer, soft X-rays detector array, neutral particle analyzer, visible bremsstrahlung, Mirnov coils array and, finally, diamagnetic coil In a second phase, the diagnostics to be introduced are VUV espectrometer, Multipass Thomson Scattering, Microwave Reflectometer, Soft-ray Spectrometer bolometry array and, possibly, neutral and charged particle beams Fig I shows the diagnostic layout for the machine z interferometeAchannel5 GH 0 e homodyn15 s th f o r e type, sin/cos detection system wil e use b r electrol dfo n density measuremen microwave Th t e source wile b l based on Impart Diode with power of 10 - 15 mW The electron cyclotron radiation for TCA/BR is expected to reach frequencies up detectioe th r Fo f thino sz radiatioGH 0 15 t wilni use e o t b l heterodyna d e detection system tune harmonid deteco dt 2n extraordinare E th t cEC y mod locae Th el oscillator will be of the BWO type with a power of 10-30 mW, and frequency sweeping time of 100 u,s or fixed frequency

54 TCA/BFi- 1 g R diagnostic layout

The interferometer and the radiometer will be bought from the commercial branch of St Petersburg State Technical University The optical spectrometer will be bought in the international market with the following characteristics focal length of 1 meter;

o resolution of 1 A , f = 6.8, ranging from 200 to 800 nm, equipped with CCD detector array. An arrangement to measure light intensity along several plasma column chords will be used to yield ion temperature and impurities radial profiles The neutral particle analyze obtaines rwa d fro Culhae mth m Laboratory, England, and has 20 detectors (channeltrons) distributed in two arrays of 10 detectors each The Mirnov detection system will be based on 4 magnetic coil arrays located in different toroidal positions insid vacuue eth m vessel Each array will hav poloida2 e2 l pick-up coils The coils will not be equally spaced, in order to take into consideration the tokamak toroidal geometry The Thomson scattering syste s mbeini g develope n cooperatioi d n witr D h M Machida from UNICAMP - University of Campinas, Brazil A FIR laser for Faraday rotation, density measurements and VUV spectroscopy are also being developed in cooperation with UNICAMP Work will begin shortly for the development of soft X-ray array, spectroscopy system, bolometry and reflectometry

55 . Plasm3 a contro datd an al acquisition

The control of the TCA/BR discharges can be accomplished with the old system, formerly used in Lausanne, which is composed basically by timers to trigger the different system tokamae th f so k This equipment, already checked, everythin goon i s gdi condition, and can be used if necessary. The Data Acquisition and Control System will have two workstations; one for tokamak control and data acquisition, and another for storing the data transferred from the first computer, with access from external users. The system is being dimensioned for a maximum capacity of 10 MB of digital information per tokamak pulse, in about 400 channels, and covering 100 ms. Due to limitations in costs, in the first phase of the tokamak operation we will have the computer capability of only 1 MB/pulse, related to some channels coverin ands m beginnin,e 0 othersth g10 n I . wile , usinge onlms w lb 0 y1 g CAMAe th C modules already existin laboratore th n gi mosd yan t Viepossiblb wLa e yth Software, for control, and IDL package, for handling the data. Fig. 2 shows the block diagram of the system. Latter on it is intended to change the whole system towards the use VME modules. This system will be developed in cooperation with the Centro de Fusao Nuclear locate Institute th t da e Superior Tecnic Lisboe od aPortugal- .

. Alfve4 systeF nR m

e radio-frequencTh y syste r excitatiomfo f Alfveno n waves wil e substantiallb l y modified wit previousle h on respec e th o yt use Lausannn di e antenn[2]e Th . a system will

Toktaulc Room

Bloc- Fig2 . k diagra controf mo datd an la acquisition system.

56 A-A

Antenna plates

Side protector

Fig. 3 - Top view of the antenna positioning and the electrical arrangement of one loop pair.

mse0 c10 1 0- 4 - Phase oacUIator Power supply t" z - & mt 0 - 20 W (4x400 V* power trlfkds) - O 2I Kt

Butler unit (21200 kff Power trkxte)

FigBloc- 4 . ksystem F diagraR e th .mf o

Fig. 5 - Time evolution of plasma parameters with Alfven heating power of 500 kW, for q = 2.7 and densities 10x 5 (B). 15 m 9 = 10x > - md 3(A)an

Time(sec) Time(sec)

57 be a multiloop antenna array consisting of poloidal half windings with phased currents A top view of the antenna location and the electric scheme of one loop pair are shown in fig 3. Four similar units wil installee b l d employing antenne th a feeding ports already available on the tokamak vessel The antenna loop will be fed though an independent matching circuit blocA . k diagra systee th f m mshowo s i Fign i . 4 . The main objectives of the new RF system for the TCA/BR project are the following 1) single helicity mode excitation, 2) the possibility of generating standing and traveling stud o wavest currens F a yR o s , tcontro o t driv d ean l plasma parameterse th ) 3 , possibility of using the same antenna array for excitation of different modes [m = 1/n = 2, powe plasmaF e R th f o o r t .W M 1 deliveo t o ) t 4 p d ru an , 4] , 6 = n 1/ = m , 4 = n 1/ m= The generatoe desigF antennR th f e no s th alreada ha f rsysteo yd man bee n completed. Construction and testing of an antenna module, using a small 30 kW generator is currently under way. The complete multiloop antenna system will be constructed locally generatoW m 1 ane rdth wil orderee b llocae th lo d t industry .

5. Transport Simulation

In order to investigate the performance expected for TCA/BR under Alfven heating, a predictive calculation was carried out using a new version of Dnestrovsky transport cod anomaloue [3]T eA Th . s electron heat diffusivir s takei y 1 l n T- fro e mth transport model [4] and additional heating is properly simulated by solving linearized kinetic equations shoe Fig n timw I e . w5 .th e evolutio maie th nf n o parameter plasma f so a discharge with an additional heating power of 500 kW. The value of the cylindrical safety x 3 = averag d Fign > i factoan 10 x 3 nq th 3i e ,electro th

58 . Conclusio6 n

The TCA/BR Projec developins i t g accordin expectatione th o gt laboratore th f so y staff Nevertheless speneffore e b ,th puttin n o i t t machine gth e into operatio initiatd nan e the scientific program is substantially large compared to the man power of the plasma group The scientific cooperation with colleagues has been very important and we expect that these cooperation can be increased and extended in order to full explore the capability TCA/Be th f o R tokamak

References

[1] - 1C Nascimento et al - "Project TCA/BR A Middle Size Tokamak Facility in Brazil" - Proceedings of the 1994 International Conference on Plasma Physics, Foz do Iguacu - Brazil, Oct 31 - Nov 4,1994, Vol 1, 69 (1994) [2] - L Ruchko, M L Andrade, R M O Galvao, I C Nascimento - "RF System for Alfven Wave Heatin Currend gan t Driv TCA/Bn ei R Tokamake "th - Pro f o c 1994 International Conference on Plasma Physics, Foz de Igua9u - Brazil, Vol 1, 364(1994) [3] - Yu Dnestrovskyy and D P Kostomarov - Numerical Simulation of Plasmas - ] Merezhkein[4 MukhovatovS V , PlovojR A , , Fizika Plasmy

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59 XAQ745743 HL-1M MODIFICATIOE TH , HL-F NO 1 TOKAMAK

HL-1M TEAM (Presented Deng)W. X. by

Southwestern Institute of Physics, Chengdu, Sichuan, China

Abstract

HL—1M tokamak was modified from the HL—1 tokamak by replac vacuuw ne a m g vessein l without copper shell,for high power auxiliary heatin mord gan e effictive diagnostic purposes feew dne baca , k plasma position control system for plasma control. Rearranged poloidel magnet ic system optimized the magnetic flux consumpotion ,so that plasma cur rent ,an druatioe dth f dischargno e ,wil increasee b l , meanwhile eth power supply system of HL—l,has been reformed to fit HL—lM,in conditioe th f higno h curren higd an th power auxiliary heating. HL—lM,the modification of HL—1 tokamak is a medium size toka ma r developmenkfo f powerfuo t l auxiliary heating hybriw ,lo d current drive technolog plasmd ean a control, including plasma density control meany b f pelleso t injectio speeiad n an puffs ga l . refore Th m laste yeare firsde on tokamath , tM 1 operatio— kHL f no

had following results: plasma current Ip=322A dischange duration Td= 1. 04s,toroidal field Bt=2. 5T,line average electron density iie~~4. 2X 19 3 10 centra/md an l electron,temperature Te~l. OKev. Comparing HL— Tokamak1 , HL demonstrate—M 1 d good capability for both plasma parameters and tokamak plasma physics investigations.

61 In this paper, the HL—1M tokamak device ,including feed back plas ma position control system, power suppl auxiliard an y y systee d ms i scribed ,the first operation results are also discussed.

1. INTRODUCTION HL tokama—M 1 s modifiekwa d fro e HL—mth 1 tokama replaciny kb ga new vacuum vessel without copper shell, for high power auxiliary heating and more effictive diagnostic purposes feew ne d a bac, k plasma position control system for plasma control. Rearranged poloidel magnetic system optimized the magnetic flux consumpotion o thas , t plasma e druatiocurrentth d f an o n, discharge, will be inerease, meanwhile the power supply system of HL—1, has been reformed to fit HL—1M, in the condition of high current and high power auxiliary heating. HL—1M, the modification of HL—1 tokamak is a medium size tokamak for developmen f powerfuo t l auxiliary heating hybriw lo , d current drive tech nologe and plasma control, including plasma density control by means of pellet in jectio speciad nan s puffga l . Comparin demonstratetokamakM 1 1 — — L L gH H , d good capabilitr fo y both plasma parameter tokamad san k plasma physics investigations. In this paper, the HL — 1M tokamak device, including feed back plasma position control system, power suppl auxiliard yan y syste mdescribeds i firse th ,t operation results are also discussed. Desig completes nwa 1992n di . Manufactur f vaccueo m vesse s finishewa l d in June, 1994. Power supply refor d othean m r facility maintenance were complete n i Julyd M ,1 1994e — firs Th L t.H operatioe th f o n tokamak under ohmic heating conditio s ha achieven d following

results; plasma current IP= 322kA, discharge duration Td= 1. 04s, 13 toroida , lin5T e . l2 average fiel= t B d d electron density n^=4. 2X10 3 cm~ centrad an l electron temperatur keV1 — e . T e Fig show1 . a s cross-section view of the device. The HL-1M main parameters are listed in Table 1.

62 Table 1 HL-1M Main Parameters

Major plasma radius, R0(m) 1.02

Minor plasma radius, a(m) 0. 26

Aspect ratio. A 3. 9

Plasma current,Ip(kA) 350 Toroidal field, B«(T) 3.0

fluH O x) s swing• V $( , 1.75

Total auxiliary heatin- g power,Pn(MW 2 ~ )

Current driv1 ~ e power, PCo(MW)

13 3 Line averaged electron density,ne(10 cm~ ) 5 — 7

Central electron temperature, Te(0)(keV) 1. 6 — 2. 0

Centra temperature,T,(0)(keVn io l . 8—10 0 . )

Energy confinement time, TE(ms) 30 — 35

2. COMPONENTS OF THE DEVICE 2. 1 TF Magnet F magne T coppe6 1 e f mads i o Th t rp u ecoil s whic usee har n di e HL-th 1 tokamak, four pancakes each 7 turn 1 r ,pancake pe s l al , connected in series and indirect water cooled. The field at the center e fielTh d . e vesseripple outeoth th 3T f t s ea i rllimitee edgth f eo r does not exceed 3%. The TF windings are installed in nonmagnetic steel cases, which transmit the electromagnetic force generated in the coil to the central rings and the adjacent supporting structure. 2. 2 PF System

The technical cosideration f reconstructeso syste F followss dP a e mar : • The iron transformer, which consists of an iron core, primary coils and bias coils s beeha ,n reuseHL—1e th n di M PFsystem. • A pair of outer PH coils is added to make full use of the flux swing of the iron core.

63 Fig 1 .Cros s Section Vie HL-1f wO M 1———iron core 5— -Outer ohmic coil 12- -Outer wedgec bi 2———Vacuum vessel 7— -Toroidal field coil 13- -Inner wedged strip 3——Vertical field coil g— -Shieid. plate 14- -Central rings 4———Horizontal field coil 9—- 15 -Basepiate -Lead wir f toroidaeo . 5——Fast feedback vertical 10- -inner ohmic coil field coil fieid coil H- -Bias coil

64 • The internal vertical field coils which were used in HL — 1 have been removed pai A f fas. o r t vertical field coil r plasms fo wer M ea1 installe— HL n di horizontal position control. • A pair of horizontal field coils, is added for plasma vertical position feedback control. The PF system of HL — 1M is composed of ohmic heating system, slow vertical coils, fast vertical coils and horizontal coils. The PF coils are symmetrically e lowearrange th e uppe d rth an t placera d n relatioi se th o t n equator plane of the torus (see Fig. 1). Because of space limitation, the slow and fast vertical field coil e closesar eaco dt h other decouplinA . g transformes wa r designe reduco dt coupline eth g between these coils. 2. 3 Vacuum Vessel and Pumping System In orde enlargo t r e window vacuu e portd th san f so m improvo vesset d an l e acces plasmao st vacuue th , m vesse bees ha l n reconstructed. vacuue Th m vesse f HL—1o l Msingls i e walled chamber divides i t I . d into four segments, each composed of one figid section made of 25mm stainless steel (304L) and one bellow made of Inconel type steel (GH39) 1mm thick with high electrical resistance. The thick wall of the vacuum vessel gives the shell effect, s timit ed constanan ture on n e toroidas 7msi tTh . l resistanc vessee th f s i elo 2mQ, whics mainlwa h y provide y bellowsb d e segmenTh . t jointe ar s fanged together with bolts vacuue Th . m tightnes t thessa e joint gives si a y nb Viton o-ring. Major radius of the chamber is 102cm, minor radius 32cm. It provide port4 5 s s giving priorit auxiliaro yt y heating plasmd san - adi agnostics e vacuuTh . m vesse bakeabls i l 150°o t p poloiday eu C b - in l duced current heating. Ther fixe e movabld ear dan e limiters madf eo graphite inside vacuum chamber. The pumping system of HL-1M consists of three turbomolecular pumps (150 / /s,each0 ) use s maida n pumps during operatioa d nan

cryogenic pump , (fo2 800//s r//sN 0 r H90 2;fo O ,- ) in use o t d crease the pumping speed for water vapour. In addition, an ion sput- tering pump , sustainin/s (20 I 0 g pressure 10~4Pa s usei ) s susda - taining pump during no-operation period.

65 3. POWER SUPPLY SYSTEM e poweTh r supply syste HL-1e th f mMo tokamak consiste th f so toroidal field, ohmic heating, slow vertical field, fast vertical field, horizontal field, auxiliary heating and baking power supplies. Two flywheel Moto-Generator are used as main power supply. e maiTh n parameter f eaco s h Moto-Generator are:

Apparent power 80MVA

Connection " wit"Y h o phastw e shif f 30°o t C

Frequency 49.5-44Hz

Line voltage 2096V

Current (for each "Y") llkA

Energy release 100MJ One of them powers the toroidal field coils via diode rectifiers and auxiliary heating system via converters. The other supplies all the poloidal field coila correspondinvi s g controlled converterse Th . specificatio e poweth f o nr suppl s showyi Tabln ni . 2 e

Table 2 Specification of the HL-1M Power Supplies

Load Voltage output(kV) Current (kA) Feature

Toroidal field system 3.86 14. 1

Ohmic heating system 0.8 8.0 12-pulse, feedback control

Fast vertical field coils 0.4 3.5 12-pulse,-feedback control decoupling transformer

Slow vertical field coils 0.2 6.0 12-pulse. programmed control decoupling transformer

Horizontal field coils 0. 01 3. 5 6-pulse, feedback control

Flux bias coils 0. 17 1. 6

Baking AC:0-0. 4 AC:0 -0. {§ feedeb d y the utility of 220V Auxiliary heating 75 30A , 3s

The control system of HL-1M Power Supply is composed of central and local control systems. It enables all or a part of the power syste operato mt e unde controe e centrath r th f o l l control syster mfo

66 experimental discharges r separatelo , y unde e controth r f theio l r n locaow l control system r testsfo s . e centraTh l control system mainly consist logia f so c controd an l program units e locaTh . l control systems have thein logiow rc control and feedback control units. Programmable logic controllers are adopte r logidfo c control personad an , l computer r prografo s m control, feedback control, engineering data acquisition and display 1 respectively.

4. AUXILIARY SYSTEM e auxiliarTh y syste HL-1e th f mo M tokamak, including plasma heating, lower hybrid current drive, pellet injection and pumping limmiter s beeha , n developed. 4Plasm1 . a Heatings

•Neutral Beam Injector Power ~1MW Energy 20—40keV Pulse length 150ms •Ion Cyclotron Resonance Heating Power ~1MW Frequenc —50MH0 3 yz Pulse length 10—100ms •Electron Cyclotron Resonance Heating W 5M . 0 Power Frequency 75GHz Pulse length 30ms 4. 2 Lower Hybrid Current Drive . Powe. r ~1MW Frequenc . 45GH2 z y Pulse length 100 —1000ms 4. 3 Pellet Injector and Pumping Limiter

•Pellet Injector solid H2/D2 Dimension of the pellets <£1. OXLl. 0 — 4mm, 4 chots o . sr 4XL1— . 4 Omm4 Ol 2 . , Velocity 700 —1200m,'s h e consumoito—t/ 4 H 6 L n

67 •Pumping Limiter Exhausted'rate e^>05 0 . Hea carboe t th loa n o dn tile 500w/cnv

5. COMMISSIONING Up to now, two runs of the HL-1M havebeen carried out. First s frowa m Oct . 19915 . Nov4o t , 199 28 .d secon 4an d from March . followine 1995secone 8 th th y . n n 199I .d ru 13 Ma go 5t parameters have been obtamded;

5 2. Toroidal field, B,(T)

2 32 Plasma current, Ip(kA)

Line averaged electron density,n,(1013cm~3) 4. 2

Central electron temperatur1 ~ e ,Te(0) (keV)

I Discharg . 5 e Cleaning The reduction of the outgassing rate and related impurity con- f primaro ten s i t y importanc HL-1Mn i e . Three methods have been adopted: a) baking up to 150°C b)DC glow discharge cleaning c)Taylor discharge cleaning e bakinTh s realizei g y meanb d f poloidao s l curren n vessei t l wall inducecurrenC A y b dt througF coilse T maximu e Th .th h m temperature difference between different e pointwal th s lesi ln i ss than 30°C, if wall temperature rise rate is below 10°C/h. After 48hrs bakin t 150°a g d 22hrCan s continuous pumping n ultimata , e base pressure is 5. 6Xl(T5Pa, and total leakage rate 1. 2X10~3Pa m3/s. Helium glow discharg s producewa e d between anelectrodd an e the grounded vessel wall without RF add. The voltage between the electrod e wal th s 400V i ld ean , curren, heliu5A . 0 tm pressurX (5 e 10-2-l)Pa. Taylor discharge cleaning is very effective for reducting oxygen impurities absorbed in the surface of the wall. The typical parame- 2 4 : Ptere K:ar s=(1-1 . 2)X10~ (300-400= Pa, B , 10~)X T, VL=10

-15V, Ip = 1.8-2kA, Td = 4-6ms.

68 Plasm2 . 5 a Position Control Owing to the absence of thick copper shell, the plasma position control in the HL-1M becomes very severe. Besides slow vertical field coils, the fast vertical field coils and horizontal field coils were installe r plasmdfo a equilibrium e fasTh t. vertical field coil supe ar s -

t 3

<3 Ar-r-^-bvP^ -I

ICO 300 600

Fig. 2 Displacement oscillograms for a series of sequent shots with the

same discharge parameters.

69 925= (94 —11— 27)

:,sec Fig 3 . Compariso plasmf no a currents since start-up.

Table 3 Main Diagnostics for the First Phase of Operations

Diagnostic Measurement

Magnetic Probes Loop voltage, Plasma current, Displacements, Magnetic oscillations. Diamagnetism

Electrical Probes Potential, density, temperature and Mach number in the edge

HCN Laser Interferometer Line averaged electron density

Ruby laser scattering System Electron temperature

Visible Spectrometer lighd an t. impuritH y spectrum

VUV Spectrometer Impurity spectrum

One-Dimentional Arra f Silicoyo n Detectors Soft X-ray fluctuation

Hard X-ray Monitors Effects of runaway electrons

Bolometers Radiated power Collective Sample Probes Effects of plasma-wall Interac-

plied by four-quadrant thyristor converter, and the P-controller is used to make the closed loop control dynamically optimum. The horizontal field coil e suppliesar threy db e phase bridge thyristor conver- tor which is controlled by the minimum time bang-bang controller. e sloTh w vertical coil 2 pulse feede1 ar s y eb d thyristor converter with programmed control.

70 3OO

2OG

Fig 4 . Plasma current,Safety facto displacemend an r t oscillograms.

Wit e abovth h e feedback controlling syste workmn i e gooth , d stability and reproducibility of the discharges were obtained. In a se- rie f shoto s s wit e samth h e discharge parameters e plasmth , a displacements can be controlled within 1cm during flat-top of the discharge (se e maximueTh Fig . 2) . m discharge duratio s 1040msni . 5Ohmi3 . c Discharge HL-1 s operateMha d sinc October4 2 e , 1994, with ohmically heated plasmas (BT.= 0..5_— 2. 5T, IP<322kA). A comparison of plasma currents is shown in Fig. 3 to show the progress made in im- provin e plasmgth a since start-up e diagnostiTh . c item r thi- fo sop s erational phase are lis'ted in Table 3.

71 The experiments have been done using hydrogen plasma. The main purpose was to get some preliminary data on the equilibrium stabilitiesD an s founMH d wa dt I . tha e stablth t e reqion.of plasma displacement s emergewa s d mainle higheth n yi r field side (stability existe e displacementh f di s leswa st thar mosfo n tm aboudis5c . 1 -t charges), and the low safety factor plasma (q

6. CONCLUSIONS e reacW h following conclusions concerning HL-1M: ) Designa , fabricaito assembld an e ndevic th f o ye e provb o t e gratifying. The engineering performances meet and in some respects surpas e desigth s n requirements Compare) b d wit e HL-th h 1 tokamak e maith , n advantagef o s

the HL-1M tokamak are: o improv•t e acces plasmao t s , e fluo makth •t x f o swinge e fulus l , o contro•t l plasma position verticall horizotalld yan feedbacy yb k system, c) First plasma operation has demonstrated good capability of HL-1 r plasmMfo a physics investigation.

ACKNOWLEDGEMENTS e authorTh s would lik thano t e e Theoretith k c Divisioy b d le n e assistancth r device fo th n n ei e Ga . designProfD . Diagnose Q . th , - tic Division led by Prof. G. F. Dong for measuring plasma parameters, many worker techniciand an s r machinfo s e operation.

72 RECENT DEVELOPMENT ADITYN SO A OPERATION XA9745744

P.K. ATREY, V. BALAKRISHNAN, S.B. BHATT, D. BORA, B.N. BUCH . CHAVDAC , , C.N. GUPTA, C.J. HANSALIA, K.K. JAIN, R. JHA, P.I. JOHN, P.K. KAW, A. KUMAR, V. KUMAR, S.K. MATTOO, C.V.S. RAO, H.A. PATHAK, H.R. PRABHAKARA, H.D. PUJARA, P. RANJAN, K. SATHYANARAYANA, Y.C. SAXENA, G.C. SETHIA, A. VARDARAJALU . VASP , U Institut Plasmr efo a Research, Bhat, Gandhinagar, India

Abstract

operatioe Th ADITYf no A tokama convertew k witne e hth r based power supplies for TF, OT and VF coils is described. We have obtained a maximum plasmf dischargA o k s 2 o9 m fa 2 currene10 duratiod an t separatn i e discharge expecd san progressiva t e improvement upt designee oth d discharge parameters. However reproducibilite th , discharge th f yo stils ei problemla . Extensive surface conditioning of the vessel wall and feedback control are planned to obtain good repeatability of the discharges.

I. INTRODUCTION

The ADITYA tokamak was made operational in September, 1989 [1]. The power supplies of the tokamak, at that time, consisted of capacitor banks for OH and VF coils and a DC power supply for the TF coils. Ini- tially, breakdown experiments were carriet toroidaa t dou l magnetic field Bt = 0.25 T in various discharges with different fill gas pressure, vertical magnetic field, loop voltages and current rise rate [2-5]. Later, physics experiments were carriestudo t nature t yth dou turbulenf eo t particle fluxes and turbulence in the edge plasma and interesting results have been reported [6-15] dischargese Th . thesn i , e experiments, wer smalf eo l duration (20-40 ms ), plasma current ( 20 - 30 kA ), average electron density = ( 3 - 6 ) xlO18 m~3 and electron temperature » 100 eV. However, the ADITYA tokamak (#0 = 0.75 m and a = 0.25 m) is designed for much higher discharge parameters: plasma current Ijj — 250 kA, electron temperature Te = 500 eV, chord 19 3 averaged electron density ne — 2 xlO m~ , toroidal magnetic field Bt = 1.5 T and discharge duration = 300 ms. The power system which will enable us to obtain these parameters are based on a combination of transformers, converters, wave shaping circuitary and a sophisticated control system.

This paper describe convertee r effortus ou s o t s r based power supplies and obtain discharge highef so r plasma curren longed an t r duration.

II. POWER SUPPLIES

e poweTh r supplies 132/1consiso tw A substationf V o 1tk MV 5 7 r sfo (peakA MV ) 0 transformer(peak5 d an ) pulsed an sconverte C dD r system with wave shaping and control systems. The incoming 132 kV is steped (peakA MV ) 0 transforme5 y b V k 1 rdow1 whic o nt s originall hwa - yde

73 (a - ) DTPS Control Scheme

Jpl(o«)

I I— °—1 pl(o£i) *-(TS) Plasmr>0 a ^^ /"v I

(b - )VFP S Control Scheme

Alf

18mH 37mQ VF Coil

Figure 1: Control Schemes for (a) OTPS and (b) VFPS

signed for ADITYA. But after considering the upgradation and reliability one more bay of transformer (75 MVA peak or, 37.5 MVA continuous) is installed which can be operated alternatively. In both cases the peak power demand is compensated to 50 MVA by supplying 36 MVAR through LC power compensators line eTh . commutate puls2 bridged1 C eD usee sar o dt power TF, OT and VF coils.

poweF T e r Th supply (TFPS designes )i r fo provid o d) t (35 A V k 0 0 e5 1 s flat-top duration. Rise and fall times are two seconds each. The scheme consists of two 6-pulse thyrister converter bridges paralleled through inter- phase transformer (IPT) for cancelling 6th harmonics. Current sharing of bridge alse sar o controlle differentiay db l current control fro sideC mA .

poweT O e r supplTh y (OTPS designes )i magnetiz A provido dk t 0 2 e± - curren g coils T in OTPe O Th o . t t S consist 6-pulso tw f seo bridges connected eacA k h0 (Fig-2 d . an la)seriee A k Th .0 s serien converter+2 i r sfo e ar s

74 fed from converter transformers from 11 kV bus. The magnetizing current is raised to a flat-top value in about 1 s and then commutated through a selected resisto develoo t r looe pth p voltage require breakdowns ga r dfo . looe Th p voltag furthes ei r controlle switchiny db g resiste t preprogramera d magnetizinA k 0 timings+2 e gTh . current provide voltsecond6 s0. s (Vsr )fo the plasma current to rise to 250 kA. Additional voltseconds required for maintaining flat-top plasma curren provides i t constana y db t dl/dt controlle kA/s8 d(5 ) zero- crossin negativd gan e converter of-2 (0.A 0k 6 Vs). poweF V e r supplTh y (VFPS designes )i F provido dt V e th e n 12.i A 5k coils for 600 ms duration (Fig. lb). In order to obtain fast response of VF current during initial rise phase of plasma current, a bias is given which avoide precharge V e deacurrentF k th s V 4 d e banA th . dn di capacitor initially raises the VF current to 600 A in 2.5 ms ( Fig. lb ).

Both OTPS and VFPS can be operated in closed loop feedback mode with actual plasma curren lood an pt voltage feedbace Th . k loo shows pi n in figures l(a) and l(b). The OTPS uses voltage control (Upt) till the re- e circuisistanc th plasmd n i an t s ei a current (Ipi) whe e resistancnth s ei bypassed. The power supply can provide 3 volts of loop voltage in negative phase operation of the converter. The VFPS uses plasma current (/j) and

plasma position for the closed loop feedback control (Fig. lb). A finp e position control can be provided by a seperate power supply (position feedback power supply) and feedback coils.

We have obtained discharges in ADITYA by operating power supplies coilT O provido s t n i A coils F maximuek a T curren 0 A e 1 ,k th 0 3 n i tmf o and 6 kA in VF coils.

III. RESULTS AND DISCUSSION

Since there exists large magnetizing currencoilT O s n evei t n before eth generation of loop voltage, the consequent stray magnetic field poses problems for initiation of discharge and rise of plasma current. For ADITYA, the measured stray field (vertical component) was found to be large (4 G/kA current)T oO f installee W . additionan da l pai coilf connected o r san n i t di series wit othee hcoilth T O r s which reduce strae sth G/ky7 fiel0. Ao f t do currentT O . Without these coils breakdows delae ga th , n yi founs nwa o dt be 5 - 8 ms which was reduced to 1 - 2.5 ms when the correction coil was brought into the OT circuit. Further, we find that plasma current and discharge duration are less than 20 kA and 10 ms respectively when the VFPS is operated without the 4 kV precharged capacitor (Fig. lb). The initial biasing of VFPS (-8 ms to +12 ms, IvF(ref) = 5 kA) is also necessary to make fast initial rise of VF current and the resultant successful rise of plasma current.

diagnostie Th c trace ADITYn a f so A discharg showe ear e Fign Th i . 2 . loop voltag shows ei Fign i . 2(a) correspondine .Th g plasma currend an t vertical field are shown in Fig. 2(b). The plasma density is measured using a 100 GHz //- wave interferometer (Fig. 2c). It rises sharply (5 x 1018m~3 initiae th n i l ) phasms e5 whe3- d plasme n an i nth A k a 0 curren1 < s i t later, it increases upto « 1019m~3 for high plasma current. The measured H signal (Fig. 2a) shows that the ionization is nearly complete before 7

. Initiams a l ris plasmf eo a densitexplainee b prefile n th yca y ldb hydrogen s pressurga e (7.5 xlO~5 torr ionization)% wit0 h10 .

75 SHOT 3968 A Y T I D A 16 NOV 95 12:17PM

-4 0.00 0.01 0.02 0.030.04 0.00 0.01 0.02 0.03 0.04 10

-5

-10 0.00 0.01 0.02 0.030.04 0.00 0.01 0.02 0.03 0.04 TIME (SEC) TIME (SEC)

Figur : Diagnostice2 s Trace f Shoo s t 3968 ) loo(a :p H voltag d an i U e a

signal, (b) plasma current Ipi and appliepd vertical field By signals, (c) chord averaged plasm ) horizonata(d a d densit an d vertica e an n yl l positiof no plasma curren trespectively y centroid8 d an x S , .

100 SHOT 4242 A D I T Y A 5 JAN 96 12:31PM (b) -I 6

,- 4

0.00 0.02 0.04 0.060.08 0.00 0.02 0.04 0.06 0.08

(d) 4

2 Ha(orb units) 0 O-ffl (arb. unils) -2

-4

0.00 0.02 0.04 0.060.08 0.00 0.02 0.04 0.06 0.08 TIME (SEC) TIME (SEC)

Figure 3: Diagnostics Traces of Shot 4242: (a) loop voltage Upi, plasma current Ipi and vertical field By (b) chord averaged plasma density n and

Soft X-ray signa ) horizonta(c l d verticaan l l position f plasmo s a current e I signalsII - 0 .d an respectivelyy a centroid6 H d ) an (d x d 6 , an ,

76 Spectrometers and narrow band filters are used to monitor optical radia- tions from hydrogen (regularly), carbon and oxygen (in separate discharges). Oxygen, which is detectable as 0 I from the beginning of the discharge ap- mose th pearte b abundan o st t impurity spectrae d Th .an I lII peakO f so 0 II follow that of the 0 I signal at intervals of about 2 ms successively. This is indicative of temperature rising to 60 eV within the first 10 ms. The subsequent evolution of a discharge shows considerable variation from shot to shot. In many shots, a high level of hard X-ray is detected accompanied by increas impuritn ei y line signals.

plasme Th a current initially rise t abousa MA/5 t 0 1 s reachint a A k 0 g4 ms. Subsequent rise of plasma current is slow (2 MA/s). All our discharges terminate in the rising phase itself either because of density/ q- limits or because of loss of equilibrium. The maximum values of plasma current and density obtained so far are 80 kA and 1019m~3 respectively. The plasma position is well maintained in such high current discharges (Fig. 2d) and they may have terminated because of density or q- limit. Although mirnov signals wer t monitoreeno discharge th n di e presente Fign di , the2 . y were recorded in many similar discharges. The mirnov oscillations ( 10 kHz ) show large increase in the magnitude prior to termination of the high current ( « 80 kA ) discharges. Other discharges terminate probably because of loss of equilibrium. This is because the plasma current is unable to follow the applied vertical field (Fig. 2b). We expect to achieve longer duration of plasma curren keepiny b t plasme gth a current (/p/) belo disruptioe wth n threshold. This can be done by reducing the vertical field (By) either by preprogrammin feedbacy b r go k contro f VFPo l S wit e actuahth (se& ! l e Fig. Ib).

Note added after TCM:

Subsequent to the presentation of this paper, we operated ADITYA by reducing the By through preprogramming and appropriately adjusting Upi e ris f plasmeo Th (Fig. a3) . currendeterminew no e verticas i tth y db l field. The plasma current is held by the vertical field upto 60 ms when loop voltage reduces to 1.5 volt. The lower loop voltage is probably insufficient maintaio t plasme nth a current, leadin dischargr fo dl/dto g0 t < e duration exceeding 60 ms. We expect to increase the plasma current as well as the discharge dura- designee tioth o nt dimprove) value(i : y sb d surface conditionin bakiny gb g and discharge cleaning, (ii) bringing negative converter in the OTPS for long duration discharge, (iii) feedback control of VFPS and OTPS by measured loop voltage, plasma curren positiond an t .

REFERENCES

. BhattB . 1. . BucBoraS .N D , t . al.,he B , India . PurnJ e Appl. Phys. 27, 710 (1992). . Atrey. Bhatt.K B . , 2India. BorP S .,ai D ,t e a. Phys nJ . 66B3 47 , (1992). . Atrey. Bhatt.K B . 3. P BorS ., D , t al.,ae India . PhysnJ . 66B1 48 , (1992). 4. P .K. Atrey, S. B. Bhatt, D. Bora et ai, Indian J. Phys. 66B, 489 (1992).

77 . Atrey. BhattB .K . 5. P S Bor,. D , t al.,ae India . PhysnJ . 66B9 49 , (1992). . . Matto. KawJha6K K R . . . S P , , t al.,oe Phys. Rev. Lett , 13769 . 5 (1992). . . . MattoKawJha7R K K . . . S ,,P t al.,oe Nucl. Fusio , 120n33 1 (1993). Matto. Kaw. . JhaK 8 K R . . . S , P , t al., oe Proc. 14th International Conf. Plasma Phys. Control. Nucl. Fusion, Wurzburg, Germany, Sept. 30 - Oct , 19927 . . (IAEA 1993) Vol. 467p , .1 . . Joseph. Jha9K R .. ,. B Kalr, ProcR ,al t ae . 15th International Conf. Plasma Phys. Control. Nucl. Fusion, Seville, Spain, Sept- Oct 6 2 .. 1, 1994. (IAEA 1995) IAEA-CN-60/A4-II-4. 10. G.C. Sethia . ChennD , a Redd ADITYd yan A Team, Proc ICPPf o . , Brazil, Oct - Nov 1 .3 , 19944 . . 37-40pp , . . G.C11 . Sethia . Chenn,D a Reddy , Procal t ,e .P.K ow f ICPPKa . , Brazil, Oct. 31 - Nov. 4, 1994, pp. 151-154. 12. G.C. . ChennSethiD d aan a Reddy, Phys. Plasma , 198s2 9 (1995). . G.C13 . . ChennSethiD d aan a Reddy, Phys. Rev. Lett xxx6 7 . x (1996). 14. JhaR . , P.K. Kaw, S.K. Matto , Physal t oe . Rev. Lett xxx6 7 . x (1996). Y.Cd an a . SaxenaJh 15. R . , "Wavelet analysi ADITYf so A edge turbulence: Evidence of nonlinear interaction", Institute for Plasma Re- search, Report IPR/RR-145/96.

78 XA9745745 IMPROVED CONFINEMEN CURREND TAN T DRIVE BY BIASED ELECTRODE IN VERY LOVf-qa DISCHARGES OF SINP TOKAMAK

. GHOSHJ . CHATTOPADHYAYP , . PALR , , A.N.S. IYENGAR Saha Institut Nucleaf eo r Physics, Calcutta, India

Abstract

Radial electric fielinduces edge dwa th en d i regio SINf no P Toka- ma meany k b biase a f so d electrod influencs it d studies ean ewa n di

vere th y Iow-g0 (VLQ) plasma discharges . Wit(qed2) h< negagea - tive bias applie improvemenn da energe th f o ty confinement timy eb observeds wa 5 factoa 1. .f o rInterestingl plasme yth a currens wa t seen to be sustained for longer time with the bias, whereas the loop voltage dropped suggesting a current drive mechanism at work. Hy- drogen line intensity decreased. Large currents were drawn to the electrode ( much above the ion saturation current ) which suggest establishment of a coaxial capacitor. The measured perpendiculer conductivity agrees withicalculatee th o nfactot a 2 f do r conductivity oleaka f y capacitor.

1. INTRODUCTION:

Since the discovery of high confinement (H-mode) discharges in the AS- DEX Tokama higy b ] h k[1 powe r neutral beam injection higo ,t thihw slo confinement mode transition (L-H transition) phenomenon has received both theoretica experimentad an l l attentio fusioe th n ni communitye Th . H-mode has since been observed in many tokamaks employing different heating techniques with diverte s wel a rs limite a l r configurationse Th . transition int H-mode oth characterises ei d mainl) sudde(a y yb n increase poloidae inth l plasma rotation improvemen) (b , energe th particl d n i ty an e

confinement time, (c) sudden decrease in HQ(Da) light emission indicating reduction in recycling and (d) gradual increase in plasma density. In the Continuous Current Tokamak (CCT) experiment, Taylor et. al. were able to induce H-mode like ohmic discharges by introducing a negatively biased electrode a few centimeters into the plasma. The CCT results demon-

strated that the radial electric field, ET, plays an important role at the L — H transition. Since then, many tokamaks [2-4] have obsserved this kin transitiof do n introducin electrodn ga e intplasmae oth .

All these biasing experements, wher substantiaea l radial currens ha t been drawn e donar , norman ei l Tokamaks having e ((ledge,2 qedgeth s > i safety factor at the edge). The present electrode biasing experiment has

bee Vere n th carrie yn i Low-q t dou aSINP-Tokama f dischargeo 2 < e sq^ k [6orden i ] studo rt influence yth radiaf eo l plasmelectriQ VL ae c th fiel n do properties.

79 e presenTh t paper discribe r finding s organisei ou s d an s s followsda . In section 2, the experimental set-up and the operational conditions are described. Section 3, discusses the experimental observations with the ap- plicatio f biad describeno an s e effec th sf highe o t r bias voltagee th n o s plasma theoreticaA . l mode calculatinr fo l g plasma radial conductivits yi presented in section 4, and conclusions are presented in section 5.

2. EXPERIMENTAL SETUP:

e SINTh P Tokama smala s ki l iron core Tokamak wit hmajoa - ra r dius of 30 cm and minor (plasma or limiter) radius of 7.5 cm. Toka- mak operational parameters during biasing experiment (toroidaT B e ar s l

magnetic field)~ 0.45 T, Ip (plasma current) ~ 26 kAmp, ne (average 19 3 density) 10x ~2 e (averag m~T d ,an e electron temperature) less than 100 eV. Filling gas is pure hydrogen.

Powersupply

Insulation Sleeve

Electrode tip

Vaccum Vessel

Limiter Radius

FIGURE 1 : Schematic drawing of the polarization set-up.

setue Th p use edgr dfo e polarisatio shows ni fig.(l)n ni insulatedn A . , cylindrical tungsten electrode of 6 mm diameter is inserted vertically from electrodporte p exposee th to variee f Th b .e o n th p ddeca ti lengte th f ho from 3 mm to 20 mm as well as its position inside the plasma. For the presen insidi.em m c limiterc e .t2 positionee5 th s stud 5. wa = .t y i r t da capacitoF m 0 A1 r ban kswitcR wit usevariabls a SC hs a hdwa a e power sourc volt0 e( -50o t electrodee 0 th volt) r fo .

80 3. EXPERIMENTAL RESULTS AND DISCUSSIONS:

The electrode size and the exposed length have no noticeable effect on the plasma parameter biasedt no s s whe. wa Without ni bia e plasme tth s th a current begins to fall gradually soon after it reaches its peak in about 1 msec fig.(2a) constanA . t negative voltag bees eha n applie t thida s instante Th . bias voltage is applied between the electrode and the limiter and is varried -35o t fro 00 mshoa -5 volt shon o t to t basis. Fig.(3) show variatioe sth n of electrode current with time for different biasing voltages.

Whe biae nth s voltag varies ei d-35o t fro 00 shoma volt-5 shon o t to t basis, a sudden increase in electrode current much above the ion saturation curren bees ha tn observed afte biasine th r g voltage of-100 volt. Fig.(4) shows the characteristic I — V curve for two different values of electrode exposed length, namely, 3 mm and 10 mm. After a threshold value of biasing voltage (-100 volt) the current drawn through the electrode remains almost the same for these two cases.

FIGURE 2(a): Time evolution of plasma current for different biasing voltages.

80 70 o ^ 60

81 600 sa,

400 -

JH J-i

200 h o

1.0 1.5 2.0 Time (msec)

FIGURE 3 : Variation of electrode current with time for different biasing voltages.

1000

800

600 - 10 mm, o 400 3 mm I 20° 0 0 100 200 300 400 500 bias voltage (volt)

FIGURE 4 : Variation of electrode current with biasing voltage for two different values of electrode exposed length.

Fig.(2a) compares the time evolution of plasma current Ip for bias voltage of zero volt and -150 volt. Fig.(2b) shows the corresponding change in loop voltage. With the application of -150 volt to the electrode the plasma current is sustained for longer time whereas the loop voltage goes down same th t ea time. This dro loon pi p voltage canno explinee b te th y db amount of drop caused by drawing excess current from the Joule Heating capacitor bank. It suggests a current drive mechanism. A possible scenerio for drivin likcurrene e b geth y thisma t : whe nbiaa applies si radiada l electric field is set up which gives energy to the current carrying electrons in directioe th n perpendicula toroidae th o rt l direction collisios A . n frequency decreases with velocity increasing these electrons carry curren longea r fo t r time. This curren te edg flowth en i sregiotokama e th f no k flattenine th g

82 150

100 0 volt cd

50 L

- bia s° volvoltag15 t e

5 2. 0 2. 5 1.

FIGURE 5(a): Time evolution of HQ emission for different biasing voltages

1.5

1 0 150 volt

0.5

bias voltage 0.0 1.5 2.0 2.5

FIGURE 5(b). 13P vs time.

150

g 100 150 volt

50 . bias voltage 0 1.5 2.0 2.5 Time (msec) FIGURE 5(c) Energy confinement tim tims ev e

83 30 150 volt d 0) 20 .5^- cc< 2 6"^ ctf 10 CL

0 0.5 1.0 1.5 2.0 2.5 FIGURE 6(a): Time evolutio f plasmno a curren r biasinfo t g voltagef o s -150 volt and -250 volt.

co O S,-j S 0 B w C

bias voltage -10 j____i_____i 0.5 1.0 1.5 2.0 2.5 Time (ms) FIGURE 6(b): Time evolution of plasma position for different biasing voltages.

current profile internas A . l inductence decrease thin si s situatione th s i o s , loop voltage.

As soon as the bias voltage is applied to the electrode the HQ signal goes down with respect to its level with zero bias. Fig.(5a). A recycling peak whic observes hi d wit zere hth o volt bia delayes si substaintiaa r dfo l amount of time on application of bias votage. This delay time gets longer bias a s voltage increases upto some threshold valu -10= e( 0 volt).

84 The poloidal beta is measured using a diamagnetic loop surrounding the plasma columnbiae th sr voltagFo . -15f eo 0 voltf3e maintaines pth i t da longea r aboufo r7 tim0. t e compare correspondine th o dt g discharge with zero bias. Fig.(5b).Energy confinement tim calculates ei d usins i g d f3pan plotted in flg.(5c). The energy confinement time improves by a factor of more than 1.5 with the application of-150 volt to the biasing electrode.

All these observations suggest tha plasme th t a confinemen improves i t d wit applicatioe hth -15f no 0electrode e voltth o t .

However, increasin e negativgth e bias voltage further, i.e. beyon- d 200 volt, plasme th a loose stabilits sit observes ya d fro positioe mth n coil, and all the observation presented above are reversed. Fig.(6a) shows the plasma current IP plotted for the biasing voltages of zero and -250 volt. A sbiae sooth ss nvoltaga -25f eo 0 volt applieds i plasme th , a currenP I t falls off very sharply to a Tninimimn value (400 amp) and sustaines for quite some time compared to the case with zero volt bias where the plasma current falls graduall zeroo yt . Fig.(6b) show plasme sth a position measured using cos9 coils. Wit applicatioe hth -25f no 0electrodee voltth o t , plasma seems to move outwards quicker than in the case with the zero volt.

This loss of stability may be initiated due to the disturbances caused by substantial local magnetic field of large current drawn through the electrod (magneticallp eti y unshielded) wit highee hth r bias voltage factn I . , calculatee th d local magnetic perturbation goes much more th ef o tha% n5 poloidal field in this case.

4. PLASMA RADIAL CONDUCTIVITY:

When the negative bias to the electrode is more than about -100 volt the probe current drawn rises much above the ion saturation current to the probe. This is apparently due to the scenario that applying a negative bias electrode th o t e result chargine th magneti e n si th f go c surface intercepted electrode byth e tip. Henc t formei potentiaa s l layer currene Th . t drawn electrodee toth , now, doe t depenelectrode no sth n do e area, whereat i s depend average aree th th f ao n seo magnetic surfac electrie th n ei c layed ran the perpendicular conductivity. This resembles the plasma co-axial capacitor wit interceptee hth d magnetic surface th eplats it formind f eo an e gon last closed surfaclimitee th t ea r formin othere plasme gth Th . a potential jrA,= e I wherd an eL T E prob e a= th t p givee eV b locationy n nb ca w no , jT is the perpendicular current density and A is the area of the average magnetic surface in the electric layer. The perpendicular current density j

can be calculated from the radial momentum equations in terms of radiar l

electric field Er and the effective collision frequency v. Which comes out as [5] . nmv

Jr = ~~-Er [1J

85 This gives the perpendicular conductivity as nmv °r = —#r M ° Experimental data shows that the conductivity calculated using Eq.(1] agrees within a factor of two to the measured conductivity.

5. CONCLUSION:

Typical increase in the duration of the plasma current has been observed in very low-g (VLQ) discharges of SINP Tokamak, by inducing a

radial electric field ina the plasma edge with biased electrode. The decrease in loop voltage combining wit e sustainmenhth f plasmo t a currena r fo t longer time indicate currensa t drive mechanism. Energy confinement time

increases ha factoa a y signamordf b H o r d el an thagoe 5 n1. s down with bias indicating a better confinement and reduction in recycling. The plasma able sustaio st n bias voltage upto some limiting value after whic t hloosei s its stability. It seems with the application of bias the magnetic surface intercepted by the electrode tip gets charged and an electric layer is formed. The perpendicular conductivity calculated using radial momentum equa- goon i tio e dnb agreemencomeo t t sou t wit measuree hth d perpendicular conductivity.

References

1. WAGNER F., et. al., Phys. Rev. Lett. 49(1982) 1408 2. WEYNANTS R. R., et. al., Nucl. Fusion 32(1992)837 3. ASKTNAZI L.G., et. al., Nucl. Fusion 32(1992)271 4. WANG E.Y.,et. al., Nucl. Fusion 35(1995)467 5. WEYNANTS R.R., et. al., Controlled Fusion and Plasma Heating (Proc. 17th Eur. Conf. Amsterdem, 1990), vol. 1, European Physical society, general (1990) 287 6. IYENGAR A.N.S. . al.et , , Pramana-J.Phys. 39(1992)181

86 RECENT HL-1M RESULTS XA9745746' HL-1M TEAM (Presented by E.Y. Wang)

Institute for Plasma Research, Bhat, Gandhinagar, India

Abstract

The HL-1M tokamak (a modification of HL— 1) is a circular cross—section tokamak and has been in operation since OCT 1994. The HL-1M character istics are as follows:R=l. 02m, a = 0. 26m,BI = 2 — 2. 5T, Ip=100 —320kA and pulse duratio Isecn~ simplA . e chamber boronization techniqu employes ewa d in HL- 1M. A distinct poloidal asymmetry was observed in the respons of the edge density fluctuation ohmie th t csa discharges, biased pump limiter LHCd an , n Dco ditions. Biased pump limiter experiments have seen that biased pump limiter with respect to the walls allows one to control the edge and core plasmas and to mea sure total particle outfolw. Impurity transport using laser jection bloi f e wof n pr sent resulte sth ohmicn si , H—mode. LHCD experiments were performed with 400KW of 2. 45 GHZ wave. The results presented in this paper arc concentrated in two aspects. The first is the dependence of LHCD efficiency on power and the second is investigation of confinement improvements by LHCD.

1. INTRODUCTION HL-1a circula s i a ) modificatioM( 1 r cross—sectio— HL f o n n tokamak, 1

with R= 1.02m, a=0.26m, B, = 2 — 2. 5T, Ip= 100-320KA,n; = 7X I0 m~,T3 (0)^-' IKeV, Pulse duration up to I sec and two full poloidal graphite

limiterc . o conduct e Maiar t nM hig1 objective - h HL powe f o s r axiliary heating (NBI, ICRH ECRHd an , currend an ) t drive (LHCD develoo t d an )p physics and technology bases for the HL—2A tokamak in SWIP. Summary of HL— 1, HL- 1M and HL—2A is shown in Table 1.

2. CHAMBER BORONIZATION A simple chamber boronization technique was employed in HL-1M^. The most radical effect f boronizatioo s HL-In ni M havtime6 — e e sr 3 prove e b o nt duction in total plasma radiated power and a decrease of the loop voltage from 1.5—2V down to 1 — 1. 5V. Spectroscopic survey of radiation in the vuv region displa totae yth l diminishin spectrae th l al lf go line f high—so Z impursity atoms. suppressioe Th oxygef no carbod nan n lineobserveds swa intensite Th . f radiyo a tion in the soft X—ray region has been also drastically reduced. With boroniza tioimprovee nth d energy confinement time observe HL-IMn di t coulm I i . e db proved by 30—40%.

87 HL—2d an M I A - parametersTablHL , . 1 SummarI e — HL . f yo

HL-1M HL— 2A parameter HL— 1 (designed value) (designed value) R(m) 1.02 1.02 1.64 a(m) 0.20 0.26 0.50 k 1.6-1.8 Ip(KA) 225 350 800-1000

Bt(t) <3 <3 3.0-3.5 F(VS) 1.7 1.7 9.0

t i ICRH(MV) L

i NBI(MW) 1 ~3-5 LHCD(MW) 0.25 i ~3-5

i ECRH(MW) 0.25 <0. 5 1

780 520

670

< 560 445

450 408 -

340 370 198 0 f*z

149 - -47 -

>J 100 - -94 -

0 40 0 30 0 20 0 10 0 50 0 40 0 30 100 0 20 500 Time ( ms) Time ( ms) Fig (a)1 . . During positive bias II., increas Fig (b).1 . During negative bia , dropsH s . Afte seriea r f dischargesso possibls i t i , e through erosio f limiteno redd ran e positio borone th f no , thaboroe th t n wall coatings fro previoue mth s discharges continue reduc o drecyclint H influe e th e th f impuritie d xo gan s fro wallse mth . The chamber coated with boron—containing carbon film remained in the HL-1M tokamak to be expossed to more than 100 pulses. Comparsion with the initial boronized film reveals the decrease in B/C radion and finally the formation of a carbon film may indicate to increase H recgcling. In that case, helium glow dis charge conditionin f HL-1go M routinels i y use reduco dt e hydrogen recycling1^.

. BIASE3 D PUMP LIMITER a)The edg cord ean e plasma contro biasey b l d pump lirniter A density rise is observed with either polarity of bias potential. The density rise during positive biasing seems to result an increase in the fuelling of plasma du enchanceo et d recycling fro wale mth s show a a increases l H Fign ) ni (a .1 . The density rise during negative biasing seems to result from improvement of par tide confinement tim shows , dropsea H Fig) n ni .(b .Effect1 f limteo s r pumping on the plasma performance are shown that the bolometer radiation is lower during pump limiter slot open comparing with slot close.

200

150

100

s 0 CD -50

-100

-150 -500 -400 40 -30 0 30 0 0 -2020 0 10 -10 0 0 Has Vottage (V) characteristiV — I e biaa Th Fig f . cso 2 . pump limiter.

Table 2. Particle confinement times from different methods. method a(m) Tp(ms) biased limiter 0.23 20 langmuir probe 0.26 30 laser blow off 0.26 25

89 b)Thc total particle outflow measurements with a positive biased limitcr characteristiFigV show2 . I— e nth biasea f co d pump limitcr insertem c 3 d beyond(i. c. towards the plasma core) the full poloidal grathitc into a hydroge dis charge. Note that the current increase with Vb for Vb^150V, then saturated. This saturation curren s measura usei t s da f totaeo l particle outflow froe mth 5 tokamak ^. The positive biased limiter current Ib (electron collection by the lim iter) equals to the total ion saturation current to all the boundaries of the toka

mak. Also we can use Ib as total outflow to estimate particle confinement time TP. T, = h/e (I) where 0= is the plasma density, a and R is the minor and major radius respective 19 3 ly. The comparison of Eg. (1) with Ib=150A, nc = 2X 10 m~ , a = 0. 23m . 02m1 an= ,R d prob e datconfinemend aan t time obtained with laser blof wof technique is given in Table 2.

4. LOWER HYBRID CURRENT DRIVE a)increase power

In LHCD experiments, the current drive efficiency r|=nJpR/PLH is in the . rangI—00 f o e . 4(l020m~2A/W); this indicates tha a higt h power muse b t use obtaio dt nhiga h plasma current recene th n I t. experiment LHCM 1 - D HL f so power was increased from less than 200kW in HL — lw up to more than

400kW. Comparision of LHCD in (Plh, n) diagram for HL—1 and HL-1M is show Fign ni . .3

U.O o HL-1 0.7 • A HL-1M

0.6 - => 0 •

0.4 ; o '~o A

^ * ^ 0.f O 3 ^ ^ O O O i Efficienc y 0.2

0.1 o :

n 0 40 0 35 0 30 0 25 0 20 0 15 0 10 0 5 0 Power (kW) ~l%- 3. Mearured LHCD coefficient versus power.

90 0 Br-2.ZT ,JI Ip -96KA • . pm-soKA o O "H

°° fo 8 3 O a ao o

''123

3 ( 1019m-e N ) e T^-LHCD/'P-OTh Fig . 4 . functioa s Ha f densityno .

confinemen) b t improvement Introducing in LH wave, particle confinement is signficatly improved. The

characteristics of TP_LHCD/TP_OH as a function of density are shown in Fig. 4 , here Tp-LHci>and TP-OH are the particle confinement time in LHCD and OH conditions re spectively observee Th . d decreas almosf ef impurito o l al t y regione lineth n si f so visible and vuv during LHCD reveals that the density increase is caused neither by recycling from the walls nor by impurity injection but by improved particle confinement. A reduction of density and potnetial fluctuations duing LHCD was observed. Electrostatic turbulence in the edge plasma in relation with confine ment during LHCD was studied and it was shown that the turbulence is stabilized LHCy b D leadin improveo gt d confinement.

. MEASUREMENT5 EDGF SO E FLUCTUATION FLOD SWAN VELOCITY Since it was studied in early 1970' s^, the instability driven by the cross — field gradient (shear) of the plasma mass flow velocity paralled to the magnetic field in an inhomogeneous plasma has been investigated extensively. Recently, the L — mode to H—-mod e (L— H) transition in tokamak plasma confinement s founrelate e presence b wa th poloidae o dt o th dt f eo l flow shear nea plasme th r a

Experimental measurement parallee th f poloidad so an l l flow carriee sar t dou on HL- 1 M with a Mach probe. The measured Mach number , ion saturation cur rent fluctuations , and poloidal velocity profiles for ohmic , LHCD conditions are shown in Fig. 5. The results present evidence that the parallel velocity changes at about the same radial location as the poloidal shear flow layer. Such flow shear may suppress the plasma density fluctuations and affect the local plasma confine ment[7]. A distinct poloidal asymmetry was observed in the response of the density fluctuations at the ohmic and LH wave injection on HL- 1 M. The fluctuations

were measure t usinfouy a db a r gH poloida l (9=0°, 90°, 180°= , r 270° d an ) 25cm positions as shown in Fig. 6. During ohmic discharges on the out boardmid plane the fluctuations decrease significantly at the Improved confinement by LHCD, while those on the inboard generally show little or no reduction.

91 O Olunic • LHCD i

0.8

0.6 O

0.4

0.2

0 3.0-4

0.03

— U.C2

0.01

•0.6

o Q. O >

O . O

22 22 24 25 26 27 28

e radiaTh Fig . l 5 .profile f mearurco s d Mach number saturation io , n current fluctuations, shear of the flow velocity paralled to magnetic field, and poloidal velocity for Ohinic and LHCD.

92 300 150

0 LHCD 300 KCZZEZSZ22X, 150

0 300 150 inboard midplane (90) (0 X 0

outboard midplane (270 °)

400 450 500 550 600 Time (ms)

Fig. 6. H, fluctuations at four noIoidal(O=0°,900,180°,and 270°) and r=25cm(a = 26cm) positions, during LHCD on outboard midplane (270°) fluctuations decrease.

6. IMPURITY TRANSPORT Small quantitie f high—so Z impurtics (Al ) havFe , e been injected inte oth HL- 1M plasma by laser blow—off to study impurity transport. The progression of the impurities into the plasma was followed with temporal resolution using vuv spectrum and was simulated using an impurty transport code yielding values for the transpor dischargo tTw coefficien. V ed regimesan D t , Ohmi biased can dH —mode plasma have been compare focussiny db analysie represeo th tw n go f so n tative discharges resulte Th . s clearly show r botfo ,h cases, that impurity trans por smalles i tH—mode th n ri e discharg eohmie thath n nci discharge valuee Th . s transpore oth f t coefficient transpore th n i s t simulatio givee nar Tabln ni . e3

Tabl . duffusioe3 n coefficients Ohmic H — mode

2 D(m /s) 1.3 0.8 at the region for q>l

93 REFERENCES

1. O.I. Buzhinsky et al. , J. Nucl. Mater 191 — 194, 1413(1991). 2. G. L. Jachson et al. , Nuclear Fusion , Vol. 30, No. 11, 2305(1990). . Diebol, RevA . . al D .t d. e Sei 3 . Instrum. 66(1), 434(1995). . , DingNuclea. T Wan . . Y al . 4X .t E g,e r Fusio Plasmd nan a Physics, Vol. 14, No. 2, 12(1994)(in Chinese). 5. P.J. Catto, etal , Phys. , Fluids 16,1719(1973). 6. X. N. Su, etal. , Phys. Plasma 1, 1905(1994). 7. J. Q. Dong and W. Horton, Phys. Plasma 1(10), 3250(1994).

94 ICRF PLASMA PRODUCTION AND HEATING XA9745747 URAGAN-3E TH N I M TORSATRON

V.V. PLYUSNIN, V.E. MOISEENKO, A.I. LYSOIVAN, S.V. KASILOV, A.I. ZHUKOV, N.I. NAZAROV, E.D. VOLKOV, K.N. STEPANOV, O.S. PAVLICHENKO Institute of Plasma Physics, Kharkov, Ukraine

Abstract

The results of experimental and theoretical studies of AlfVen wave production and heating of

plasm frequence th n ai y range cyclotro n beloio e wth n frequenc o>y( coo< presentee ci)ar d Several type f antenno s a have been studie r plasmdfo productioF R a d heatine an nth n i g ) T 3 1 < B , cm URAGAN-35 2 1 < , a . Mcm torsatro0 10 = R n( - a frame type antenna (FTA) conventionally used for plasma RF production and heating with the best

operational propertiemoderatd an w lo et s a plasm a densitie 5x10*< e n~ s( 2 cm"-'), - compact three-half-turn antenna (THTA) proposed for plasma heating and density ramp up (up to 3 3 3x10^ cm' ) after the low density target plasma ( n"e >4xlO^ cm' ) had been produced by FTA. - recently proposed antenna of combined type ( "crankshaft" ) , which has the best properties of both above mentioned antennae in the whole range of densities e excitatioTh e electromagnetith f o n c field URAGAN-3n i s M plasma FTAy b s . THTd Aan crankshaft antenna has been studied numerically using 1-D wave code To study the dynamics of RF plasma production in the URAGAN-3M torsatron the 0-dimensional code have been used The results of calculations showed better performance of crankshaft antenna compared with FTA and THTA in the whole rang plasmf eo a densities When using the THTA at the scenario with FTA as a target plasma sourse. the experiments 13 3 performed showed the possibility of dense plasma production ( ffe > 2xl0 cm' ) and heating, which had not been obtained earlier in the URAGAM-3M torsatron The shifted towards the plasma core power deposition profil THTf eo A resulte modification di plasmf no a density profil improvemend ean plasmn i t a confinemen e preliminarTh t y experiments with crankshaft antenn plasmn ao a production showed that this antenn ^ cm' 0 producn 1 ca a o t ) plasm URAGAN-3e dense p th th eu n ( ei a M withouy an t 3 additional sourse of target plasma and it can be used for subsequent AlfVen heating

INTRODUCTIO therw extendeno s e i o t p U Nd amoun f experimento t plasmn so a productioF R toroidan i l magnetic device range th f frequencien eso i s beloe th f wo ion cyclotron frequency. These experiments show that the production of plasmas

with moderate value densitf so y realize e (10^b ^ cm' n 10 2 ca 3< d cm' ) e usinn 3< g e conventionath l antennae, suc s frame-typha e antenna (FTA) typically user dfo plasma RF production and heating with the puporse to study plasma confinement propertie URAGAN-e th f o s URAGAN-3d 3an M torsatrons [1,2 r ]Nagoyo a type antennI II a use targer dfo t plasma build-u torsatroS CH n pi n [3].

95 Both type f antennao s e have significantly developed current-carrying surfaces in toroidal direction. These features of antennae allow to excite effectively in near-antenna regio e parallee magnetith nth o t l c field componenF R e th f o t electric field. Electrons accelerate y thib d s component perfor e neutramth s ga l ionization and provide plasma denstiy rise. After neutral gas burn-out the subsequent heating of produced plasma is carried out by the Alfven waves. The mechanisms of fast wave (FW) conversion into a strongly damped slow wave (SW) in the region of Alfven resonance layer and the Landau damping of resonantly excited Global Alfven eigenmode basie th thif se o ssar heating scenario. e typicaTh l plasma parameters achieve e URAGAN-3th n i d M experiments 12 3 13 following:nare eth e «(2-3)xl0 cm' e «200-30 ,100-20» T e «10j n T , , 0eV 0eV 3 cm" , Te « 40-60 eV, T; «60-100 eV in the wide range of confinement magnetic field

values 0.2 < coo/cocj < 0.9 for fixed RF generator frequency coo. These experiments were characterized by excitation of strong parametric instabilities of IB\V on plasma periphery resulting in turbulent heating of ions and electrons and formation of hot ion tails with energy up to several keV [4]. The attempts to produce plasma with 3 3 higher densit (l-l.S)xlO^> e n y( cm" increasiny poweF b R ) e r strongeo rth f go r neutral gas puffing were giving small or even negative effect. This happened due to deleterious effect of RF power depostion profile shift onto the plasma boundary with the increse of plasma density. This deterioration of FTA features at high plasma density force fin o t dn appropriat a d e concep f antenno t a desig r densfo n e plasma

heating at the (oo

=(a>02/c2)6j is realized on the different magnetic surfaces ( where n is longitundinal

mode e majonumberth s e i dielectrirth e R radiutoruss ,i th l f & o c,s tensor componen . Thi) t s heatin moss gi t efficient whe e usedmodee ar nth 1 7 , s« wit a hl^ wher plasms i ea a radius. This conditio s beenha n explicitly formulate Refn i d ] .[5 d followan s also from Ref.[6]. Note tha thin i t s cas e wideth e spectru f modmo e number s excitee antenni s th y b d a system t possibl.no This wa se when helica] [7 l multi-half-turd an ] antenna[8 n e have been used. Moreover, these antennae havea complicated design and could excite only single Alfven resonance. As a result their power depostion profil s verwa ey sensitiv e variatioth o t e n plasmi n a density resulting from AR condition. To avoid these disadvantages and to find an effective denswaa f yo e plasma productio valuee heatind th f (2-5)xlfllno an t so p gu 3 cm"3 the concep compacf o t t three-half-turn antenna (THTA bees )ha n formulated [9]. However, this antenna could work only A afteproduceFT r e targeth d t 3 plasma ( ne> 4x10*1 cm" ). The existance of plasma density lower limit for THTA is connected with a considerably suppressed excitation of long wavelength AR, denswhicw lo en hi plasma e responsiblpoweF sar R e rth inpur efo t onto electrons. In orde improvo rt featuree eth densitTHTf w o s lo An i y range while preserving it's

96 features for dense plasmas the antenna of combined type - "crankshaft" antenna [10bees ]ha n proposed recently. NUMERICAL MODELIN orden I o studt r G y antenna system parameters o optimizt d an e plasma build-up scenario e URAGAN-3th n i s M torsatron, self- consistent numerical modeling with two codes: 1-D RF code and 0-D transport code anothelinkeo t e beeds on rha n performed [11]. codD 1- e e solveTh boundary-valusa e proble Maxwelr mfo l equatione th n si mode f nonuniforo l m plasma cylinder with identical ends e plasmTh . a dielectric tensor taking into account collisional energy dissipation as well as electron Landau damping was used. Ion gyroradius was assumed to be zero. Therefore, the excitation, propagation and mutual conversion of the slow and fast waves are correctly describe frameworn di e modelth f ko , while kinetic ion-cyclotron waves (EBW neglectede ar ) antennae Th . modelee ear s externada l currents which meea t condition dJV j =0. The dynamics of plasma build-up at the stages of breakdown, neutral gas burn-out and plasma heating has been studied using a 0-D transport code. A set of transport equations included the equations of energy balance of electrons, ions and atom s wel a ss particl a l e balance e electroTh . n energy balance includeF R e th d power term, energy losses due to radiation and ionization, Coulomb energy exchange with iontranspord an s t losses describe energy db y life-time scalinr gfo helical devices (LHD-scaling). The ion energy balance included the RF power term, the energy exchange with electrons and atoms due to Coulomb and charge exchange collisions as well as transport losses. The atomic energy balance included energy exchange with ions and energy flux with sound velocity to the wall of vacuum chamber. The charged particle balance included ionization term and term responsibl r lossesfo e , whic s evaluatehwa d scalingD fro me dynamicLH Th . f o s system with a constant number of particles was considered, i.e. the 'hot' ion lost at the wall was replaced in the vacuum chamber volume by the 'cold' atom (assumption of complete recycling). The antenna size and scenarios of RF heating was optimized by means of 1-D code F parametee R Th . optimizatiof ro ratie th f antenn oo s nwa a loading resistance

to plasma average density (Rp|/ne) in the whole range of densities. This parameter is directly linked to the plasma production (ionization) rate. The average radius of power deposition ( alss o)wa considered. The 1-D code generates the dependence of power deposition to electron and ion components on plasma density. These dependencies were substituted into the 0- D transport code.

97 NUMERICAL AND EXPERIMENTAL RESULTS Fig.l demonsrates the computed valueratie th f antenn o f o s a loading resistanc absorbtioo t e edu y nb

electrons to plasma average density R.,]/n and dependance of average radius of e

power deposition vs plasma density. The Rp|/n" parameter for THTA is e significantly less than for FTA one in low density range that explains the reason why FTA worked effectively at relatively low dense plasmas while the use of THTA becomes impractica than i l t range. Fro othee mth r hand usine th , THTf go morAs i e preferabl r densefo e plasma case follows fro significantle mth y improved power deposition profile manifasting itsel reducen fi value> d

1 OE»6 1 OE»7 1 OE«8 1 OE-9 1 OE*10 1 06*11 1 OE*12 1 OE'13 1 OE»14 n (cm3}

R In ?• pi e 1 OE-12 —

OE-1 OE'1 6 OE-71 06-1 B OE»191 OE-101 OE«11 OE*121 OE*131 4 n (cm3)

Fig.1 dependense Th . f averageo e radiu f poweso r deposition (upper chartd )an

Rpl/ne parameter deposition (upper chart) and Rp|/ne parameter (lower chart) versus plasma density for double frame (70x23 cm) and THT (24x40 cm) antennas.

98 URAGAN-3M, #49468, B-0.45 T, Hydrogen, f1=f2=5.4 MHz pulsA .FT e (RF-1 THTd an ) A pulse (RF-2).

,1.E12 cm(-3)

t.ms Fig.2. Temporal evolutio f plasmno a density . THTA increased density of the target plasma produce FTAy db .

URAGAN-3M, #48492, B«0.45 T, Hydrogen f1-f2«5.4 MHz, The density threshold of THTA operation evaluated at 4E11 cm(-3).

,1.E12 cm(-3)

Fig.3. Temporal evolutio f plasmno a density (RF-1A THTd FT . an ) A (RF-2) pulses. Prf1, Prf2 are about 100 kW.

99 the it's efficiency at dense plasmas own by the shift of RF power deposition profile ont plasme oth a boundary experimene Th . t results confirme conclusione dth s made on the base of theoretical consideration and numerical modeling. In Fig.2 the temporal evolutio f plasmno a density valu s presentedi e , where THT pulsF R A e started whe e targeth n t plasm s preliminarlwa a y produce y FTAb d , Fig.3 demonstrate e existenc th densitsw lo f o ey threshol f THTo d A operation. This

_ 13 -3

20 30 40 t(ms)

Fig.4 URAGAN-3M, #57138. The temporal evolutions of some plasma parameters: plasma density Hbeta-signal, ion saturation current on Langmuir probe and Soft X-rays signal in the typical discharges d THTan whe AA workenFT d consecutively

100 O • frame antenna D - double hame A - chain antenna antenna R / n pl e

' """' '""" ' "'""' "'""1' '"" II I I ' ""'i^"""" ' '"»•! ] TT™1 lllmli ' """1 "'""I ' """'I ' I1IIIB| ' '"""I 1 OE-6 1 OE»7 1 OE«8 1 OE-9 1 OE-10 1 DE-MI I OE-12 1 0£«13 1 OE»14 10E»6 1 OE»7 10E*8 1 OE-9 106*10 1 0€«11 1 OE«12 1 OE»13 1 OE»14 n (crrr3) n (cm"3)

Fig.5 dependense .Th averagf eo e radiu f poweso r deposition (left chartd )an

Rpl/ne parameter (right chart) versus plasma density for for frame (2x40 cm), double fram chaied (4x4an n) (4x40 cm antennas) 0cm .

O & -crankshaftantenna D -THTantenna 10E-11 ~|

Ve * 1 OE-12 -=

10E*« 106*7 10E» C€»1 8 9 10E»101 OE«11 10E»1 OE»121 OE* 1 OE*1 3 OE»6 1 106*17OE»1 8 OE«1 1 9 4 OE-O0 1 OE»11 OE«121 OE»131 4 n (cm-3) n (cm'3)

Fig.6 dependence .Th averagf eo e radiu f poweso r deposition (left chartd )an

Rpl/ne parameter (right chart) versus plasma densit crankshafr yfo t antenna (24x4, 0cm crankshaft amplitud e2cm- ) circle sco- cj=0.8, triangle sco/cocj=0.6,.an antennT dTH t a

(o/cocj=0.6 (squares).

101 3 antenna couldn't increase plasma density when it's valu less sewa than 4x10*1 Cnr . Typically, (see Figs.2,3) the growth of plasma density during the FTA pulse was saturatin ordee levee th (5-7)xlOf th f r o o lt ga ^ cm'^. During THTA operatioe nth 3 substantial increase of ne up to 3x10^ cm" was observed at approximately the same RF power value. The transition to a qualitatevely new stage of plasma discharg s observeewa d during THTA pulse (RF-2pulsA FT s opposee a )e th o dt (RF-1) (Fig.4). The significant rise of plasma density during THTA pulse was observed mainly at the plasma core while the ion saturation current registered by the expernal Langmuir probes dropped sharply. The signal of the Ha line measured outsid e plasmth e a core decrease same th t eda time e measurementTh . f radiao s l distribution of plasma density showed that the density profile steepened near the periphery. Such a behaviour of plasma parameters has never been registered using the FTA at any level of RF power. Despite that thes experimente eth s demonstrate successfue dth l productiof no dense plasma, it should be mentioned that the realization of plasma RF production usin o antennagtw complicatedo e disadvantageeto th seeme f b o o t e s e On .th s i s necessit havo yt e several device portth antennar n si efo severad ean l separatF eR transmitters. To optimize the antenna capable to produce a dense plasma alone the numerical studies have been carried Fig. e outFig.d Th .5 an 6 sho comparative wth e

voltage—RF

1 - Central modulated strap (I = I rf) - sid -0.2= ) 3 , e 5 I I strap = I s( f r 3 rf 2 rf 4, 5 - longitudinal strap conductors Fig.7 Crankshaft antenna view

102 URAGAN-3M, 13.03.1995, #65070, B3 MHz= 8. 0.6 = f ,, 4 T Hydroge n Plasma RF build up by Crankshaft antenna

, 10**12 cm(-3) Irf.a.u. 500

- 400

20 t, ms

URAGAN-3M, 14.04.1995, #65108 B3 MHz= 8. 0.6 * f ,, 4 HydrogeT n Plasma RF build up by Crankshaft antenna

, 10**12 cm(-3) Irf.a.u. 10 500

400

300

200

100

20 40 t, ms

Fig.8

103 results of numerical studies on optimization of antenna which will be capable to perform the neutral gas ionization and to provide the plasma density rise up to (2- 13 3 5)xl0 (secm'A value eFT .Th Fig.lRf e eo pth i/n r normalls i )fo e y higher than required for plasma production in the low density region. So the addition of small- size frame type currents to THTA should make it capable to perform neutral gas ionization. This addition is desirable not to deteriorate THTA properties at high

densities. The behaviour of Rpi/ne parameter vs plasma density for the frame antennae with small parallel length (4cm) show obvioue sth s minimu ^ -1010 t m1a 1 cm'3. This minimum is a result of poor excitation of long wavelength electromagnetic modes, which makes worse power deposition profil t a hige h densities e situatio e improveTh b . e shif n f th longitundinao ca tny b d l currents poloidal spectru e higheth o mt r wavenumbers. This allow o enchanct s e th e excitation of modes absorbed deeper in plasma column and to suppress the excitatio f modeno s absorbe plasme th t da a periphery realizee .b Thi n sca d usinga chain of frames stretched in poloidal direction (Fig.5). Obviously it is more preferabl n additio a e THTA s th a e o t n . Therefore t followi s a , s from Fig.6e th , proposed antenna of a crankshaft type is modified variant of THTA by the combinatio f THTfram4 no d Aean antennae placesentrae th t da l stra f THTApo . New antenna's properties at low dense plasmas are similar to FTA ones. At high

densit e Rth py|/n e paramete f thio r o differens to antenn t no ts i a fro e sammth e parameter of THTA.

URAGAN-3M, #65070 and 0-D modeling with radiation losses (effective impurity concentration 25%).

,cm(-3) 1.UC»T» - O O o O O O ^ -^u —iy 1

1.0E+12 s ^__^^^""^

1.0E+11 | Q oj/^

1.0E+10 I / 1.0E+09 1 F 1 OE+O 8t —— calc ex

i nc*.n-7 F i i , i -.,l ... i . I 0.005 0.01 0.015 0.00.0225 t,s Fig.9 The qualitative comparison of calculate experimentad dan l plasma density values evolutions.

104 The crankshaft type antenna (Fig.7) has been manufactured and installed in URAGAN-3M torsatron. The first pulses showed that this antenna can produce plasma alone. Fig.8 displays plasma density and antenna RF current evolutions in one of the first experiments on plasma build-up. In this regime

, ®o/eokG 5 cj6. =0.8 = o , hydrogen(B , withou s puffing)ga t , plasma densito t p yu 10*3 cm~3 was achieved. The decay of plasma density during the RF pulse is illustrating the common situation for the first shots characterized by strong impurity release from poorly conditioned inner metalic surfaces. The influence of e impuritieth dynamice th n o s f plasmo s a build-up were modeled numerically with the help of 0-D code by inclusion of additional radiation losses from impurity with e samth e radiative capabilit o t hydrogey n atom d witan sh density being proportiona e plasmth o t al density e comparisoTh . f temporao n l evolutionf o s plasma density in experiment and obtained by 0-D code showed the qualitative agreement between them (Fig.9) f radiatin o inclusioe % Th . 25 f gno impurit y with respect to plasma density leads to the saturation of plasma density at the level of 10*3 cnr^ for RF power input 200 kW while the same density without impurities is obtained at 30 k\V. CONCLUSION. The studied scenario of plasma build-up at the range of frequencie scyclotro n beloio e wth n frequency using optimized varian f crankshafo t t antenna seems to be promising both for stellarator and tokamak type devices. The development of such an antenna is important not only for URAGAN-3M experiments but also for production of dense target plasmas in stellarators and for RF assistance to Ohmic discharge start-up in tokamaks. This should provide the savin f inductogo r magnetic electronflud xan s pre-hea s welle studiea te Th .th f o s performanc f crankshafeo t antenn URAGAN-3e th n ai M torsatro progresn i e nar s now.

ACKNOWLEDGMENTS The present report was supported by International Science Foundation contracts No.UATOOO and No.U32000. The presentation of this report was made possible by International Atomic Energy Agency sponsorship.

REFERENCES [1J V.V.Bakae 10te Proc. th aJ h f t vo .Inte . Conf Plasmn o . a Physic Contrd san . Nucl. Fus. Research. London 1984, IAEA, Vienna, 1985, v.2, p.397 [2] Ye.D.Volkov et al. Proc. of the 14th Int. Conf. on Plasma Physics and Contr. Nucl. Fus. Research. Wurzburg, 1992, IAEA, Vienna, 1993, v.2, p.679 [3] K.NISHIMURA et al. IAEA Technical Committee Meeting VIII Stellarator Workshop, Kharkov, 1991, IAEA, Wienna, 1991, P.235 [4] N.T.Besedin et al. IAEA Technical Committee Meeting on Stellarators and other Helical Confinement Systems, Garching, Germany, 1993, IAEA, Wienna, 1993, p.277.

105 [5] V.E.Moiseenko, IAEA Technical Committee Meeting VIII Stellarator Workshop, Kharkov, 1991, IAEA, Wienna, 1991, P.207 [6] S. Pun, Nuclear Fusion, 27 (1987) 229 S.N.Golovato] [7 , J.L.Shohet, Phys.Fluids (19781 2 , ) 1421 [8] A. de Chambrier et al. Plasma Physics, 24 (1982) 893 A.I.LYSOIVA] [9 Inth 5t . , Tokal t Nie Conference, Toki, Japan, 1993 Fusion Engineerind gan Design 26 (1995) 185 [10] V.E.MOISEENKO,et al. 21st EPS Conf. on Contr. Fusion and Plasma Phys., Montpellier, 1994, V.18B, p.980. [11] V.E.MOISEENKO, et al.Sth Int. Toki Conference, Toki, Japan, 1993, Fusion Engineering and Design 26 (1995) 203.

106 XA9745748 DAMAVAND - AN IRANIAN TOKAMAK WITH A HIGHLY ELONGATED PLASMA CROSS-SECTION

R. AMROLLAHI Plasma Physics Department, AEOI, Teheran, Islamic Republic of Iran

Abstract

The "DAMAVAND" facility is an Iranian Tokamak with a highly elongated plasma cross-section and with a poloidal divertor. This Tokama advantage th s kha alloo e t plasm e wth a physics research undee th r conditions simila thoso t r ITEf eo R magnetic configuration r exampleFo . , the opportunit reproduco yt e partiall plasme yth a disruptions without sacrificing the studies of: • equilibrium, stability and control over the elongated plasma cross- section • processes in the near-wall plasma • auxiliary heating systems etc. The range of plasma parameters, the configuration of "DAMAVAND" magnetic coil passivd san e loops theid ,an r location withi vacuue nth m chamber allo creatioe w th plasme centevacuue e th f th th n o t f aa ro m chambed ran the production of two poloidal volumes (upper and lower) for the divertor.

1 Introduction

e "DAMAVANDTh " facilitn Iraniaa s i y n Tokamak wit a highlh y elongated plasma cross-sectio d wit an na poloidah l divertor (Fig.l). This Tokamak has the advantage to allow the plasma physics research under the conditions similar to those of ITER magnetic configuration. For example, the opportunity to reproduce partially the plasma disruptions without sacrificing the studie: f so - equilibrium, stability and control over the elongated plasma cross-section, - processes in the near-wall plasma, - auxiliary heating systems etc.

The range of plasma parameters, the configuration of "DAMAVAND" magnetic coils and passive loops, and their location within the vacuum chamber allo creatio wplasmthe centevacuu the the the nof aat of r m chambethe and r production of two poloidal volumes (upper and lower) for the divertor.

107 8-POLOIDAL COILS 4-INDUCTOR

2-SUPPORTING RINGS 1-TOROIDAL COILS

5-VACUUM VESSEL FLANGE

3-SUPPORTIN* RING,

SUPPORT

7-POLOIDAL COILS

6-PASSIVE SYSTEM

SUPPORT

DAMAVAN: 1 Fig. D sections

2 DAMAVAND

2.1 Main parameters

The main parameters of the "DAMAVAND" facility are given in Table I.

108 Tabl Mai: eI n Parameter f "DAMAVANDso "

Parameters Max. Values

Major radius of the torus R0 = 36 cm Transversal cross-sectiof no vacuum chamber 2a/2b = 20/65 cm

Toroidal field T 2 1. B= t Discharge duration t > 15 ms Elongatio f plasmno a cross- section k" = 1 ——> 4 Plasma density n(0) = 3 x 1019 m'3

Electron temperature* Te(0) = 300 eV

Ion temperature* Ts(0) = 150 eV Plasma current !„ ~ 35 - 40 kA Under Ohmic heating.

plasm e ratie th th f os o i a * *k semi-axes b= p /aK ,p

2.2 Toroidal field

A toroidal solenoid Fig. 1(1), Fig , separat0 mad.2 2 f o e e uniformly- placed sections (pancakes), produce stabilizine sth g magnetic field. These sections are fastened to four supporting rings Fig. 1(2) and to two load-bearing ones to provide the necessary mechanical strength. The radial mechanical forces are absorbe frama y db e supporting externall inductoe yth r Fig. 1(3madd )an f eo glass reinforced plastic pipe. The inductor frame, as well as load-bearing and supporting rings, resis e effec th f tbendin o t g moments resulting froe mth interaction betwee currente nth s throug toroidae hth l fiel poloidae d th coil d san l magnetic fields. Electrical parameters of the solenoid are given in Table II.

109 TOROIDAL COIUS

PUMP PORTS

SUPPORT

VACUUM CHAMBER

PUMP

Fig. 2 : DAMAVAND viewed from the top

Table II : Parameters of Toroidal Solenoid

Inductance Ohmic resistance Maximal current Maximal magnetic

Lt(H) rt(Ohm) I,(kA) field at R0 = 36cm B,(T)

5xl03 5xl02 -15 1.2

magnetie Th c field ripplequivalene th n ei t solenoid plane doe exceet sno d 2%. The toroidal field coil is cooled with water allowing the toroidal field impulse secondsw fe f o s .

110 2.3 Vacuum chamber

vacuue Th m discharge chamber torua , s with rectangula trapezoidad an r l cross-sections composee ar ,larg o tw e f sectiondo s (toroidal angleo 144°tw d )an small (toroidal angle 36°) ones. The sections are jointed together via the flanges thae watear t r cooled along their perimeters e forminar d a an dielectrig, c azimuthal gap. The vacuum chamber is made of non-magnetic stainless steel of thickm m .5 The small section of the chamber Fig. 1(4) has a rectangular cross-section wit eacn hI . inneh smalmm r 0 dimensionl 68 sectio x 0 n s20 singlthera e ear e vertical diagnostic o windotw d wan havinm m 0 ga cross-sectio 59 X 0 7 f no trapezoidal windows acros e e averagth th s t a ed cross-sectionan p to e th t a , bottom, 35 mm in size. The large section has a trapezoidal cross-section shown in Fig. 1(5). The angle between lateral faces Fig.l (5a) and the vertical axis is equal to 13 degrees. The largest dimensions of the section are equal those in the small section. Along the whole external perimeter, the diagnostic windows are located withi centrae nth l degreeangl8 1 gape f eth o sn si betwee toroidae nth l field coils. The window dimensions within the large section are: upper and lower windows e Th . mm 5 12 x 0 6 d an m m 0 window; 22 middle mm x th 0 n 0 8 i se7 x 5 2 total number of windows for diagnostics and evacuation is 76. Withi large nth e chamber sections, angle piece weldece sar . The loade yar - bearing chamber belts serving for supporting the set of copper passive loops whic protectes hi d agains direce th t t contact wit plasme hth a wit hthia n stainless steel liner. Thus the plasma column is surrounded with the walls of stainless steel. In one of the cross-sections a stainless steel limiter, 3 mm thick, is located. limitee Th r consist independeno tw f so t loblimitd an scross-sectioe sth e th o nt dimensions 2b, = 32 cm in the vertical direction and 2a, — 16 cm in the horizontal direction.

Ill 4 Passiv2. e stabilizing Held

In a Tokamak with an elongated plasma column cross-section configuration "thicka e us "n housinca e on preservo gt presee eth t equilibrium. However, after lapsa f skin-timeeo , = ^TC a^a **~~c= r' ~

presee th t equilibrium wil destroyede b lplasm e th d aan column wil roundee b l d and displaced to a greater major radius. In this expression c is the average mino e electricath r s e s housingthicknessi it radiuth o s f i l o d s6 ,an , conductivity. Therefor orden ei preservo t r presee eth t equilibriu longea r mfo r

time ( t z T4 ) one needs a set of passive loops and the corresponding vertical magnetic field produced by the currents through external conductors. A set of passive loops is necessary in order to provide the MHD-plasma stability (above verticae allth , l stability) reduco t d verticae an eth , l instability incremena o t t value inversely proportiona skie th n passive o timt lth f eo e set than I . t cased an , then the function of elongated equilibrium plasma column production is imposed upon the poloidal field produced by external currents. Four section passive th f t Figm o s ese m . 0 1(61 f )o separatep ga a y db wide, alon torue gplacee th sar d withi vacuue nth m chamber fous it rn i ,sections . Each section of the passive stabilization set includes two independent elements: interna externald an l . Each element turns it madn s i i , fouf eo r saddle-like loops connected with each other at the end faces. Large sections of the passive set, along wit connectione hth facesd en e , th hav t sa e four jumper meridionae th n si l plane. Thus each large sectio mesha s ni , interna externad an l l ones. Since eth whole passive stabilization set is made of copper, it is coated with a foil of stainless steel.

2.5 Poloidal field

The poloidal field coil set up is symmetrical with respect to the equatorial plane of the facility. An air inductor is used for the plasma current excitation and

112 sustainment structurs it n I . inductoe eth threa s ri e layer solenoid highm c 3 , 12 , wit averagn ha e radiu f 11.2 o sFig m c 5. solenoie 1(3)Th . insertes di d intoa glass-cloth-base laminate pipe and filled with an epoxy compound. The external pipe absorbs, efforts emerging fro solenoide mth wels a ,serves a l fixina s sa g element and a support for the toroidal field coil. The three windings of the inductor are usually connected in series. 3 The total inductance of the inductor coil is L; = 2.4 X 10~ H and its 2 Ohmi cx 10" resistanc3 Ohmmaximae = t r Th . s ei inductole fielth n di s i r limite mechanicas t it A y d. b currenKA a t 0 la 2 strengt , f equaT s o t i 6 d o t lhan the maximal curren inductoe th t penetrates ri magnetia y db c flu . 0.2f xo Vs 2 r shapinFo e plasmth g a cross-sectio r controllinfo d an n e plasmth g a equilibrium a number of poloidal field coils are used Fig. 1(7). One of these coils, controlling the plasma column cross section, is placed between the vacuum chambe toroidae th d an rl field pancakes Fig. 1(8). Other Coils, producine gth poloidal magnetic field, are placed outside the toroidal field pancakes, Fig l.(7). coile type o usede th sar f Tw so . Some coil placee sar d withi cute nth n s i special holders producin ring-likga e configuratio loopse th f multiwirA o .n e copper conductor, 75 mm2 in its cross-section, within a polyvinil insulation, is

used for these coils. Other coils are made from copper pipe (busbar) having a cross-section 12 X 15 mm an2 d curved into a ring. The loops are insulated from each other wit glasha s ribbon. The parameters of the coils and their destinations are summarized in Table III. Since the poloidal field coils are symmetrical with respect to the Tokamak equatorial plan totae th , l numbe f turno r s wil twice b l e those give tabln ni e III. In thi coordinate- s caseZ e coile th , th sf so locate lowee th n dri semi-space will equae b thoso t l similaf eo r coil uppee typeth n sri semi-spac taket ebu n witn ha opposite sign. The excitation coils are connected in series with the inductor, being used for compensation of the scattered field from the inductor in the region occupied by the plasma column.

113 Table III : Parameters of Poloidal Field Coils

Coordinates of the Cross-section Number cross-section centre dimensions Coil type of turns r,cm ± Z,cm lr,cm Iz,cm

5 23.5 _ 1 _ C.c 10 27 47 6 3 C 10 27 52.5 3 5.5 C 4 45 54.5 3 3 C 4 45 51.5 3 3 C 2 48 54.5 2.5 1 C.c 2 50.5 54.5 2.5 1 Ex 3 48.5 53 4 1 C 8 49 51 5 2 V.c 10 45 47 6 3 C 2 73.5 23.5 2.5 1.5 Ex 8 74.5 21.5 5 2 C 3 74 20 4 1 C.c 16 74.5 17 5 5 Eq 10 66 15 6 3 Eq 5 61.5 16.5 2 3 H.c

C Configuration C.c Configuration control Ex Excitation V.c Vertical control Eq Equilibrium H.c Horizontal control

114 e equilibriuTh m coils produc approximateln a e y verticale fieldth n i , plasma column zone, confinin plasme th g a column with respec e majoth o t rt radius. The configuration coils produce a quadrupolar (stretching) magnetic field extendin transversae gth l plasma column cross-sectio verticae th n ni l direction. The horizontal control coil is used for correction of the plasma column position with respect to the major radius by a feedback process. The vertical control coil, together with the passive stabilization loops, sustaiallowo t e plasme on nth s a colum verticaa n ni l directio feedbaca y nb k process. Finally configuratioe th , n control coil allowcorreco t e transversae son tth l plasma column cross-section configuration (e.g. its triangularity).

3 Evacuation System

The evacuation system consists of two identical modules. Each module include turbomoleculasa r pump, HBT-240 forvacuua d an , m pump, HBP-. 5DM Moreover, the evacuation modules are supplied with the valves for operative and emergency disconnection (connection) from/to the vacuum chamber of the facility s wela ,wits a l h technological gauge meter r controllinsfo vacuue gth m directl eacn yi h module. The ultimate vacuum in the chamber of the facility, reached by both modules operatin paralleln gi 10~ X 7 1 Torrs i , . The vacuum control system allows one to measure the gas pressure directly in the vacuum chamber, in various parts of the evacuation pipeline, as controo t wel s e spectraa l th l l structure e operatin th thad th f an o f t o es gga residual gas. Three vacuum gauge meters, are used for vacuum measurements. A mass spectrum analyzer is used for the spectral gas structure control. The control over all these active elements in the evacuation system is realized from an independent unit located next to the Tokamak control board. [1]

115 4 Gas Puffing

The gas puffing system includes a combination of gaseous vessels with palladium filters, which allows one to purify and to mix hydrogen and deuterium desirey inan d proportion piezovalva d an , e through whic operatine hth g gaseous mixture is puffed into the vacuum chamber of the facility.

5 Preionization

e preionizatioTh n syste productioe th ms use i r fo dcola f dno weakly- ionized plasm discharge th n ai e chambe facilite th f o ry befor emergence eth f eo a longitudinal electric field initiating the breakdown and for sustaining the discharge Tokamake pulsth n ei . preionizatios ga e DAMAVANTh e th n i D Tokama achieveks i d eithey rb microwava spara y b r ko e discharge. microwave Th e preionization system producallowo t e son emicrowava e discharg e chambe th e facilite electro th th n i e f t o a yr n cyclotron resonance frequency resonance Th . e conditio reaches nfieli e th : d r dfo

B. =

e toroidath whers i t l B emagneti c field induction e frequenc th d <• s i an >, f o y oscillations. The electromagnetic wave is excited with a magnetron operating at wavelengte th e seconTh . d cm hharmoni 2 A~ . f thico s frequency corresponds magnetie toth c fiel~ 1,0 dB 7 (thiT s valu fiele closs ei th d o et magnitud e th t ea

radiu 0 —cm)6 sR 3 . Operating condition magnetroe th r sanod: fo e nar e voltage,

; pulskV e 2 /us 0 duration1 generatee U20 ;th = a . = t kW A ,d 0 power4 ~ P ,

6 Glow Discharge Conditioning

Production of a pure plasma with a low level of impurities depends, to a great extent, on the conditioning of vacuum chamber and limiter walls. The

116 conditionin vacuue th f go m chambe s realizei r d wit gloha w discharge. Glow discharge electrodes are made of stainless steel rods, 10 cm long, and ~ 1 cm in diameter. The e introducear y d int uppe n a loweoa r o r r port througe th h insulators within the large section of the vacuum chamber in two cross-sections. A potential positive relative to the chamber is applied to the electrodes. The conditioning is accomplished by a discharge in argon (xenon) for 2 hours, in helium for about 4 hours and in hydrogen (deuterium) for about 4 hours. The discharg . Afteee kV parameterth r, 1 voltag A — 5 : curren0. e U e — ar s I t evacuatio chambee th f rathena o o rt Torr)6 r hig10 hx , vacuufurthe1 < mrP ( conditioning up to the operating conditions is achieved by Tokamak discharges.

7 Feedbacks

identicao Tw l system automatif so c contro usee stabilizatior lar dfo e th f no plasma column position with respec majoe th o rt radiu witd san h respece th o t vertical direction e automatiTh . c control system circuit include e plasmth s a column (the object of the control), the diagnostics of displacement, a regulator and an executive device, a control coil, and a power supply source (capacitor bank) including a voltage converter. The diagnostics of displacement include a Rogowski coil and magnetic probes, integrators, a divider, a programmator and a summator n autonomouA . s voltage inverte s a converterusei r s a d e On . capacitiv powes usea i s , da rkV e suppl1 storage r = fo 0.0= y U C ,9, F voltage inverter botn i s h system automatif so c control (wit. Z) hd respecan r o t

The operating voltage is Up = 250-400 V. Both systems operate in a free- runnin kHzg1 , mode~ f maintainin t a , positioe plasmge th th f no a column with respect to r and Z, within an accuracy of the order of 2 mm. Tokamae Onth k ther systea s ei m compensatin electromotive gth e force induce controe th n di l coils with respec radiue th witd o t an sth respece th o t t configuration, by the inductor, by the equilibrium coil and by the configuration coil.

117 8 Power Supplies

Toroida poloidad lan l magnetic fields powecoil8 Tokamay f sb o rd fe e kar supply systems. Every power supply system includes main elements shown in Fig. 3. Each poloidal coil (Table III) can be connected with different power supply. The parameters of power supplies are presented in Table IV.

-380V

: Fig3 . 1:3- Phase thyristor switch rectifier: 2 ; curren: 3 ; t commutator; 4 : device for the automatic control over the level of charging capacito : resistor R , C r .

118 Table IV : Parameters of Different Power Supplies

No. Destination C Rectifier Commutator V0 Ipulse, F kV kA

1 T.F. 0.21 5kV,5A 4 Spark gaps 3 15 HPT-5

2 SIB 0.06 5kV,5A 2 Spark gaps 3 15.5 HPT-5

3 FIB 0.0066 5kV,3A 1 Sparp kga 5 13 HPT-5

4 EFB-1 0.01 5kV,3A 1 Spark gap 2.5 4 HPT-6

5 EFB-2 0.01 2kV,5A Spar1 p kga 2 3.7 HPT-6

6 SFB-1 0.03 5kV,5A 1 Spark gap 2.5 9 HPT-5

7 SFB-2 0.03 5kV,5A Spar1 p kga 2.5 9 HPT-5

8 FBB 0.09 lkV,lA 2 Voltage 0.4 0.5 Invertor

T.F. Toroidal field bank, S.I.B. Slow inductor bank, F.I.B. Fast inductor bank, E.F.B - 1 Equilibrium field bank-1, E.F.B - 2 Equilibrium field bank-2, S.F.1 - B Quadropolar field bank-1, S.F.B-2 Quadropolar field bank-2, F.B.B. Feedback bank, Working voltage.

119 9 Diagnostics

The initial set of diagnostics for DAMAVAND includes various diagnostic means providin measuremene gth controe maie th th d nf l to an plasm a parameters, necessary at the start-up and at the adjustment stage, as well as in further research program. Moreover, the diagnostics provide signals for controlling the systems producing and sustaining an equilibrium plasma configuration. maie Th n diagnostics means include t initiasummarizee ar th t n di se l n di Tabl . LayoueV Tokamaf o t k diagnostic shows si Fign ni . .4

Tabl Initia: f Diagnostic o eV t Se l DAMAVANr sfo D

No. Diagnostics Measured Parameters 1. Magnetic probe flud sxan loops Plasma configuration, MHD- perturbations.

2. Spectroscopy in visible range radiatio f hydrogeno d nan impurities

3. Langmuir probes temperatures Te(r,t), T;(r,t) density n;(r,t) at the periphery

4. Bolometers radiation losses

5. Scanning Atomic particle ion temperature T;(t) analyzer with a solid target distribution io n function

6. Scanning photoelectron electron temperature Te(r,t), spectrometer electron distribution function, radiation of impurities

7. Diagnostics of hard X-ray leve f haro l d X-rays radiation

120 Opti.c* Magnetic Probes

F(.»..>: Loop

7 - s c b o Y 1 c i. I y

Optics

/ V< i

Xcu.lrul -\nalizcr AJdss Anulizcr

Afa(/•)<•< I i.c P Li'inilcr

. Fig. 4 : Layout of Tokamak diagnostics

10 Plasma Parameters

tablpresentee e Inth ar eI manimae dth l plasma parameters obtainee th n di first experiments.

Typical plasma current oscillogram loo, sIp p voltag 0 verticad eU an z A l horizonta positior plasmlA e th f no a column, hard X-ra , opticayHX l lines D/8, Carbone III, Carbon givee ar Fig.5,6,7 n no V e I . Value plasmf so a electrod nan ion temperature is presented on Fig.8.

11 Conclusion

In conclusion, the construction and the operation of Damavand, a medium- aspect-ration small-sized Tokamak will enabl A.E.O.e eth provido It valuablea e contribution to the National Fusion Programme.

121 40- V

30-

20-

HX,a.u

lljry«

4 6 8 10 12 14 t,ms

: Har5 Fig. d X-ray diagnostic

122 t.ms

: SpectroscopFig6 . y diagnostic

123 0

Fig. 7: Spectroscopy diagnostic

124 ev 350

3 **»ie* 300

250

200

T 150 ;

100 j 50 11 , 1 1 ... —— 15 0 01 5 t, ms

FigTemperature: 8 . electrof o s iond nan .

125 Plasma Physics Department of Atomic Energy Organization of Iran believes that, based on the research opportunities offered by Damavand new possibilities Islamie th , c Republi Iraf makcn o n ca importan n ea t contributioo nt the international fusion programme, and increase its own technological base on this field.

Reference

[1] AMROLLAHI, Reza : "Iran et Tokamak", Doctoral Degree Thesis, University of Paris, 1994.

126 niiii — in"11 ••'•' PRESENT STATU RESEARCF SO H ACTIVITIE E TH T SA XA9745749 NATIONAL INSTITUT FUSIOR EFO N SCIENCE AND ITS ROLE IN INTERNATIONAL COLLABORATION

J. FUJITA National Institute for Fusion Science, Nagoya, Japan

Abstract

In the National Institute for Fusion Science (NIFS), Japan, a helical magnetic confinement system named Large Helical Device (LHD undes )i r construction with objective comprehensivf so e studie higf so h temperature plasma helicaa n si l syste investigatiod man helicaa f no l reacton a s ra alternative approach. Superconducting 2,m= majo, = coil produc , 10 l m s of r 9 radiu3. esteada = sR y state helical magnetic field for confinement, together with poloidal coils on LHD. The magnetic field strength on the axis is 3.0 T in the phase I and 4.0 T in the phase II experiment. The plasma major radius averaged an 3.7s i , 5m D do plasmplasme fLH Th . a m wilradiu produce6 e lb 0. s si heated dan d with ECH, and further heated with NBI and ICRF. It is also planned to produce a steady state plasma in LHD. It is expected to have the first plasma in 1998. Small devices such as CHS and others are under operation in the NIFS for supporting the LHD project. The Data and Planning Center of NIFS is collecting, compiling and evaluating atomic and molecular data which are necessar nuclear yfo r fusion research. The talk will include the present status of the construction of LHD, research activities on the development of heating and diagnostic devices for LHD, and experimental results obtained on CHS, JIPP T-IIU and other devices. The role of NIFS on promoting IAEA activities to bridge large scale institutions and small and medium scale laboratories for world-wide collaborations in the fiel f plasmdo a physic fusiod san n research will als introducede ob , together wit iden ha f ao organizin gregionaa l cente Asian ri .

1. Introduction The nuclear fusion research is a long term project. It is necessary to gather every possible brains fro l ovem worlal e rth realizo dt efusioa n reactor flexiblA . e strategy t onlno ,y alone gth tokamak line t alsbu ,o wit varietha f alternativyo e approache s necessarysi , with close international collaboration. Based on these considerations, the main research activity in the National Institute for Fusion Science (NIFS), Japan bees ha , n directed toward stude sth plasmf yo a confinemen heatind an t n gi a helical system complementara s a , y approac f tokamaho k researc comprehensive th o ht e understanding of magnetically confined toroidal plasmas. For this purpose, a helical device called Large Helical Device (LHD beins )i g constructe cite Tokf sitth yw o n ei ne i a [1,2] t JIPe da Th P. T-IIU tokamak and the Compact Helical System have produced important experimental data to support projectD theLH . Super computer usee largr sar dfo e scale plasma simulatio powerfua s na l tooo t l understand physical phenomena in plasmas.

127 Among various activities at NIFS, diagnostic developments for LHD will be a typical example which bridges the works on the large scale device and those on small devices. These activities, together with some of experimental results obtained on JIPP T-IIU and CHS will be explained in the following.

ProjecD LH e t Th . 2 physice Th projecD s subject LH summarize s i te th f so followss da : 1) Produc higea h nrrcurrentless plasm studd aan y transport proble morden i obtaio rt n basic data extrapolatewhice b n hca reactoa o dt r grade plasma. 2) Realize a high /? plasma of averaged beta value higher than 5 % which is necessary for a reactor plasm studd aan y related physics. 3) Install diverto d obtaian r n basic data necessar r steadfo y y state operation through experiment controllinf so g quasi-stationary plasmas. Stud) 4 y behavio higf ro h energy particle helicaa n si l fielmakd dan e simulation experiments of alpha particles in reactor plasmas. 5) Mak complementarea y stud tokamaf yo k plasma wideo st deeped nan comprehensive nth e understanding of magnetically confined toroidal plasmas. targee Th t plasma parameter stude th f abovr yo sfo e subject settlee ar s d under three operation modes: 0 2 3 . T 4 = B , s 3 0. - 1 0. = m' E 0 T , 1 keV14 = ) - v Hig a mode7 3 ,n r = h « t T : m T> 1020 keV nV3s ( Q ~ 0.35 ). 19 3 2) Hig mode{ hT 10. x T 2 : keV4 0 7,(0 = rn 1 v = a = « ,),B . T 2 - 1 Hig) 3 = hB , (3% mode 5 > t :f

SC Poloidal Coil Cryostat

Fig. 1. A conceptual drawing of LHD

128 Tabl Machin. e1 e parameter Large th f eso Helical Device (LHD) under constructio NIFSt na .

Parameter Phase I PhasI eI Major radiu) s(m 3.9 3.9 Coil minor radiu) s(m 0.975 0.975 Averaged plasma radiu) s(m 0.5 - 0.65 0. 50.6- 5 Plasma aspect ratio 6-7 6-7 / (pole number) 2 2 m (pitch number) 10 10 Plasma volume (m3) 20-30 20-30 Magnetic field Cente) r(T 3 4 Coil surface (T) 6.9 9.2 Helical coil current (MAt) 5.85 7.8 Coil current density (A/mm2) 40 53 Liquid helium temperatur) e(K 4.4 1.8 Plasma duration (s) 10 10 Repetition time (m) 5 5 Heating power ECRH (MW) 10 10 NBI (MW) 15 20 ICRF (MW) 3 9 for steady state operation (MW) — 3 Neutron shotr yielpe )n d( — 10x 4 172.

machine Th e parametersummarizes i D LH f so Tabl n di . e1 Superconductin g coil f polo s e number / = 2, pitch number m = 10, major radius R = 3.9 m, produce a steady state helical magnetic field for confinement, together with inner and outer vertical coils, and shaping coils. The pitch modulation which is employed in LHD brings a better helical symmetry and clear helical divertor structure. It is designed so that the magnetic field configuration can be changed as flexibly as possible for a variety of experiments. Because the helical coils are wound continuously around the machine, the windings are being carried out on site instead of constructing the device in the factory and transport it to the site. The helical coils are cooled with a pool-boiling manner, while the poloidal coils of cable-in-conduit type are forced-flow cooled. The whole machine is set in a large cryostat. The construction will take more thayearo ntw s from nowf 1997o wil d d completee ,conceptuab an A len . e th n di l shows i drawinD Fign i LH . f .1 g o

129 Table 2. List of Diagnostics for LHD.

Diagnostics Purpose Descriptions

7D, plasma pressure, Rogowski, Mirnov, Flux loops Magnetic Probes positio shapd n an plasm f eo a Microwave Interferometer nj wavem 2mm1m ,/ single channel

FIR Laser Interferometer nj(r) 1 19 um-CH3OH laser, 10-channel

Microwave Reflectometer ne, ne fluctuation under development

Thomson Scattering re(r)./»e(r) 200 spatial points

ECE Tt (r, 2} 2-D imaging X-ray Pulse Height Analysis Tf, impurities 20-ch Si (Li), 4-ch Ge detectors Neutral Particle Analyzer r,,/(£) radial scan Charge Exchange TM,V,(r) us diagnostif eo c neutral beam Spectroscopy / Polarimetry q(r} 4 X-ray Crystal Spectroscopy Tt(r) 0.1-4nm, ?L/AX:10 3 Neutron Diagnostics neutron fluxt T , NE-213 detectors counterse H , , high energy particles activation foil Bolometers P* (r) metal film, silicon diode, pyroelectric detector 4 VUV Spectroscopy impurities, Tl 1 -200nm, ^/AX:10 4 Visible Spectroscopy «0(H),Zeff 200- 700 nm, MAX: 5 \ 10

Langmuir Probes ^«e Fast scannin fixed gan d probes Visible/Infrared TV plasma positionPWI , TV systems wall/limiter temperature Soft X-ray Diode Array MHD Oscillations silicon surface-barrier diodes MW/FIR Laser Scattering microinstabilities 1 mm/ 195 jam multichannel Heav Bean yIo m Probe plasma potential, fluctuation AuT or Tl+, 6 MeV, 100 \i A Diagnostic Pellet particle transport Hydrogen/Double laye pellete ric , C, Li High-energy Particle high-energy particles beae LH ) MeVmi/ 2 ( mA 0 ,1 Diagnostics probe, particle detector probes

3. Plasma Diagnostics for LHD Although LHD is not a tokamak but a helical system, the diagnostics for LHD are not much different from those for tokamaks. The essential difference from a tokamak is that the plasma shape has no axial symmetry, and an elliptic plasma cross-section rotates poloidally along the magnetic axis. Then, 3-dimensional diagnostics shoul providee db r LHDdfo . Moreovere th , plasma should be diagnosed through long and narrow observation ports which are drilled across cryostate th . Therefore diagnostice th hav,D eLH r varioufo s s features different from thosn ei

130 conventional tokamaks othee th n advantagrn a handO . s i t i , f helicaeo l syste mhavo t centeo en r pol windingr eo transformere th f so thao s , t inboard observation port available sar LHDn eo . It is also planned to produce a steady state plasma in LHD. A proper function of divertor is vitally important in a long pulse machine. Special caution should be taken in the design of divertor diagnostics. However, the divertor diagnostics are still in a phase of under development. Although no D-T operation is planned on LHD, the D-D operation will produce a significant amount of neutrons. Major parts of diagnostics are placed in adjacent rooms or underground of the main experimental hall, whic biologicalle har y shielded with thick concrete walls. The diagnostics summarizee whic preparatioi ar h e D har LH r Tabln dfo i . e2 Man f thesyo e diagnostics utilize the results of developments achieved on the supporting machines, JIPP T-IIU and CHS. Among a number of diagnostic developments, that of a heavy ion beam probe (HIBP) wil explainee b l examplen a s da .

4. Development of Heavy Ion Beam Probe plasme Th a potential profile radiar o , l electric field distribution importann a s i , t quantite b o yt measure helicaa n di l system, becaus strongls i t ei y relateplasme th o dt a confinement properties. Althoug measuremene hth poloidaf o t l rotation spee utilizes di fino radiade t dth l electric fielda , direct method to obtain the potential distribution is necessary. A heavy ion beam probe has been designed for the potential profile measurement on LHD [3 - 5]. Because of a large size of the plasma and a high magnetic field intensity on LHD, a heavy ion beam such as gold should be accelerated as high as 6 MeV for the beam to penetrate the plasma, even at the magnetic field intensity of 3 T in the first phase. Due to non axi-symmetric configuratio helicaf no l magnetic field injectee th , d beam doe t remaisno toroidaa n i l planet bu , traverses bot toroidan hi poloidad an l l directions e secondarth r e samfo Th s ei .y beama s A . consequence, the secondary beam comes out from the observation port located toroidally next to the beam injection port. A schematic drawing of HIBP system for LHD is shown in Fig. 2. For accurate measurement of plasma potential and its fluctuation, the beam should have an energy sprea smals da possibles a l . From this viewpoint gola , d negativ sourcn eio plasmaf eo - sputter-typ bees eha n developed [6]. Negative ions thus produce acceleratee dar tandea n di m manner, and converted into positive ions through a stripping gas cell which is at the high voltage end. A test stand has been constructed for studying negative ion production, extraction, energy dispersion and charge stripping efficiency. An energy analyzer of a high resolution and of high voltage characteristic alss si o being designed. On JIP HIBV Pke P T-IIU0 syste50 a , ms operatingi interestinn a s A . g result plasma , a potential change associated with the sawtooth crash has been observed. As a prototype of the HIBP system for LHD, a 200 keV system has been constructed and applie CHSo dt uniquA . e ide f activao e trajectory control metho s confirmedi woro dt thin ko s system.

5. International Collaboration The promotion of international collaboration is one of the important roles of NIFS. The diagnostic developmen gooa s i t d exampl collaborative th f eo e work basiA . c idea will comt eou fro smaly man l laborator fror yo m individuals testee b ide e n dTh aca . usin gsmala l device before constructing an actual equipment for a large device. At the same time, this kind of work will

131 Charge Exchange Gas Cell

'G^*— Au" Ion Source 0 1 2345m

Basement

Fig. 2. A schematic drawing of the 6 MeV heavy ion beam probe.

132 provid a gooe d opportunit r younfo y g scientist e traineb o t sd wit a hparticipatio e th n i n experiment. The establishment of a regional center for training young scientists and for encouraging the growth of innovative ideas will be of great help to promote the IAEA activities to bridge large scale institution d smald mediuan san l m scale laboratorie r world-widfo s e collaborations.

6. Conclusion At NIFS constructioa , devicD preparatord LH progressn ean i f no e ar yD . workLH r fo s Basic researches using smaller device alse sar o being carried oute diagnostiTh . c developments gooe ar dD examplfoLH r combininf eo worke gth wida n si e rang fieldf e o places d san . The NIFS is ready for contributing international collaborations with an idea of regional center in Asian district.

References

[1] A. liyoshi, M. Fujiwara, O. Motojima, et al., Design study for the Large Helical Device, Fusion Technology, (19907 1 ) 169-187. [2] M. Fujiwara and the Large Helical Device Project Team, Status of large helical device project, Fusion Engineering Design,d an (19953 2 ) 547-561. . FujisawaA ] [3 . IguchiH , . Taniike,A heav. Hamada V . Sasa Y bea n M ,d Me y io o 6 man A , probe for the Large Helical Device, IEEE Trans. on Plasma Science, 22 (1994) 395-402. . SasaoM ] . Okabe[4 Y , . FujisawaA , Yamaoka. H , . . FujitaWadJ M , d aan , Developmenf o t negative heavy ion sources for plasma potential measurement, Rev. Sci. Instrum., 63 (1992) 2726-2728. [5] A. Fujisawa, H. Iguchi, Y. Harnada, M. Sasao and J. Fujita, Active control of beam trajectorie heavr s fo bea n yio m prob helican eo l magnetic configurations, Rev. Sci. Instrum., 63 (1992) 3694-3700. [6] T. Taniike, M. Sasao, A. Fujisawa, H. Iguchi, Y. Hamada, J. Fujita, M. Wada and Y. Mori, Ion sourc stripped ean cels rga l tandedevelopmentV Me m6 heave th bean r ysio fo m probn eo the Large Helical Device, IEEE Trans. on Plasma Science, 22 (1994) 430-434.

SEVf iB£

M.M. MASOUD Plasma Physics Department, Nuclear Research Centre, Atomic Energy Authority, Cairo, Egypt

Abstract

statue Th plasmf so a physics research activitie Egypn si revieweds i t . There ear nine institutes with plasma research activities. The largest is the Atomic energy Authority (AEA), which has activities in fundamental plasma studies, fusion technology, plasm d lasean a r applications d plasman , a simulatione Th . experiments include Theta Pinches, a Z Pinch, a coaxial discharge, a glow discharge, a C02 laser, and the EGYPTOR tokamak.

Plasma physics laboratory researc Egyphn i t started 195 Nationan i 9 l Research Cente groua y rb p workin Z-pinche th n gi glo C ,wD discharge Hollow cathode discharge dischargeF ,R , using electric probe othed san r conventional diagnostics. In 1962 two experimental plasma physics groups are formed in Atomic Energy Authority (AEA) to study cold and hot plasmas where National Research Center group joined them. First group worked in hot plasma machines such as, Theta and Z-pinches, shock wave tubes, using electric and magnetic probes, microwaves, high speed streak camera, pic coilsp kspectroscopu d ,an diagnosticsr yfo secone Th . d group studies were dischargess ga , in sourcen ,io differenf so t types, acceleratio chargef no d energparticlew lo yd beaan s, m injectors. plasme Th a physics progra slowes mha d down fro mseverao t 197 e 198o t 0du 5 l reasons. Today ther severae ear l research institutes workin plasmn gi a physics theor experimend yan t These institute theid san r facilitie- : e sar

1- Plasma Physics Department, Nuclear Research Center, AEA. - Tokamak, Plasma Focus, Theta Pinch, Z-Pinch, Coaxial Plasma gun, Glow discharges, Lasers, Theory, Simulation, groups

2- Plasma Laboratory, Physics Department, Facult Sciencef yo , Al-Azhar University, Cairo. - Plasma Focus, Z-pinch, R.F. Discharge, Glow Discharge.

3- Plasma Laboratory, Facult Engineeringf yo , Zagazig University, Zagazig. - Plasma Focus, Z-pinch, Microwave, R.F. Glo, ,wDC Discharges.

4- Physics Department, Faculty of Science, Zagazig University, Zagazig. - Z-Pinch.

5- Physics Department, Facult Sciencef yo , Cairo University, Cairo. - Glow Discharge, Vacuum spark.

135 6- National Institut Lasef eo r Enhanced Sciences, NILES, Cairo University. - Laser produced plasma, Glow Discharge.

7- Electrical Engineering Department, Facult Engineeringf yo , Cairo University. - Glow Discharge, Z-Pinch, Theory.

8- Nuclear Engineering Department, Faculty of Engineering, Alexandria University, Alexandria. - Glow Discharge, Simulation.

9- Physics Department, Facult Sciencef yo , Assyout University, Assyout. - Glow Discharge, Arc Discharge.

This is beside some individual works in the other universities and research institutes.

This doe includt sno researce eth h institutes work acceleratorsn si sourcesn io , , bead an m injector than I s. t fiel maie dth n Institut sites en i AEAIo dn i , Source Acceleratod san r Department

In the following sections a short description of the research facilities and activities in Plasma Physics Department, AEA.

researce Th h activit departmene th n yi divides ti d into four groups: 1) Fundamental plasma studies, theoretical and experimental. 2) Fusion technology, theoretical and experimental. 3) Plasm lased aan r applications. 4) Plasma simulation.

1) Fundamental Plasma Studies Theory theoreticae Th l studie mainle sar y directe investigatdo t probleme eth : f so - Nonlinear interaction wavd san e generation plasmasn si . - Instabilities specially, Beam-plasma, Electro- acoustic, current convective, Buneman Drifd ,an t - Surface waves, excitatio interactiond nan .

The new trend in the theoretical study is the use of the analytical method to describe the nonlinear plasma dynamics in situation closely to the experiments.

Experiments a) Z-Pinch Four meter z-pinch [1] is used for pre ionization and preheating of the rest gas in the 3.5 m. theta pinch experiment. It is also combined with the theta pinc produco ht stablea e screw pinch. Another 0.3 m length, 0.25 m diameter, z-pinch operated with relatively slow ris e[is.)0 tim1 - capacito e( r ban uses ki producdo t stablea plasmt eho r afo the stud additionaf yo l heating problems plasme Th a. temperature reached

136 50 eV and electron density up to 1()14 cm'3. Microwave and particle beams will be introduced to the pinched column later. mose Th t interesting result smaln i s l z-pinches tha higf i t h current discharge flows throug curren, hit t disruption occur accompanie] s[2 y db hydro -magnetic instabilities.

b) Theta-Pinch Theta-pinco Thertw e ear h machine operation si n durin las e years0 gth t1 e th , 3.5 m. and 0.8 m.. Both machine's capacitor banks were developed several times. The 0.8 m. 6-pinch is one of the old machines in the plasma department. With capacito. kj.4 kV f 5 -o 2 the re us bank pea,a k discharg. e currenkA 0 12 f to riss witu e0 h tim1 obtaineds ewa peae .Th k plasma temperatur densitd ean y were 180 eV. and 1.4 x 10*5 cm'3 respectively. A burst of microwave radiation referred to the excitation of the lower hybrid electron cyclotron oscillations and its harmonics was detected. Most of the radiating energy was carried by the second harmonics. Magneti cinterestine reconnectioth f o e on g s nwa phenomena observed, which imposes the modification of the system. The latest modification of the 0.8 m. 6-pinch is the use of two banks, fig.(l), to enhance magnetic field reconnection currene Th . t research course th s ei stud relatioe yth n between magnetic reconnectio othed nan r process sucs ha energetic particle differene th d san t instabilities.

4kJ bank discharge tube

to vacuum pumps

FIG. Modified 1. 6-pinch. cm 80

137 glass window glass-tub*

v coil discharge \ chamtwr liner Z-pinch- cables. hot wire leakage » —• . System.

cables to the main bank

FIG. 2. 3.5 meter 6-pinch assembly.

138 6-pinc. m ex-CuIhan 5 a 3. s h i e machinTh K mU e re-installe modified dan o dt reduce its electrical noises and to suit the safety regulation in the laboratory. consistt I ris. us e 7 maitime)f - s o . n. -70kA ,ban kj 0preionizatio 8 k(4 n z-pinch bank(3 kj.-125 kA.-1.5 us.),bias magnetic field bank(4.8 kJ.-3.5 kA.-2 ms.),and screw pinch bank, z-pinch bank,( kA.-0 7-5 . 6us)kj . Schematic diagrame th f so 6-pinch coils and z-pinch assemblies are shown in fig.(2). The pinched plasma column temperatur4x10*d an . 5eV cm'densit0 d 3 e15 an e respectivelyyar . The current course of study on 3.5 6-pinch is directed toward the dynamics of plasme th a sheath heatine ,th g mechanis pinchee th mf o d column spectrad ,an l line's transition probabilities.

Fusion Technology Experiments

maio tw n e experimentTh Egytoe th e rsar tokama Atod kan n plasma focus.

Aton Plasma Focus A powerful plasma focus system "Aton", Mather type, has been designe d1992, , fig. (3), constructe operated dan 1993n di . The power supply of the system consists of two modules, each has 25 capacitor, 1.5 uF. 45 kV., which can store 38 kj. energy. Each module gives a peak current of 0.5 MA. for charging voltage of 35 kV with rise time of 3.2 us.

3-Ovter 1-lntuMw i window Intel bgneslltt 7-tocwim fkigr t-ftftnan chamfetr

•-to ncwm >nnm

7S*I 7i«f T.SJll 7-Sjrt 7»»t

. <*>

FIG. 3. Cross-section of the plasma focus apparatus (1 M.A. Aton).

139 researce Th h program carrie plasmn do a focus from 199 199o 3t 5 was: -- Matching condition. -- Axial and radial phase dynamic. -- Self induced axial magneti focu e effecs cth it fielsn d to stability dan . -- X-ray and energetic particles emission. - Introduce other diagnostics, laser shadow graph, spectroscopy.

The result sho] s[3 w tha radiae tth l plasma sheath curren curvilineaa s tha r structure that is responsible for the induced axial magnetic field. This field reduce sheate sth h velocit depressed yan instabilitiese sth .

increaso T plasme eth a focus temperatur densityd ean fasa , t rise time high discharge current, single pulse, wil useddevelopee A b l . d desigo nt chang capacitoe eth r bank Marko ,t s type bees ha ,n don usiny e b sam e gth e component systee Th . mundes i r constructio wil d operatee nb lan d 1996.

Egytor Tokamak Egytor tokamak is a small one, Taylor type, which was operated successfull severar yfo l years (Unitor Dusseldorn )i f university bees ha nt I . reconstructed in 1995 in Plasma department, AEA, Egypt. Schematic view of Egyto shows ri figurn i e (4).

OH Tranforaer Cor Ohmic heating TF Toroidal field coils VF vertical field coils PL Poloidal liaiter (optional) TL Toroidal liroiter P Plasiaa. P Window

FIG. Schematic4. view tokamak. ofthe

140 Egytor main parameters are:

Large radius R = 0.3 m. Minor radius a = 0.1 m. Discharge Duration = 50 ms. Loop voltage U = 1.2 V. Ohmic power = 60 kW. Toroidal field = 1.4 T. Poloidal field = 0.08 T. Vertical field = 0.01 T. Filling pressure = 1 .SxlO'3 mbar Plasma current = 36 kA. Electron temperature = 180 eV. Ion temperature = 80 eV. above Th e parameters have been obtained previousl ] accordiny[4 e th go t current running conditions. An additional bank's energy will be demonstrated withi nexe nth t yea increaso rt plasme eth a current. Also auxiliary heating methods will be considered later.

The training of the young scientist is the main goal of the running program. They are studying, the cleaning discharge, continues scanning of the mass spectrum, discharge sequences, ohmic current shaping bacd ,fe k system using the microprocessors, laser scattering, microwave interferometer, laser induced fluorescence, spectroscopy, data acquisition conventionasysteme th d ,an l diagnostics. The tokamak simulation group have started their task. They use the available codes beside they start to do their own code tailored to the problem under investigation theoreticae .Th familiae l grouar w prno wit tokamae hth k physics. The involvee yar studyinn di transpore gth t phenomen edgd aan e physics problems.

The proposed experimental program that will star 199n ti : 6s i i —Plasma stability dependence on the discharge and plasma parameters, ii —Eroded material from the limiter, and plasma behavior near limiter. Hi-Introducing other diagnostic techniques.(CCD camer Heliud aan beamn mio )

Plasm Lased aan r Applications

The main interest of this group is to be familiar with the technology of plasma and laser for material science and medical uses. The running machines gloe ar w discharge, coaxial gun, small plasma focus tunabld an , lasere edy . A powerful C(>2 lase designes ri industriar dfo l application wild san l be constructed this year. gloe wTh discharge simple sar e experimen t rich tphysicbu n i d san technology. The study of normal glow discharge reveals a complicated structure of the electron beams of the discharge. It has been found that the detecting of different electron beams energy sometimes was due to the diagnostic method. Henc diagnostie eth c tools fundamental theorconsideo t s y distributioha e rth n function specially in the overlapping area of two electron beams. The sputtered material parameter from the glow discharge electrodes for different condition undes si r investigation. The coaxial plasma gun produces a stream of wide range of plasma parameters, plasma stream velocity, temperature densityd ,an researce Th . h

141 studie carriee sar 4-1n d o . coaxia0kj l discharg controo et stabilized lan e sth plasma. The interaction of the plasma and different material substrate will be carried later. A small plasma focus works with heavy gases is used as a x-ray source for medical application study.

References [1. Thesis]Sc . MansouM , , FacultA. . r A Science f yo , Cairo University, (1995) [2] Masoud M.M., Bourham M.A., Sharkawy W., and Saudy A.H. Z. Naturforsch. 41a, 120 (1986) [3] Soliman H.M. and Masoud M.M. Physica Scipta 50, (1994) [4] Uhlenbusch J., Third Workshop on Plasma and Laser Physics, Ismailia, Egypt, Oct 3-7,1993. Forschungszentrum Julich, Bilateral Seminar of the International Bureau, Volume 15, p.47, (1994)

142 SLOW BANK SYSTEM OF SBVP-TOKAMAK: XA9745751 A SHORT REPORT

R. RAY, P. RANJAN, S. CHOWDHURY, S. BOSE Plasma Physics Group, Saha Institute of Nuclear Physics, Calcutta, India

Abstract

SIMP Tokama mads kwa e operationa Julyn i l , 1987 powee .Th r supply system of the tokamak at that time was designed for a plasma duratioa pea r k fo arounf plasmn o s m . d2 akA 5 curren7 f o t Efforts were directe increaso dt e this duratios witm e 0 th 2 h o nt helsloa f wpo bank system designe woro dt conjunction i k n wite hth original fast bank system desige Th . n aspect e systeth f so m were complete systee th bees d mha n an d partially executed [11. Subsequen o thit t s partial implementation, efforts were directed to incorporate the necessary control system and interface facilities between the existing fast bank and the developed slow bank systems e significanTh . t feature controe th f so l circuite sar that they work according to a well thought out sequences of logic and are designed to guard against possible failures in the existing or the developed power supplies. Efforts have been put to make the operatio systee th mucs f mna o h user-friendl s ycoula e workeb d d out within certain practical constraints. The control circuit and interface facilities have been put to extensive tests and are found to work satisfactorily. The entire power supply system is now in differenr activfo e eus t research programme groupe th n si .

1. Implementation of Slow Bank System for SINP-Tokamak, R. Ray et.al, Saha Centenary Symposiu Nationah 8t d man l Symposiun o m Plasma Science & Technology, Allahabad, October 11-14, 1993.

Introduction:

A progransn undertakes ewa exteno nt e plasmdth a duration

from around 2 ms to 20 ms by adding a second bank of

capacitor to the existing bank. This bank has a separate

power supply system for charging but this bank discharges

through the same set of ignitron switches.

This second banbees kha n named Slow Bank System

143 Basic Idea:

Both the fast and slow banks are separately charged from

two different power supplies. Both the banks are kept

isolated through an appropriate diode stacks.

Charging patter decides wele ni th l y knowdb n equation

L.n i ' T ril T " ' ji" = t...... vly dt C

where E is source voltage for charging.

During discharge, the controlling equation is

0 = t id / - + i R + ^ L dt C Three solution possible sar threr efo e set conditionsf so .

4L2

ii) -2

2 LC 4L2 LC

(iicase& non-oscillatore ) )ar s(i d casan ye (iiis i ) oscillatory.

Solution for non-oscillatory cases (i)

(- a -f b) t (- a - b) t Q i = I, I, o [e - e ] 4 (R2C2 - 4LC) and

2 2 ( a + b> t n rRC M(R C - 4LC) - e [^..,Q q= —...^ ° 2^ (R2C2 - 4LC) RC - ^(R2C2 ------. -:i————Lui.mîûîiiiiiiiiiiiino™™ ______\ (R2 2- C4LC2 )

144 R wher ™= d ™a ean

For case (ii), above solution valie sar d. 0 wit = hb

Oscillatory case 2 Q.e~at t (3 n Si " = i 4 (4L - CR 2C2)

* e C ^L Q 2 and q = ° Sin (,

where 2

LC 4L2

- R2C2) RC

Design considerations:

fase Th t ban charges ki highea o dt r voltage compareo t d

the slow bank closinn O . ignitroe gth n switch e fasth ,t

bank discharges wit tims hit e constant s tilvoltagit l e

falls to that of the slow bank which then starts

discharging and maintains current for a longer duration.

The acceptable dro capaciton pi r voltag oveV eA timra t eA

and the value of bank capacitance C are related as

C =

V Vr*L1" J *"• '

145 For 10% drop

C - T,At

Present status : banH J kd ratinan F TFV , g

Name______Voltage____Capacitanc n millifarai e d TF______1200V______500 VF 250V 800______JH0 40 750V

JH and VF banks are ready while TF bank development is

going on. Capacitors used :

: Rescon (IndiaMake )

Value : 10 millifarad/250V Leakage current : 2 mA

Power supplies : Full wave bridge Transformer Phase Voltage KVA TF 3 22*0 600V?12QOV 25 each phase VF 1 0 25022 V 1.5 JH 1 220 750V 7.5

Control circuit :

It ensures that the dump switch is liftted before the

SB can be charged or discharged. A few other interlocks

e okayear d befor e operatioth e f o e charginnb n ca g

initiated. The scheme is shown as follows.

146 CHARGE

MCB DN ± INTERLOCK]

LIFT DUMP INDICATE 'f DUMP SV i J ni/ '? DUMP LJIN FAULT \ f INDICATE YE s DUMP SWIT- ^ 2" CH DK •" 1 i AMCl O I DN >/ rsID INDICATE MCI ^» I I O X DK? + Vc ap_V 'pres»(ta. ^ K n Jfc VI~C F MCOF I INDICATE CHARGE MC2 DN READY i INDICATE ND ? K D 2 MC INDICATE MC2 FAULT YES, MC2DN

A V QPTOTRIAC OFF ? N ND D F H

V. prese p ca t.

DUMP CLOSE

147 Result :

Shot no1. 23593 Date' sept2 2 . , 1995 fflEBEtMiaa Tiwe s scale'5u = 1 . 71 '""" 1POJH) 5v

Fast Bank Slow . BankA k2 3 = I P

TF = 5 KV JH = 255 V 10.= t 5A BBS

V 2 9 = F V V K 4 1. = H J

V K 4 1. V F=

-1 12? 7

148 XA9745752 JUST: JOINT UPGRADED SPHERICAL TOKAMAK

E.A. AZIZOV, N. Ya. DVORKIN1, O.G. FILATOV2, G.P. GARDYMOV1, V.E. GOLANT3, V.A. GLUKHIKH2, R.R. KHAYRUTDINOV, V.A. KRYLOV2, I.N. LEYKIN1, V.E. LYKASH, A.B. MINEEV2, G.E. NOTKIN4, A.R. POLEVOY3, K.G. SHAKHOVETS3, S.V. TSAUN4, E.P. VELIKHOV4, N.I. VINOGRADOV3 Troitsk Institut Innovatior efo Thermonuclead nan r Research (TRINITI), Moscow, Russian Federation

Abstract The main goals, ideas and the programme of JUST, spherical tokamak (ST) for the plasma burn investigation, are presented. The place and prospects of JUST in thermonuclear investigations are discussed.

Introduction At present in the world about ten devices (HSE, ROTOMAK, FBX-II, SPHEX, START, TS-3, HIT, CDX-U), named spherical tokamaks (ST) are in operation. Impressive succes f thio s s direction, specially results, achieven o d START device potentiad an , l advantage low-aspecf so t tokamaks have resulten di developing of work on creation of devices of the following generation. At present, ST with plasma current of scale 1 MA (GLOBUS-M, MAST, FAT, NSTX, USTX) are being developed and constructed. Fast dynamics of ST direction development, low in comparison with usual tokamaks of significant limiting factors (capital costs, required power consumption, time duratio buildr nfo ) make urgent developmen creatiod an t f no devic reactof eo r scale wit hwhicn plasmo , h breakeveMA a 0 curren1 ~ d p I tnan thermonuclear burning modes can be achieved. It is necessary to note the following important circumctances: resultse 1Th . , which wile followin obtainee th b l n o existinn d gdo an T gS generation of devices, such as GLOBUS-M, MAST, STX and etc., can allow to specify modes of operations of JUST device, to change a features of JUST design, and to work out the research programme in detail. 2. When designing numbea , JUSf o r T systems shoul executee db d as"base" or "modular". It means, that used decisions and construction's principles can be hereinafter base for the subsequent modernization of JUST device and future experimental ST-reactor concept creation. 3. Expenditures of the structure of JUST complex can be essentially reduced while creating it in scientific centres with already well advanced infrastructure. e uniquTh e situatio thin i n s respec f Troitso s i t k Institut r Innovation'fo e & s Thermonuclear Researches (TRINITI), there power site of above 300 MW capacity experimentan a , l hall with biological shield, tritium system designer dfo work with tritium mass up to 10 g and developed infrastructure are existing now.

'State Enterprise "Lenmgradsky Severny Zavod", Russian Federation Efremov Institute of Electrophysical Apparatus, Russian Federation A F loffe Physical-Technical Institute, Russian Federation RNC "Kurchatov Institute", Russian Federation

149 Parametrical analysis

The level of approaching to the reactor parameters can be characterized by

o facto= PpUS/PAU Q r X- rati f thermonucleao o r power PfUSt plasma heating power PAUX-

In the deuterium-tritium experiments on the largest existing tokamaks with strong auxiliary plasma heatin achievee gth d valu f Q-factoeo abous i r t 0.3e se , Table 1. Table 1 Value f Q-factoso r achieve experimente th n di s TFTR /!/ JET 111 PAUX, MET 39:5 15.3 PFUS, MBx 10.7 1.8 --> 4.5* 0 0.27 0.12 --> 0.3* # - recalculation from 10 % tritium contain (experiment) to 50 %.

So the transition to the regimes with Q « 1 (breakeven) and specially with Q = 2-5 (regime close to self-maintaining burn) is the task of the next generation of tokamaks.

The main goal of parametrical analysis of JUST is determination of "pay" for the achieving the range Q > 1 in spherical tokamak configurations. PARVNS uses dcodwa unde/ e/3 r analysis.

1 2 3

Fig. 1. Parameter Q = PFUS/PAUX influence on tokamak power supply at different value aspecf so tplasmd ratian oA a major radiu W, sR Z.s 1?

150 MA

1.5

1.0

0.5

Q 0 1 2 Fig.2. Parameter Q influence on plasma current Ip , toroidal field on the axis Bto and averaged neutron wall flux PN - R = 1.7 m, A = 1.5, k = 2.5, PN = 3

-2.

1.0

0.5

Q 3 5

Fig.3. Paramete influencrQ normalizen o e d beta value neutrod an n fluN xp R = 1.7 M, A = 1.5, k = 2.5, Bto =1-34 T

151 Calculations show, that reductio powef no r site consumptio correspondE nP o st s increasin e normalizekg 5 decreasin 9 th d q s pi an d N gan N dg(P plasma beta, plasme th k9 s 5i a elongation safete th s yi 5 factor)9 q , d uppee an Th . N rP limi f o t k95 was fixed at the levels PN = 3 and k95=2.5 . The dependence Q95(A) of minimum safety factor value from the aspect ratio corresponds to the Hender approach /4/. The value of auxiliary heating power PAUX is 15 MW. It is assumed that central inducto totae th f l o flu r% xgive 0 swing6 s~ .

The dependence of power site consumption PE from Q, A and major plasma radius R is shown in Fig.l. Very high value of Pz at low aspect ratio A < 1.3 corresponds to the strong increasing of: power heating of the central inductor and problems with placing , mechanical strength and heat removal from the central kern (central inducto inned coils F necessars i T an t re I . H parth r O f share o ty th p decreasin coilH O s f gmagnetio aspec w decreasine clo th t flur ta ratioxfo 2 P f go . The main task of the flux creation in this case can be transferred on the external poloidal field coils. But attempts to solve this task (at 15-20 % level of OH coils flux) come up against the difficulties with vertical and MHD plasma stability durin e currenth g t rise phase , undeSo . r JUST design s considerei t i , d that approximately equa l e totainputh lo t tflu x fro coilH externad mO an s l poloidal coils is reasonable.

Aspect ratio and plasma major radius increasing also help to decrease P£ and problems wit centrae hth l kern.

But on the way of aspect ratio increasing some important advantages of low aspect tokamakloste b .o t Namel e sar medium-to-nigr yfo h aspec- 8 t1. rati> oA difficultie1 ~ Q r s fo wit d h2an plasma heatin oc-particley gb s appear n thi(i s s cas- minima p eI l current, necessar confinement)e th r yfo importans i t I . noto t t e that with aspect ratio decreasing natural plasma elongation increases and configuration with "natural" divertor is formed.

This consideration results the selection of the aspect ratio value A = 1.5. Analysis show s necessarthai t e i t th r o hav yt fo e major th em 7 1. radiu « R s achieving of Q = 3 at the level of Ps ~ 300 MW. e dependencTh e plasmth e f froao currenmtoroidae Q th , Ip t l magnetic field e plasm e averagth th t a d a ean axi t neutroB s e firsnth t flu n walo x l (durine gth plasma burn) pn is shown in Fig.2 for fixed values of A = 1.5, k^ =2.5, q95 « 3, PN = 3 and R = 1.7.

In the range PN > 4, further increasing of Q can be achieved up to transition to e self-maintaininth g regime e respectivTh . e dependenc t a fixef Q(pno e )d toroidal magneti shows i c T Fig.3n fiel1.3= ni 4t B d . resulte th conceptuae n I th , e showar 3 nl2- parameter~ Q , 3 ~ f JUSN so P r Tfo in Table 2.

152 Table 2 JUST preliminary parameters Plasma major radius] [m R , 1.7 Plasma aspect ratioA , 1.5 Plasma minor radius, a [m] 1.13 Plasma elongation5 k£ , 2.5 Plasma safety factor5 09 , 3.3 Toroidal magnetic field on axis, Bt [T] 1.34 Toroidal plasma beta, Pt [%] 27 Plasma current, Ip [MA] 14 Auxiliary heating power, Paux [MW] 15 Fusion power, Pftis [MW] 40 Q facto- r 2.7 Averaged neutron flux MW/n p , m 0.22 burn time, Atburn [seel 10 J Plasma density 0 m~[1 e n , ] ~0.6 Plasma temperature [keVT , j 10 Energy confinement time IE , sec 1 Fuel d-t

Some detail f concepso ideologd tan f yJUSo T tokamak

Tokamak JUST was planned as polyfunctional stand for work on: - regimes, making condition breakever sfo n achievement, transition into burning and knowing of the limits of working parameters of installation; - solving of the technical problems (perspective conceptions of divertor, availabl quickld ean y rechangable unit inductof so inned ran rF parT f o t system, divertor plates, fuel cycle); - technological problems (perspective materials, behaviou materiaf ro l propertie t higsa h neutro head nan t fluxes).

followine Th g principle base layee th s ar e s da concep JUSf o t T project: numbeA 1 .f JUSo r T basic system supposes i s mako dt s modulaea r blocks. So, relatively to the divertor, this is understood as reservation of large enough free space inside the vacuum chamber under poloidal divertor and a number of big ports for simple and comfortable manipulations with different perspective divertor equipment, this facilitates utilization of considerable stream energyf so severaa r Fo . l periodically changable JUST systems, sucs ha central inductor and inner part of TF system, modular principle means both possibilit f changino y d necessaran g y modernisatio f somo n e constructive element installatione th f so least placine A . th t strengthenind gan outee th f gro coils of poloidal system must admit the changing of position for the optimization of scenario of discharge. At that unchangable elements of tokamak's construction are:

153 e maith e vacuun- th parf o mt chamber with port r auxiliarfo s y heating systems, diagnostic serviced san ; - outer part of toroidal system. In this mean modular constructioe th e bas r th fo eappeay e b ma n o t r following development of the installation. 2. For the most of the JUST regimes of work the consecutive division on base regime (or standart regime, to some extend) and advance regime are carried out. Base is the regime with Ip ~ 8 MA, normalized beta pN « 3, Q - 1, standart configuratio f divertoro n , puls e, energ s duratio 3 - y 1 confinemenn t time satisfy to H- mode (TE - TEH)- Advanced is the regime with IP - 12-14 MA, PN > 4, Q > 2-3, divertor configuration which simplify the heat removal (natural divertor etc.), pulse duratio o 10-1t p u n 5, energs y confinement time satisf VH-modo t y - E (T e

3. It is supposed the maximum using (during JUST design) the newest elaboration technologiesd san , develope Russiay db n industry.

Breakdown plasme th d aan ! starp u t

Plasma is accepted to be form near the outboard of the vacuum chamber and smale th s (selm iha initiat5 e 0. Fig.4) < l siz0 a .e2 Realisation of the breakdown by vortical electric field meet with difficulties becaus e stronth f eo strage limith yn fielo t 5-1< ( d 0 Gauss securind )an e gth plasma position localization.

Estimations show thalevee th straf t o l yplasme fielth n di a regio JUSf no T aciev Gauss0 5 e thin I . s casmors i t ei e reasonabl creato t e forplasme efh a with assistanceR EC breakdowe Th . realizee b givee n th nca n d i volume wite hth pressure of the neutral gas ~ 10~5 torr by cyclotron resonance of electrons. Interaction of vertical magnetic field with Pfirsch-Schlutter current secure the plasma equilibrium along the major radius.

During ECR assist discharge the increasing of the plasma density changes the effectivity of the cyclotron absorbtion. The cyclotron mechanism of electron's acceleration stop its work for the plasma density ne >0.01-ncr because of the plasma influence on the wave polarization. It is replaced by another mechanism of absorbtio transformatioo t e ndu f faso nt electromagnetic wave inte slooth w plasma wave, whic intensivels hi y fading. Effectivit f absorbtionyo , associated with the process of the wave transformation near the upper hybrid resonance, is closely to unit for the plasma concentration up to critical one ne - ncr if the wave doesn't reflected on the way to resonance. In the result, the aggregate of this two resonant mechanism absorbtiof so f electromagnetino c waves (cyclotron typd ean upper hybrid one) ensures the gas breakdown at low neutral gas pressure in the require assisR d volumEC t nplasme wito < t crford th e e ,hn f ewhican m o du a h fill up the region with coo >o>Be (o»o is the wave circular frequency, coBe is electron Larmour frequency).

154 Bondage

5? Plasma ¥£& Plasma

rvi 1m 0

Fig.4. JUST conceptual design. Cross section.

155 s breakdowga e e th th n JUSr i n Fo m T1 tokama3. e regio= th C n R i kn hyrotron frequency is 20 GHz ( AO « 1.5 sm). The source of ECR power heating - 50-10 witW 0hk time duration 10-2 necessaryjbs i s 0m creatioe th r plasmf no a 11 3 with ne ~ S'10 sm" , Te ~ 50 eV in the volume ~ 10 sm . Plasma, appropriat scenarir efo o calculations, with closed magnetic surfaces i s formed, when plasma current achiev 30-5~ , verticap eI 0kA l magnetic- fiel30 - d 50 Gaus minod san r plasm. a m radiu 2 0. s~

Discharge scenario

Standart variant of the plasma scenario during the current ramp-up phase with usin f Ejimo g p ^ -I alead =C R o formul- AT resistivt Ejirnif s -u / /5 ae flux consumption at the level of 10 Vs. In one's turn this allow to achieve the maximum plasma current only about 8 MA. This value of Ip turn out to be near the current Imjn , which is necessary for the fast a-particles confinement. It is necessary to note, that in the standart variant of scenario, the velosity of current rise dlp/d s limitei t e plasmspeey th db f do a current penetratio d doet an nno s exceed, as a rule, the value 1 MA/s.

In advanced varian scenariof o t supposes i t i , increaso dt levee e th dlp/d lo t p u t 10 MA/s due to effect of current form layers and to decrease resistive flux down e leveth le developmen 1.5-Th . 2Vs f suco t h e varianth f o f scenari o te on s i o principal poin JUSf o t T work necessars i t I . solvo yt e several tasks: - choice the poloidal magnetic system (position and current scenario) witwhole e satisfactioth th h r e fo scenari " f conditio2 o n > f 5 oo Q9 n nominae th o t currenp lu ) tkA breakdowf riso 0 5 ed -fro= en p e m(I th n current; - definition of the maximum plasma current for given inductor construction; - choic e geometrth e d material yan e passivth f so e stabilisation system, which secur plasme eth a vertical stability.

DIN A code /6/ is used for the plasma parameters and dynamics calculations e frefoth re boundary propoundin axisymmetricaD 2 n gi l geometry t giveI . e th s following results:

e choseth n I 1n. geometr e poloidath f yo l syste s possibli mt i achievo t e e eth plasma current rise up to the 14 MA. Under this the DN plasma configuration is created with parameters: R = 1.7 m, A = 1.5, k95 = 2.5, 6 = 0.3, lj = 0.5, P Pe plasm th =0.1 s i a6 4( triangularity s internai j l , l inductancee th s i p p , plasma poloidal beta) e resistivTh . bure eth flu nn o xphas t emor musno ee b t than 2 Vs. e maximu2Th . m e currenvaluth f eo poloidan i t l field coils doe t exceeno s d8 MA e maximuth ; m currene centrath n i t l inducto 0 MAturns6 s i r e Th . maximum field in the inner part of the central inductor is lower than 13 T. passive 3Th . e stabilisation structur securn eca characteristie eth c tim verticaf eo l plasma position instabilit levee th lt y a 10-10 . 0ms

156 a

- (sU o S o

(U 6 Z I 0 T- 2- C- -^ o S C3 JS O "

ID « oo 00

_O ^^

o 8 O .i ea §5 E -o 6 2 I 0 I- 2- 6- t-

-B C\2 KS

"S W3 CO 1.8 v~4 II ob

o II I 0 2- B- t-

157 The developing of the plasma scenario allows to closer define the poloidal coils position and form, material and details of construction of the vacuum chamber passive th f o e d stabilisatioan n structure. Some characteristic pointplasme th f so a scenari showe oar Fig.5n ni .

Plasma confinement

plasme th r aFo parameter JUSf so T (Table resulth f calculatio) o te2 e th f no necessary values of the confinement improvement factor H in relation to the known L-mode scalings (see /5/) is shown in Table 3.

Table 3 Valueconfinemene th f so t improvement factoH r Scaling ITER-P-89 Kayej-Goldston Lackner-Gottardi H 2.5 1.6 3.4 H* 2 1.3 2.6

valuundee H th f conditione s o ei th r - Her H e , that a-particles heating does t wors no plasme eth a energy confinement.

s interestini t I noteo gt , that during las e DIII-t th tim n eo D tokamak, with dimensions and magnetic field values near to JUST ones, the now scaling was acieveintegrae th r dfo l type parameter p'tE (see examplr ,fo e /?/):

22 4 S R (P 10-• 5 • *

where S=Ip'qx|//a'Bt calleo s s oi d shaping. For the JUST parameters S is «32-38 (if q^ ~ 3.3-4), so there is the possibility to achieve P'T£ ~ 0.2-0.28. This product is near to the data of Table 2.

Plasma current drive

The current drive effectivity is characterised by the value I -n-R p [MA, 1020 m"3, m, keV, MW] T-P,AUX For the neutral beam, ECR or FMW current drive cases the value of n = 0.025- 0.03. That gives for the parameters of the Table 2 the necessary value of current drive power about PCD ~ 30 MW (small value of bootstrap current - 0.2 is taken into account) s possibli t I . o decreast e e this valu o somt f poweo ee edu r

158 increasing of the ratio T/n. So if n ~ 0.4' 102 0 m" 3 and T ~ 16 keV, the value PCD decreases down to ~ 15 MW.

Conceptual analysis of the device design

Major elements of a device design namely vacuum chamber and electromagnetic system consistin f toroidago l (TF poloidad )an l (PF) systeme sar considered f possiblbelowo e e sketcon Th .e f ho variant a genera f o s l desigf o n JUST tokamak is represented on fig 4.

1. Toroidal system.

The design of a toroidal system should answer the following requirements: - coiltoroidaa f o s l field shouild creat= 1.5...1.5 y 1. radiun B e o = 7T R s m; - curren separata n i t e coil shoul; t exceekA d no 0 d25 - the quantity of coils should be not less than 24 (the field ripple must be 0.7% at the vessel wall and the 0.35% at the outer edge of a typical plasma discharge); - duration of a working puls - 20...30 s, temporary interval between working pulset morno s esi tha minutes0 n3 ; - the system of current streams should provide symmetry radial error field 4 relativel equatoriao yt 10'< | j ;B l / plan r B e| - the contact units should provide held down effort not less than 3 M Pa o providt d an e safet f contac o l dynamiyal t a t c effecta toroida n o s l magnetic system; desige toroidath a f - no l system shoul dismountable db provido t d en an ea opportunity of realization of repair jobs; s supposedi t I , thae toroidath t l system will consist fro 8 consistentlm4 y connected turns. A current in it should be 240...250 kA. Each turn (see fig. 6) consists of a central part (1), external part (3) and two compensatory plots (2), serwing for indemnification of temperature expansion of an internal part, which can make size up to 20 mm. e externaTh l tura par f no t represents type-setting water- cooling copper conductor with a contour, close to instantless. Copper conductor is supposed to conclude in an easy steel bearing skeleton. The geometrical parameters of external

parts of a turn should provide the reliable long-term work at current density in conductors 10...15 A/mm, 2 speed of water in channels 5...7 m/s, the heating of copper - up to 25...35°C. A steel skeleton should provide nessesary rigidity and strength of a design at all acting of electromagnetic, thermal and weight loads, convenienc a device f eo , opportunit f independenyo t fastenin f eaco g h external part of a turn on the general base of frame, connection of all external parts of turnunifora n i s m rigid system wit e hel hth f demountabl po e horizontal straps and top bandage. With the purpose of increasing transparency of the device it is suppose grouo dt p external part turnf so pairsn si .

159 3

To previous turn

Fig.6 systeF T , m turn.

160 Outer TF OH Wedge clamp

Banding

Expansion compensator

Fig.7 externan A . l branc internad han ltur F rodT nf eo connection .

161 In this case, distance between turns in an equatorial plane of the device will mak centraeA abou. m l 1 tura part f no t looks like itself type-setting spiral-like conductor, made from copper cooled rods. The internal parts of 48 turns of toroidal system will form an external part of central pop. A winding coiner of an internal part of each turn - 7.5°, i,e. corresponds to surrounding step of turns. Consecutive connection of turns (see fig. 6) without using of the conventional circuit with return turd thuan ns easily execute a dhig h degre f symmetro e y relativel equatoriao yt l plan radiaf eo l component errof so r field achieveds i s . The electrical connection of external branches with internal rods can be executed with the help of force contact on advanced contact surfaces with using a load-bearing bandage from composyte material and wedge clamps (see fig. 7). Use of water cooling is supposed for both: for internal, the most energy-tensed parts of turns, as well as for external D-shaped sites. The preliminary estimated accoun cooline th f o t g syste mads mlimitinr wa efo g cas e- stationar y mode.

Geometrical parameter internae on tota e f so th ; l rodarem lengtf 9 e ao :th = hL

2 cross-section current conducting elements - Sj = 2.44-10" 3 m; current density - j = 98.2 A/mm2; summary area of cooling channels -So — 0.566 1O3 m2 ; arrangement of channels of relative conductor - longitudinal; direction of current of water in adjacent channels - is opposite. Released heat power (on one central part of a turn) is approximatlely 3600 kW. At the mass charge of water - M = 8.5 kg/s; water- flow m/s5 speed1 ;a = pressurf d V - an a e differencMP 8 = P A e temperature drop on a border "water- copper" - AT — 60°C a film heat transfer coefficient will make ~ 37 kW/m2 C. At this parameters from condition of thermal balance follows, that operating temperature of a centre part of a turn in a stationary mode shoul t exceedno d 150°C. e externaTh l part turnf so s coolin t difficultno s gi , because current density there is rather low (~10...15 A/mm2). However, the compulsory cooling using permits essentially to reduce their dimentions, and so to increase "accessibility" devicee ofth .

. Poloida2 l magnetic system.

consistF P f threo s e subsystems: system f excitatioo s maintenancd nan f eo plasma curren ; system P I r positionin- t fo s e plasmgth a radiall verticalld yan y and for determing the plasma shape. Mos f difficultieo t s while spherical tokamak developing, connected with ohmic heating system creating. Therefor necessars i t i e dwelo yt thin o l s question in more details. In considered device, the zone for ohmic heating solenoid accomodation is limited in radius of 460 mm, and it must create maximum flux swing 10... 12 V-s. e mosTh t perspective solenoiH varianO f o td desig s considerei n e th d following foura : - layers inductor, with each layer made from dual continuous copper conductors with radial gaps between them and axial stream of cooling water on these gaps (see fig. 8). The difficulties, arising when realization of specified variant, first of all are as follows: availability of electrical contact of

162 Dual copper conductor

Fig.8 solenoiH O . d conceptual design.

163 copper wit a coolinh g liquid; necessit f providino y a gsmoothnes n axiaa f lo s channels walls, excluding cavitation watee th ; r deliver systep ta ym& complexity. e morTh e detailed studa desig f d formulatioo yan n e technicath f o n l requirement o poloidat s l syste n completmi e volum s possibli e e after testing accounts of plasma magnetic configurations equilibrium, according to the plasma scenarios.

3.Vacuum vessel.

The engineering features incorparated into JUST device should allow flexibility and efficiency in addressing the tokamak program gols. At the heart of the JUST - is D - shaped vacuum vessel. It should be designe a continuou s da s resistive shell o avoit , d problems inheren electrican i t l breaks or bellows - type coustruction. desige vacuua Th f no mvessea l should answe followine rth g requirements: - used materials, the methods of design elements connection and internal surface conditions should provid a basie c pressur e vesseth n i le less than ; 10-Pa 7 - toroidal resistance mus e sufficiencb t y hig o hfacilitatt ) (0, Q e1m plasma initiation; - constant tima vesse f o e l mus e enougb t r maintenancfo h f verticao e l plasma stabilit / R~3 ) L y( s 0m - the mechanical strength and stability of vacuum vessel should be saved t effeca f atmospherio t c pressure mechanicad an , l loads, arisin plasma t ga a cuurent disruption; firse th t - f walo chambel r should thermaa carr n o y l poweo t p u r 0.3 MW/m divertod 2an , r2... o targett p 4u MW/m s- ; 2 desige th - n should provide preparate technologica; °C 0 l heatin20 o t p gu - diagnosti d technologicaan c l port n equatoriai s l plane should provide radial, vertica d toroidaan l l e plasmviewth d vessef o san a l interiod an r connectio f additionano l heating system wels sa s acces a n l a stafa f o t so f interial element chamberf so ; - the chamber should have two toroidal divertor zones with following dimentions: heigh - 1...1.t , widt5m - h1...1. wit2m h port r divertofo s r targets service and for pumping. A demountable vacuum chamber creation is allowed. It's general view is shown on fig. 9. It must be designed as a modular construction one, and will be consist from the following main parts (see fig. 10): central part (1), divertor module (2), external part (3) diagnostid ,an c centrae portTh s. bell) par(4 t s i t supposed to be executed from stainless steel of minimum thickness (2...3 mm) to increase an ohmic resistance of a chamber. Whereas the central part of a chamber adjoin o centrat s l a toroidapar f o t l system (becaus f necissito e o prividt y e required aspect ratio) thermaa , l lengthenin f vacuugo m chamber's central part during t operatini possibles gi s indemnificatio it r Fo . zona n f centrai neo l part d divertoan r module connection,a axian i sn radia i s wel a ls a ll directionsn a , equalisers (5 ) in a king of ring-shape goffered membranes are allowed. Divertor module is supposed to be executed from sections, representing threestrate cellular designs ( fig. 11 ) 50...80 mm thickness. It should have elements for fastening

164 Fig.9. JUST vacuum chamber. General view.

165 Fig. 10. JUST vacuum chamber. Main elements.

166 Outer casing

Honeycomb structure

Inner casing

FigThreestrat. 11 . e panel.

167 replaceable divertor targets divertoA . r module's volum defines i e proceediny b d g from a required divertor targets configuration and necessity to arrange them on certain distance from plasma surface e divertoTh . r module shell shouln a e b d additional elemen f regidito t a rin s gya n externa a edg r fo e l a chamberpar f o t . Divertor module should have 8..pumpin2 1 . g branchpipes (pos , fig.6 . 10)e Th . external part of vacuum chamber (pos. 2, fig. 10) should be executed from the threestrate cellular panels 80 mm thickness as well. A soldered panels with outer and inner casings from stainless steel and honeycomb structure from steel poil between the e expectedmar e diagnostiTh . c ports bel assumes i t a thinshee s da t edged design. To maintain required ohmical resistance a number of the constructive decisions in particular assembly of elements of the construction parts, separated by dielectric linings is possible.The essential significance also has application in desig f cellulano r panels, whic f havinhi g enough large bearing abilities have high ohmical resistanc botn ei h meridianal, toroidal directions offeree th , dSo . cellular panald witan f hgenera o sm externad m thicknes an d l0 interna8 an lf o s l environments till lmm,of bearing ability under external pressure are equivalent to a continuous steel wall of thickness 30-35 mm, and as for ohmical vesistance to tree mm environment. The design of a chamber should provide an opportunity of installation from the external party of blocks of neutron protection or blanket. On the whole the performed calculations show, that at given parameters of elektromagnetic system the operation is of its most responsible of elements of it will be executed when having significances, of density of current, temperature of materials, speeds, quantit d distillatyan e pressur channeln i e f coolingo s t wilI . l require during development and realization of additional researches on breadboard models and styles with the purpose of refinement of really permittable borders of specified parameters. Thus Russian experience in creation of super- power the electrogenerators and domestic airspace technologies will be maximally used.

Divertor

In JUST tokama concepe kth higf o t h enough divertor volum accepteds ei . This allow to check a number of different divertor systems. The divertor sizes in vertical directions are characterized by value hx-p - distance between X-point and the divertor plates for the maximum plasma current. This value for JUST and for large existing tokamak DIII-d an T showDs i sJE Tabln ni . e4

Table 4 X-poin diverto- t r plate distanc vertican ei l direction Tokamak DIII-D JET JUST hX-Pm , 0.3 7 0.0. 6- 1.35 1. - It can be seen that hx-p for JUST case is much more than for DIII-D tokamak (with the same value of major radius) and is about twice larger in comparison with JET (R = 3 m). Besides, the length of the magnetic field lines

168 (plasm - adivertor ) increases when aspect ratio decreases /?/ lase .tTh argument also simplify the decision of the heat removal problem. e valu Th f poweo e r move o eact d h divertor zon f JUSo eN plasm (D T a configuraton) is estimated as 6 MW. So for the standart variant of divertor the 2 specific heat flux pn reacdi ca v0 MW/m h2 f plat(i s placei e d normae th o t l separatrix) and 5-7 MW/m2 - for angle inclination 15 - 20 °.

configuratioe Inth f "naturalno " divertor levee th ,scrapf o l f layeeo r thickness increasing can be 5 -10 near the plate. So pdiv can be significantly decreased down to - 2-4 MW/m2 even for the normal flow relatively to the divertor plate. There is the possibility of further relaxation of the specific heat flux due to neutral gas presence in the divertor volume or by inclination of divertor plates.

e result Is nallowablth i t i , checo et JUSn ki T divertor system r verfo y s a s intensive heat fluxes - 20 MW/m (standart divertor, plate is placed normal to the separatrix) r medium-to-lofo s a , w hea MW/m4 t - fluxe 2 < s2 2(natural divertor, inclined plates).

Reactor aspects of JUST

At the beginning some notes about the place of JUST in the thermonuclear programme. JUST is large, principal (may be main) step in the ST investigations. During the JUST work it is supposed to give the decisive answer in relation to limit f workino s g regime (plasmT S f o s a confinemen , plasmt a beta, fluxes)n I . e morth e broad plan JUST represent polyfunctiona lworkine stanth r dfo g out: different scenarios and regimes of plasma burn; perspective divertor systems and material r fusion s plannedfo si t I . , under JUST design, usag a enumbe e th f o r newest, but tested in industry, elaborations and technologies. The peculiarity of JUS minimun i s Ti m cospowed an t r site consumptio e solitioth r f sucnfo no h tasksf wido t ese .

In this respect JUS importann occupn a T ca e unique b yth n tca e t i plac d ean stag n fusioi e n investigations t JUS a large-scalBu .s Ti e sted onlan py after realization it will be possible the assurant extrapolations (specially for plasma beta confinementd an tokamak-reactore th r fo ) . Nevertheless, some opinion about ST-reacto formulatede b n ca r . Todae yth following reactor applications of tokamaks are under actively discussions: 1. Volumetric neutron source for the material testing (note, that certain part of this programme can be, in principle, realized already in JUST); 2. Hybrid (fission-fusion) thermonuclear plant; 3. Installation for nuclear wastes transmutation; Pure thermonuclear plant. There are reasons to suppose, that ST ideology is more adequately transformed inte firsoth t three possibilities. Under this main reactor requirements muse b t taken into account:

169 s necessari t 1I . n comparisoy(i n wit e experimentath h l installation o increast ) e significantlr fo 5 e availabilit0. th y - 3 0. y = e leve coefficiena th K lo t p u t neutron sources and up to Ka = 0.7 - 0.8 - for power plants. So in JUST it is necessary to devote the special attention to the stationary work regime achievement (current drive systems, magnetic system, heat removal systems) o east d fas an d yt an rechangin f somo g e elements (central kern, divertor plates). 2. It is necessary placing between plasma boundary and toroidal field coil - radiatio ne inne th shiel blanked t ra e oute an leg( dth shield ) t ran a t leg- d . The inner shield thickness di is defined by maximum allowable fluence on the

insulatio reasonably b d nan e tim r thiefo s fluence achievingn e timth es I . liff o e 23 internal kern is 2-3 years, djn is 35-40 CM for ceramic case with fluence ~ 10 2 unde\r value th r f specifieo c neutron MW/m1 flu ~ x r A12O2Fo . 3 ceramics it is possible to decrease the value of din down to 0.2 m. Another variant is in refusal from insulation usage in electric kern (single turn coils with parallel feed) make it possible to decrease djn practically down to zero. The blanket shield an d thicknes . t outboarsa m 1 ~ s di 3. Because of blanket and shield presence in the outboard, external poloidal system becam e placeb r froo et e plasma dfa m th . Thi n resulca s expedienn i t t of superconductive external poloidal field coils usage. Besides, this situation (hig p betweega h n plasm e som d coilth n lea an aeo ca )t d simplificatiof o n external poloidal system and to the decreasing of the number of poloidal field coils thin I . s case speciath e l attentio e dischargth r fo n e scenario realisation must be made. f possiblo t se ee parameterTh neutroe th f so n source with neutron wall load- 1 MW/m2 (abou same th t e parameter hybrir fo e dsar power base plantth f eo n o ) JUST approarch are shown in Table 5 in comparison with some other similar projects. At that pos. 1 corresponds to the stand for the material testing with single turn toroidal fiel dpowe T coil S did e nsran =0.0 th plan o y t /4/b 5 m ;- t I I pos. Robinson's ideolog (START-->y/8/ M AST- installatiothe >to 11)- ; nfor Ill pos . the nuclear wastes transmutation by Peng's ideology /?/ and at, least, pos. IV - to the variant of JUST type neutron source with inner shield thickness 20 sm.

Table5 Design parameter somf so reactor T eS s I II III IV Neutron wall load, MW/m2 2 2-4 1 1 Plasma major radius R, m 0.8 0.79 2.5 Plasma minor radius a, m 0.5 0.6 1.67 Aspect ratioA 1.6 1.3 1.32 1.5 Plasma current Ip, MA 9 30 7.5 26 Toroidal magnetic fiel T axin , do sBt 1.7 1.9 Fusion power PpUS> MW 30 3000 32 360 Powe f heating/CDo r W M , 30 24 30

170 Conclusion

Concep f tocamao t ) multifunctionaMA k 0 JUS >1 1.5~ P TI (A , l stanr dfo working out of physical modes (discharge scenarios, mode of burning, limits of working plasma parameters), perspective divertor devices, realization f testo f o s material states si t whilda e having neutro thermad an n l loads. Analysi resultes sha rougn di h parameter unite th f :so R=1.7m =1.5 2.5A = ; k ; ; BTo=1.34T 10...1= , p I ; 4MA = mord . s an , e) 0 5 t 1 bur2. ~ no t p u ( PAU1 > XQ 15...2 ; 0MW Preliminary study of a design of tokamak has been fulfilled. Main ways of its realization are determined. Thus, maximum widely usage of experience of a Russian industry in creation a super-powerful electrogenerators and high technologies of airspace complexes should be applied.

Acknowledgments

A special thank Intelligeno st t Resources Inc. (Fax. 7-812-312-90-29D 3- r fo ) modelin producC PT y gb t PRO/ENGINEER.

References

/ /! K.M.McGuire, H.Alder, P.Allin t al.ge , Physic f Plasmaso s , 1995, v.2, no.6, p.2176 TeamT JE e , NucleaTh / /2 r Fusion, 1992, 7 v.3218 . ,p /3/ A.B.MHHeeB, B.B.cDHJiaTOB, FIpenpHHT HHH3OA Fl-0935, M., !JHHHATOMHHdx>pM, 1995 /4/ T.C. Render Tight Aspect Ratio Tokamak Neutron Source /5/ ITER Physics, ITER Documentation Series No.21, IAEA, Vienna, 1991, p.236 /6/ Khayrutdinov R.R., Lukash V.E., J. Comput. Physics, 1993, v.109, p. 193 I M.PenP g Statu Prospectd san Sphericaf so l Tokamak Research, loffe-Culham- ORNL Symposiu GLOBUS-Mn mo , 16-20 Oct. 1995 . Petersburg,St , Russia /8/D.C.Robinson Small Aspect Ratio Tokamaks, Int. Workshop "Tokamak Concept Improvement", Varenna, Italy, Aug.29-Sept.3, 1994 miniuI I"!"'*--* min 1 ------A NONLOCAL MODE ELECTROR LFO N XA9745753 HEAT TRANSPORT IN TOKAMAKs

A. DAS . KAW,P . DASTGEES , R Institut Plasmr efo a Research, Bhat, Gandhinagar, India

Abstract

Some recent transient experiments on TFTR as well as TEXT tokamaks have indicate nonlocada l respons electroe th n ei n heat conductivity presene W . t n locano l w modelfe a r electrofo s n heat conductivity which qualitatively exhibit the same features as those observed experimentally. We present the stationary equilibrium solutio sucf no h model discusd san s their stability- Fi . nally we present results to show the relevance of such models to explain the experimental result TFTf so TEXTd Ran .

1. Introduction

Some recent experiment TEXn i s d TFTTan R tokamaks have shown1'2 that an injection of impurities at the edge, cools the edge immediately due radiativo t e loss fro impurite mth y atoms t producebu , concomitana s t rise of central temperature. The experimental observation further rules out the possiblit f increaseyo d ohmic heatin e coree suddeth Th t . ga ne th ris n ei central temperature can then only be explained on the basis of a drop in e electrovalue th th f eo n heat conductivit e centreth t a e . yX This turnn i , , suggest nonlocaa s l e temperaturdependencth n o e X f eo e profile.

presene Inth t manuscrip propose w t electromodee a th r lfo n heat diffusivity, which gets influenced by both local as well as global physics. We may thus write , . Xo

The numerator has been chosen to take account of the local contribution arising fro plasme mth a turbulence whic knows hi increaso nt e towarde sth tokamak edge e denominato.Th r contain e terth sm which depende th n o s global temperature profile through the function F; the extent of the nonlo- cality gets determined throug kernee hth x — globa e K( x'}.l Th l dependence has been chosen to be operative instantaneously in the model.

The choice of the funtion 'F' is motivated by an attempt towards modi-

fying an existing local model to obtain a corresponding non-local model. We note that Hinton 3 has proposed a model Xe to provide an explanation of the L-H bifurcations observe Tokamaksn di e obtainemodes b Hi n . ca l d from Eq(l) by the choice K(x — x'} = S(x — x') and

J \dx,dx '

173 Hinton's model is based on neoclassical theory and assumes that the velocity shear suppresses the local fluctuations and is proportional to the square of temperature gradients, i.e. (^f) oc (|^) . Our physical motivation for the nonlocal mode s tha i r lanomalou fo t s transport also feee w ,l that velocity shear at one location can influence the fluctuation level and Xe everywhere. This is because the fluctuations can be highly extended radially(as is expected on the basis of toroidal and turbulent coupling4 and has been observed exper-

imently 5) and/or the fact that the fluctuations from one location can radially propagat otheo et r locations. 6 This propagation can principlen i , e verb , y rapid as compared to the typical diffusion times and is our justification for takinlocan no lg a mode l which instantaneously responds everywheree Th . kernel integrae th denominatoe e functio th d th , f an o l F n expression(lf ro ) has been deliberately introduced, to model these effects. We shall use several r subsequenou powe n i form w F la r f o ts discussions. 2. Stationary States

heae Th t transport equatio slae th b written e nmodei b n s ca nla

dT 3 dT rr +H 2

We choose the thermal diffusivity x *° nave the form

Our model shows that x depends not only on the location x but also de- overale pendth n so l temperature profile (whic soughs hi t fro e solutiomth n of the transport equation). The relevance of the local dependence through xi and the global dependence on the temperature profile through the Kernel K has already been discussed in the last section. In equation(2) 'H' stands for heae th t source(e.g. ohmic heatin tokamakr gfo )choses i ter d mnan uniform for simplicity. We choose to normalize distances by minor radius a ,time by the diffusive time a2 /xo and the temperature by heat input per unit diffusive time (a2/Xo) per particle heae Th .t transport equatio then writtee nca nb , s na

/V1+ 1 dx dt ) ~ (4 +1.0~ An equilibrium temperature profile can be obtained by solving eqn(4) for equatioe solvee Th b . n 0 d nca = numericall - |^ arbitrarr yfo y choic' 'p f eo and complicated functional dependence for the Kernel. However for a simple unifora d an m 2 ever 1 Kerne= — casyp f whereK e( o l equatioe th ) n nca be solved analytically and yields two stationary states as we shall now see. puttind an tim1 e Substitutin gth = ex —) derivativeK( x' , 2 = gp s zerosa , the differential equation for the stationary state is given by,

dx C

174 Note tha have w t e replace denominatoconstante a , dth (7 y b consida ,x f ro - erable simplification which results fro choice m th unifora f eo m kernee Th . l solutio above th f no e equatio simplgives w i , no d nby s nei an

Here we have made use of the boundary conditions of having |^ = 0 at . 1 = x r fo 0 T(x)d = an 0 x=

The constant C in this case is to be determined by

Substituting the expression for ^-7 we get the following quadratic equation for determining C,

AC2 - C + 1 = 0 (8) where,

x2dx "fJo (1+Xi*2)2 Thus there are two possible values for C implying the existence of two possible equilibrium states. The larger value of C implies a higher temperature at the core region (x = 0); consequently we will call the solution with higher value of C as the hot solution and the one with the smaller root of C as the cool solution. Taking the limit of xi —> 0 we obtain a parabolic profile for the equilibrium temperature.

2 ) (9 rsi(z )= -(l-z ); C in this case is the solution of the equation

——C2- (7+1. 0 0= (10) 3xo

3. Stabilit f Stationaryo y Solutions:

We next examine the question of stability of the above equilibrium solutions. We present an analytical derivation of the stability of the two equlibrium solutions for the case of xi = 0. The linearized perturbation about the stationary temperature profile satisfies the following evolution equation:

2 8*f 8 Tdfst

175 Her value Xsth diffusivits f i teo stationare th r yfo y temperature distribution and y1 (df\ = xst I * -=r 1dx Jo \dx )

is the linearized change in diffusivity due to T(x,t). We assum time eth e dependenc fore th mf eo

which reduces eqn(ll followine th o t ) g ordinary differential equation

\° Jo Xst Xst Ox* e eigeth Her s ni value7 ee determine b whic o t s hi d fro boundare mth y conditions.lt should be noted that equation(12) is an interesting variety of ordinary differential equation righe Th . t han dequatioe sidth f eo n although a const ant (definite integral) can not be treated as an inhomogenous term of the differential equation, as it explicitly depends on the overall structure of the temperature eigen function. Neither does the differential equation come unde e clas th rf integr o s o differential equation e terth m s a sinvolvin e gth integra definita s i l e integral. followine th t ge e g w solutio f"(z0 e th — r ) ,x t n fo t Applyina 0 = ~ g|

a fl T(x,t) cosh(/?z)-——= —/ xsmh(/3x)dx (13) Xst/JJo where

Xst

The application of second boundary condition viz T = 0 at x = 1 yields eigee th n value conditios na

o (14)

The roots of the above equation yield the value for f3. It is clear from the plot zere oth fo contou reae imaginard th an lf o r y functioe partth f so n givey nb Eqn(14) that both of them are simultaneously zero for real values of /3 in the case of cool solution and purely imaginary values of 0 for hot solution(Figl).

176 10T

7.5

-7.5

-10 -7.5 -2.5 2.5 7.5

Fig(l : )Zer o contour plot f real(dottedo s imaginary(boldd an ) ) parf o t / plane/? , R . P Not e intersectioe t casLHth e th ho f eqn(14n ei So e th r fo )n / finite/? d occur ,a —an f showin r 0 sfo g tha t solutioho t unstables ni .

Since

t solutionho r fo It0 . indicateThi> 7 coosr d impliefo an ls 0 tha< s7 t that the cool solution is stable whereas the hot solution is unstable to temperature perturbations.

Numerical solution of the evolution equation with an arbitrary initial profile shows tha t alwayi t s collapses toward e cooth s l stationary solution thereby confirmin r analytiou g c conclusions have W .e verified numerically that even for cases with finite values of xi as well as for stationary states with cylindrical symmetry r conclusioou , n regardin stabilite gth equilibf yo - rium states remains unaltered.

4. Respons Edgo et e Temperature Perturbations:

We next consider the response of the equilibrium temperature profile to the injection of a cool pulse. The evolution equation can be written as

(15)

177 Here Ledge®(t)Q(tp ~ 0 models the injection of cool square pulse at time t — 0, which lasts upto t = tp and is operative only at the edge. Different pulse shapes have also been considered for our numerical calculations. The numerical solution of Eqn(15) shows an appreciable rise of central temperatur t solutionsho e wite e cooline case th edge hth f th Th eth o . n ei f go t dramatino effec s wa coor t cfo l solutions. The heating at the core along with the cooling at the edge is not difficult to understand in the context of our model. The value of x nas an inverse dependenc integrae temperaturth e n th eo f o l e gradient suddee Th . n cooling at the edge causes an increase in the value of temperature gradient locally, thereby reducing x everywhere through the global function in the denomi- nator. This then is responsible for a rise in the core temperature. Since the hot equilibrium solutions were inherently unstable we have repeated our studie another sfo r model wher choose ew —eF ^ (f|) . Thi s choicf eo F permits only one stationary solution whose stability was confirmed numerically. We see (Fig 2) an appreciable immediate rise of central temperature with an injection of cool pulse. We also note that core heating persists for a while even after switching off the pulse. These features are also observed in the experiments. It shoul mentionee db d here that core heating artificialln case a th f eo r fo , y imposed boundary conditio fixef no d temperature(s edgha e s th e(a t a ) 0 T= been considered in all the studies reported here) is very sensitive to the kind of cooprocese th l n puls i f incorporatin so w e no tha choosese e on ar t e W g. sele th f consistent evolutio e boundarth f no y conditio edgee th t .na These results wil presentee b l d elsewhere.

'rOOO.dat' 'r1200.dat' — 'r800.dat' o

0.4 0.6 0.8 1 1.2

Fig(2) : Temperature Vs radius r. The curve rQQQ.dat corresponds to the initial state and rl200.dat corresponds to the evolved state at time 1200 units. The step due to cool pulse at edge at t — 0 is also shown(box)

178 Conclusion

We have propose da nonloca l mode r electrofo l n heat diffusivity whic- hex hibits interestine somth f eo g features observe transienn di t transport experiments at TEXT and TFTR. We have presented the stationary state equilibrium solution havd an s e also discussed their stability responsd an , edgo et e temperature perturbations.

References

1. K. W. Gentle etal Phys. Rev. Lett 74, 3620 (1995). 2. M. W. Kissick etal Nuclear Fusion, Vol.34, 349 No.3(1994)

3. Hinton etal Phys. Fluids B(3), 696 (1991)

4. Kishimoto etal Plasma Phy. &; Controlled Nuclear Fusion Research. IAEA-CN-60-60/D-10 5. R. J. Fonck etal Phys. Rev. Lett 70, 3736 (1993)

. Garbe6X . t etal Nucl. Fusion ) 967(5 , , Vol.31,No.9 (1991)

EXT PAGE(S) Seft S?LA.\"K

179 NONLINEAR SATURATIO E TH F NO XA9745754 RAYLEIGH TAYLOR INSTABILITY

A. DAS, S. MAHAJAN, P. KAW, A. SEN Institut Plasmr efo a Research, Bhat, Gandhinagar, India S. BENKADDA, A. VERGA Equipe Turbulence Plasm 1'URAe ad , Universit Provencee ed , Marseille, France

Abstract

The problem of the nonlinear saturation of the 2 dimensional Rayleigh Taylor instability is re-examined to put various earlier results in a proper perspective. existence Th varieta f eo finaf yo l attributee stateb n differencee sca th o dt s e choicinth f boundareo y condition initiad san l condition earlien si r numerical modelin numericagn studiesow r Ou l. simulations indicatT eR thae th t instability saturates by the self consistent generation of shear flow even in situations (with periodic boundaries) where principlen i , infinitn a , e amount of gravitational tappede energb n yca . Such final e achievestateb n sca r dfo suitable valuePrandte th f o s l number.

I. INTRODUCTION

The Rayleigh Taylor (RT) instability has been extensively studied in various physical contexts ranging from neutral laseo fluidt ] r[1 sablate targetF laboratoro dt IC ] [2 s y [3,4] as well as ionospheric plasmas [5-9]. While its linear characteristics are easy to investigate through analytical calculations, its nonlinear evolution has mostly been studied by means of numerical simulations. Surprisingly, there appears wida e b e varianco t resulte e th th r n ei sfo final nonlinear states obtaine thesn di e studie cleaa d r an sunderstandin r discussiogo n no this problem is lacking. Our aim in this paper is twofold - firstly, to present a brief discussion on the past work in order to understand the differences in the results and attempt to put them in a proper perspective and secondly to present our own numerical investigations of this problem which demonstrat eselfconsistena t saturation arising fro generatioe mth f sheano r flows wile W l. restrict ourselves her dimensionao e onltw o yt instabilitT l R studie e th f so y as three dimensional studies are yet in a very preliminary stage.

181 II. REVIEW OF PAST WORK

A number of numerical as well as analytical works have been carried out in the past to study the interchange instability in varying contexts. The conclusions regarding the final nonlinear states in these various RT related studies, however, seem to vary and can be broadly categorize followss da :

severan 1I . l numerical studies, particularly those associated with periodic boundary conditions, no saturation of this instability has been observed.(e.g. in [10])

othen 2I . r studies, where conducting boundaries have been employed sela , f consistent generation of shear flow has been observed and the instability has been found to saturate.(e.g. in [11])

thirA . d3 clas f studieo s s claim o observt s turbulena e t final state wit ha powe w la r spectrum for the energy distribution of the fluctuations, (e.g. in [10])

The use of spectral codes is fairly widespread in the investigation of the RT instability and hence the choice of periodic boundary conditions is a natural one. Infact, spectral codes automatically enforce such conditions. However such boundary conditions seem quite inappropriat thir efo s problem since the ysituatioe leath o dt n wher fluiea d element leaving the lower boundary with a particular vertical velocity will appear from the upper boundary with the same velocity and fall through the height determined by the box size again and again thereby gainin infinitn ga e amoun f gravitationao t l potential energyt . no Thus s i t i , surprising tha e instabilitth t y continue finae th gro o lt sd statuninterestinn wan a s ei s ga well as an unphysical one comprising entirely of vertical flows. Unless the instability can saturate by some mechanism within the length of one box (and we will soon comment on such a situation) periodic boundary conditions are not the appropriate ones to use for examining the final nonlinear state. The choice of conducting boundary conditions in the direction of the equilibrium gradient (and periodic boundaries in the other direction) avoid the problem associated with an infinite gai f potentiano l energy moss i t I t. convenien finita suc n i e t eha casus differencino et g scheme along the gradient direction and a spectral scheme in the periodic direction. An additional simplification is to take a constant value of (implied by d/dy = 0) and to assum perturbee eth d density valu also et o vanisconductine th t ha g boundariesw fe a n I . studies spectraa , l schem bees eha n employe botn di h direction conductind san g boundaries have bee ndirectioe mockeon n i restrictiny p nb du initiae gth l stat havo et e onlmoded yod s in that direction e characteth e evolutiof I th . f o r n equatio o preservt s suci n s a h e eth symmetr modese th f yo , the initiae nth l conditio boundare th t na preserves yi l timeal t da s appeard an s lik boundarea y condition. Howeve trua t re thi no boundar s si y condition n(i e imposeon contras finita e n th di e o t tdifferenc e code d numericaan ) o tw l resulte th n i s cases sho wdistinca t difference.

182 othee Th r major sourc variatiof eo varioue th n ni s pas t choice studieth n initiaf i eo s si l conditions and selection of the range of physical parameter space for investigation of the instability. Broadl classeo ytw initiaf so l conditions have been considered -(i) random distribution of density and potential fluctuations (ii) states with a few long scale modes. The physical parameter instabilite spacth f eo governee yar followine th y db g relevant quantities: strengte (i th )appropriat e th f ho e gravitational force (e.g. natural gravit ionospherie th n yi c problem, the curvature of the magnetic field for the curvature induced instability etc.) (ii) equilibriue th m density scale length (iii) Diffusivity (iv) Viscosit lengte th ) hy(v scale along periodie th c directio d (vilengte nan th ) h scale alon directioe gth equilibriuf no m gradient. Actually only five of the above six variables are independent when the model equations are suitably renormalised. Numerical studies in the past have spanned over a large range of parameter space and have differed in other important aspects like the boundary and initial conditions. A relevant question to ask then is which are the crucial factors that influence the final state of the RT instability. Partly motivate thiy db s questio partld nan gaio yt n some understandine th f go experimental result steada n so y state toroidal plasma experimen t IPRa t have w , e carried numericaa t ou l simulatio e magnetith f no c curvature drive instabilityT nR wile W l. first discuss this work and its results in the next section before returning to a discussion on the general aspect problee th finae f so th lmn i section .

III. MAGNETIC CURVATURE DRIVEN RT INSTABILITY

The governing equations relevant for the curvature induced R-T instability are easily obtained from the continuity and the momentum equations with the additional assumptions ) quasineutralito(1 f y (ii) cold ions d (iiineglece an ,th ) f paralleo t l dynamics e coupleTh . d set equationf .o writtee b n s ca n a

-^+V (V^+ (I) ~V }-^zxV^-Vn+ DV= 2n at nayn g ay

dt ' 9dy ?V (2)

cHer= /(.Rfig e V gravitationa th cs ,i ) e ionl th drif sf o arisint g throug e curvaturth h e

diamagnetie c= /(Z2 th V terms f) i )d an s c drif sounn t io speed(c e d th speed s s i i R , n n ct s

e majoth r radiue curveth f o sd toroidal machine cyclotron io e th , ns fZi , frequencd an y c

Ln is the equilibrium density scale length). The equilibrium density profile is chosen as

N0 — N00exp( —x/Ln), 4> is the perturbed potential and n is given by n — \n(N/N0).

We have chose normalizo nt equilibriue th 0 eN m densit Ny yb 00 T/e,y (b f , tim fiy d eb c ,an

length by cs/fici.

183 The linear growth rate, obtained from the above two equations, is given by the real part of i if, z-2 -] 1/2 2 g (V 4V n- V + g) D)k (3)

2 1/2 12 2 T k 1 ) (4 - Rea± 7,= l g -k(V^V ynV- V + g )\ z L K 1

which indicates tha l thosal t e wavelengths whic longee har runstablee thaar c nA , where

' (vV —V} J ( n Vg)

e squarTh e showvariableo b e tw n integrald yo satisf t n>ca e an e followinth n sth r y fo s g equations.

2 2 — / n dx dy + (Vn — Vg) I n~ dx dy = —D I (Vn) dx dy (5)

and

) (6 y d x d ) ip V ( I = — a y I <+>-^-d g x V d — / y (Vw d — x d ) ' ^ J dy J J ot

The first equation ( 5), which has been obtained from the continuity equation can be inter- termy d x prete ,d identifie e whic2 b n followss n d a / "pressure a hca e s da Th . " like quantity, change inequalite th o incomine t timn si th e f eydu o outgoind gan boundo tw g e fluxe-th t a s

ariese fluxesTh . ,e term whicth e se nonzer har proportionaar V s d lonoa s an g a V o t l g n n and d(j>jdy are appropriately phased. The second equation ( 6) describes the temporal change totae th f l o kineti c energy whic agais hi nsame finitth r e efo reason , tha , becaus s i t e of the phase difference between n and d^jdy. This equation has been obtained from the evolution of vorticity which essentially represents the quasi-neutrality condition. Vorticity evolve e differenca resul th s f a so t e between diamagneti x boundarie o tw c e fluxeth s t a s arising fro magnetie mth c curvature difference Th . e betwee B fluxenEx s cause chango sn e in vorticit s the ya e sam yar r bot efo h iond electronsan s e seconTh . d equation ca ) 6 ( n also be interpreted in a slightly different way where we rewrite (/ dn/dy) as (— fnd/dy

d interprean verticae th t , x d/dydxdy g ev viewe e change b th n s gravitationadf ca a e o l potential energy. Thus this equatio mera s ni e statemen f kinetio face d potentiam th t can su f thao te th t l energy inviscin a s i d constant e absencinth f diffusioneo . This energy interpretatio howeven nca r provid informatioo en n about the direction of cascade as the potential energy term cannot be cast as the integral of a square of a variable.

184 equationo combinee tw b e y Th sma d resultin followine th n gi g equation,

d /".-,_ x9 vg 21 ff 2 \2 Dv // \2 V ( ) // axa y———+ —/ (Vn) (7 ) dxdy g VgJ — n V J 3 V — n V J Ot Althoug e lefhth t han t doei dt , see 0 sidno s mthi= n ei z / s invariann cas— a s ei D r fo t hav o t obviouy ean s physical significance. Neithe t tel i anythins u ln ca r g abou e poweth t r spectrum as it is the difference between two positive definite quantities. We will now present the numerical solutions of equations (1) and (2) in the next section and then discuss the characteristics of the final saturated state.

IV. NUMERICAL RESULTS AND DISCUSSION

We have use pseudospectrae dth l method modes8 12 ,x studo witt ,8 evolutioe h12 yth n of the two coupled equations ( 1, 2) In order to investigate the dependence on boundary conditions (as discussed in section I), we have made independent simulation runs with two different sets of boundary conditions, namely (a) periodic and (b) conducting in the x direction. In case (b) the conducting boundaries are simulated by restricting the initial condition o contait s n only modes witd symmetryod h l run e madAl ar s. e with random initial conditions. We observe that in both cases, the potential acquires long scale structures whereas the density show predominancsa shorf eo t scalebeginninge th n si . . Eventuall(Figure2) d an s1 y however, both acquire long scale vortex structures. The final state thus comprises of long scale coherent structures. Sinc randoea m initial conditio evolves nha d int overa y coherent state, it makes no difference as to what initial state one chooses. interprey Wema observee th t d behaviou followss a r . Around saturation, equation) (1 s and (2) are essentially decoupled. Thus eqn.(2) becomes like a conventional driven two

dimensional Navier Stokes flow (with Vgdn/dy as a stirring force). This equation is known hav o inviscit o etw d invariants which wele leadth l o knowst n accumulatio powef no t lona r g scales potentiae Th . l thus moves towards long scale t saturationsa densite Th . y equation, on the other hand, is more like the turbulence of a passive scalar spread by a turbulent flow. It has only one square invariant (in the absence of diffusion) and transfers energy to short scales. Thus initially density shows significant short scale structures. Eventually however gravite th y stops pumping energy int systee oth m becaus velocitf eo shore th yo t s shea d an r scale oute sdi . In earlier studies, that are similar to our case (a), Hassam et al had noticed no saturation and had argued that the lack of saturation arose from the use of periodic boundary conditions which artificially made possible the tapping of an infinite amount of potential energy. Our simulations show, however, s possibli tha t i t obtaio t e n saturatio ne periodi eveth n i n c case throug e selfconsistenhth t generatio f sheano r flowe finaTh l. stat obtaie ew s verni y similar to that observed earlier by Finn et al in their simulations, where they had studied

185 A B t=0 t=0 120 100

0 12 0 10 0 8 0 6 0 4 0 2 20 40 60 80 100 120

t=40 t=40

0 12 0 10 0 8 0 6 0 4 0 2 0 12 0 10 0 8 0 6 0 4 0 2 V y

Figure 1: Time evolution of density (A) and potential (B) for n = 0.1, D =

initiae 0.8= Th n . V evolv) l0.03= 0 stated g 0.1 V = 6an , e t s( rapidl y into long scale structures in the linear stage (upto t = 40).

nonlineae th r interactio modew fe a s f usinno gfinita e difference code with real conducting boundaries in the x direction. The crucial parameter in our runs (as compared to Hassam et valudiffusivitye e th th al f s e)i o . When diffusivit mads yi e stronge t leadunsaturatern i a o st d state with entirely vertical flows. In case (b) we have chosen the initial state to have only those modes which have odd symmetry thin I . s case alsr finaoou l state comprise f largo s e scale structures. However, shea thin i r s cas lesss ei , compare case periodif th eo o dt c boundaries, presumably because there is little gravitational energy to gain as the flows are restricted in the vertical direction. Thus only a small amount of shear can saturate the instability. We do observe a power law e spectrafoth r . Howeve e autocorrelatioth r n functio a stron s ha ng oscillatory character,

186 A B t=85 t=85 120 100 80 60 40 20

20 40 60 800 12 10 0 10 12 00 8 0 6 0 4 0 2

t=130 t=130 120 100 80 60 40 20

0 12 0 10 0 8 0 6 0 4 0 2 20 40 60 80 100 120 V y

densite th , y85 retain= t Figurt sA smal: e2 l scale structureth d an ) (A s potential shows coherent long scale structures. At long times (t = 130), the final states for both density and potential show long scale coherence.

which indicates that the final state is coherent and not turbulent. Thus it is incorrect to invoke Kolmogorov's similarity argument explaio st observee nth d power laws. To summarise, our two dimensional numerical simulations indicate that the final state for instabilitT R e th invariabls yi y coherent provide boundare dth y condition value th f ed o san diffusivity and viscosity are chosen appropriately to allow for nonlinear saturation. Within the two dimensional model there is no scope for obtaining a final turbulent state and of explaining powe spectrw la r termn ai f Kolmogorov'so s inertial cascade arguments. Obser- vational data ,e ionospherbotth n hi laborator n i s wel ea s a l y experiments, however point towards a turbulent final state for the RT instabilty. This indicates a serious shortcoming dimensionao tw e oth f l mode d stronglan l y argue a close r fo s r examinatio e threth f eo n dimensional effects that could be crucial for generating turbulence.

187 REFERENCES

. ChandrasekharS . 1 , Hydrodynamic d Hydromagnetican Stability (Oxford University Press, London, 1961) and references therein.

2. W.L. Kruer Physicse Th , of Laser Plasma Interaction (Frontier Physics;73n si , Addison Wesley, California, 1988) and references therein.

3. D. Bora, Phys. Letts A 139 (1989) 308.

4. G. Prasad, D. Bora and Y.C. Saxena, Geophys. Res. Letts 19 (1992); 241; ibid 245.

. M.H5 . Johnso E.Gd nan . Hulbert, Phys. Rev (19509 .7 ) 802.

6. J.W. Dungey, J. Atmos. Terr. Phys. 9 (1956) 304

7. B.B. Balsley . HaerendaG , R.Ad an l . Greenwald . GeophysJ , . Res (19727 7 . ) 5625.

. P.K8 . Chaturved P.Kd an i . Raw, Geophys. Res. Letts (19752 . ) 381.

9. M.K. Hudson, Jnl. of Geophys. Res. 83 (1978) 3189.

10. A. B. Hassam, W. Hall, J.D. Huba and J. Keskinen, J. Geophys. Res. 91 (1986) 13513.

11. J. M. Finn, Phys. Fluids B5 (1993) 415.

188 SHEAR REVERSAL, IMPROVED CONFINEMENT XA9745755 AND ION TEMPERATURE GRADIENT MODE

N . SENS SE . A , Institut Plasmr efo a Research, Bhat, Gandhinagar, India

Abstract

Recent experiments [9] on the TFTR show that while the electron particle diffusivity can be reduced by a factor of ~ 50 to near the neoclassical particle diffusivity level during the reversed shear mode (R/S mode), the ion and electron thermal diffusivity reduce only by a factor of 2-3 and that too not in all discharges. In order to investigate the apparent insesitiveness of the thermal transport towards shear reversal we have carried out a stability analysis of the ion temperature gradient mode with negative magnetic shear l range^pf-,al — r f L>Tiwavelengthn fo f so ) = , 0 i < ff T , S wher1 ( — , g < lon eb e g(b eTi gyroradius)n io e th s i i Te/T,p ,T= intermediat d an ,d an ) 1 e> e (4/b /> 1 /2 founis It d shor tha1). shea the tt> (er be reversa T> ; weaa klhas stabilising influence unstabl e . o) Thinth ' therefors i s^ e mod> r}^= n e ; e(? consisten1 77 . t wite hth TFTR result. Hence, the best way to reduce the thermal transport would be to optimise the equilibrium profiles so as to keep TH < r/f lt and also 5 < 0 in the plasma interior.

. IntroductioI n

A new plasma configuration has recently been proposed for improved plasma performance in advanced tokamak experiments [1]. The principal feature of this configuration is a region of negative magnetic shear (dq/dV < 0, where q is the safety factor and V is the poloidal flux label) ove centrae th r l regio plasmaf no . Recen] t [2 experiment T JE e th n si and in the DIII-D [3] have indeed produced such reversal in shear in the plasma interior. JETe th n ,I this regim bees eha n obtained with pellet injectio nn mode P i (th d PE e an ) DIII-the rampinDby plasmgthe a elongation. Tha negativa t e magnetic shea completeln ca r mode D ideao yo stabiliss MH lsi = n e eth know r sometimnfo . Recently5] , e[4 , ther s beeeha ngrowina g interes stude th f yn o i t the drift-type microinstabilities, e.g., trapped particle or ion temperature gradient driven modes presence th n i , negativf eo e magnetic . Thishea7] primarils , si [6 r y because these instabilitie e usuallar s y regarde s beina d g responsibl e experimentallth r fo e y observed anomalous transport in tokamaks. Thus stabilisation of these instabilities by the shear reversal could produc effectivn ea e transport barrie plasme th n ri a interior.

In this work, we investigate the effect of shear reversal on the stability of the ion temperature gradient driven (7/i, where, T/J = 4' )r mode17 . This mode is widely regarded e primarth s a y candidat o accoune anomalout e th r fo t therman io s l transpore th n i t high temperature fusion moddevicesi r\ s e destabiliseei Th . d whe e parametenth r - rjiex ceed criticaa s l value T?;C, whosc T d , wheree T an valu ; ~ eT i e T lie r sfo betwee2 d an n1 electroe th respectivele d ar nan n temperaturesio e yth consideo T . e unstablth r e mode we thin i , s work, confine ourselve limie > r/iothen i I th c .rj to t sr interwordse ar e -w , ested in the discharges with flat density profiles, i.e., the VT,- mode. As also pointed . Tany [8]al b ,t limit e ge eve th ou zerf o tn n i o density gradient frequencw lo , y microinstabilities can persist because of the temperature gradient. Our full analytic stability

189

analysis shows thasheae th t r reversa weaa s kha l stabilising influenc unstable th n o e e mod ej?4 have "') r (r)iW > = 7? -e % ffocusse• ** l ,range al n f wavelengthsdo o s , short 1 fy,L= ? >b>l, ; T (4 -(VTi/T;)-e T= i , ^ wher T= e= /Ti,e r ,eb and ps

gyroradius)n io e th , intermediate (ehavd >&$>!an e) foun1 C < lond dg an )gtha (b t1 /2 i

this observation is independenT t of the wavelength regime considered. This is consistent with recent reversed shear experiments in TFTR [9] where no appreciable reduction in the ion and electron thermal diffusivities are observed by the reversal of shear (although the particle diffusivity reduces by ~ 50 times). So, the best way to stabilise these modes woul optimiso t e db profilee equilibriue th th f o s m quantitie describes sa referencn di ] [6 e so as to keep T?; < 7/f ** and also 5 < 0 in the plasma interior.

II. Stability Analysis

Using fluid descriptions a simple eigenvalue equation for the ITG mode can be derived in a straightforward way [10-13]. The leading order equation in the ballooning formalism [14] can be written as

2 2 ) (1 b Se+ (le -(cosd)+ SesinO)}^+ 0 = CIO

extendee th Here s \ i - 8 ,d ^ poloidau , ^ T ^ l . variable , sheare Vq'/q= = th t s 5 f ,i , CJBC^ I) UD = (Ti^Ti, B is the equilibrium magnetic field and R is the major radius. We will now solve this eigenvalue equation in various limits:

1 /2 strone , thith 1) s si g CAS> balloonin e b : (eE I > r't g limit thin I . s casmode eth s ei localise narroa o dt w rang , typicall0 f eo y nea outside rth toruf eo s wher curvature eth s ei unfavourable. treao T t this cas expane ew d equatio smalr fo ) ln (1 mod e eth widtd an h0 eigenmode equatio simpla s ni e Weber equation:

lowese Th t order eigenmod gives ei y nb

with,

190

r growinFo g mode , nondivergen(Imf0) > i t solution require choice lowee sth th f reo sign in equation (3) and the lowest order growth rate is given by (for fire^, 2 « 1)

(4)

The expression for the growth rate (4) contains the geodesic curvature contribution as exemplied by the (5 - 1/2) term. Let us now examine this for the two cases 5 < 1/2 and S > 1/2.

For 5 < 1/2,

evidens i t I t from equatio tha) sheae n(5 th t r reversa stabilisina s lha G IT ge effecth n o t mode. However typicar fo , l secone tokamath , d1) k> e parameterb . 1 > q s, 1 (e.g. ~ S , square terth mn i e bracket neven ca s r compete wit firse hth t term. Shear reversaa s ha l weak stabilising effect on the ITG mode and the mode is unlikely to be stabilised by the reversal of shear. The negative shear can give a substantial stabilising effect only when q howeves i t I . is 1) rsmal interestin~ q ( l noto gt optimisee th tha r fo t d inverted ^-profiles described in reference [6] and for the reverse shear experiment in TFTR [9] q < I is never observed ove entire th r e plasma cross section. This resul closn i s i te resemblance wite hth numerical work of Kim and Wakatani [7]. Our result therefore brings out a very important outcome: "the best way to stabilise the ITG modes would be to optimise the equilibrium profiles so as to keep rji < 7/f ** and also S < 0 in the plasma interior as envisioned by Rewold . [1993]"al t e t .

On the other hand, for 5 > 1/2, the growth rate is given by

t = Q d an r where2/ = P ,

Here also we find that the reversal of shear has a stabilising influence on the ITG modes (through the third term, as the second term just affects the real frequency). However, as before e shea effece th th f , ro t reversa mode th n eo l stabilit rathes yi r weak.

191 2 mode th thi, n ei 1 s limi> stil n g treate e stronb ca e ltb > th n CASdgi f 4 balloonin : EII g limit. However e smalth , l Larmour radius approximation assume derivinn di g equation (1) in the fluid approximation may not be valid. This case has been treated by Quo et al.-, unstable [15]Th . e root obtaine thin di writtee s b limi n [15]s ca nta :

= HereI0 (be)exp(-be).0 T , = /i(6g)exp(-&e I T e eigenfunctioth d an ) s givei n y b n where

(8)

Again choosin lowee wil) gth (8 lr q giv sige nondivergene en th i t solutio growine th r nfo g mode. Following the same procedure as in case I, it can be shown that for both 5 > 1/2 mode th weakls ei 2 1/ < y stabilise5 d sheae an th y r db reversal .

CASE III: be < 1, this is the weak ballooning limit. The mode structure is now extended alon magnetie gth c fiel dmultipla lind ean e scale analysis [16-18appliee b n orden ]ca di r solvo t e equation (1). Basicall alloe yw w variatio lengto tw hn ifn ni o r scales ~ 0(1 0 9 , ) and #1 ~ (bg1' ) ~ O(X~l) > 1, where A is a smallness parameter, to account Tor the rapid variation due to the periodicity of the equilibriu-a and to account for a lower secular sheao decat e ry du respectively expane derivative0 W . e dth s sa

JL A d + d8* 80~ 0

2 We solve equatio ) orden (1 orde y b r O(Ao t r averagd )an e ove fase t th rtge scalo t 0 e0

Note tha toroidae th t l effects ente eigenvalue th r e equatio Pfirsch-Schlutee th a nvi r factor lase th t n termi . Equatio standara s i ) n(9 d Weber equation equatios A .invarian s i ) n(9 t unde e changth r —5 sign , ei f > shea-5 no r reversa t expecteno s i l havo dinflut y an e - ence on the stability of the mode. However, to get insight into the nature of the roots

192 stabilised/destabilise e reversath y d b f shea o lwil e w rl proceed furthe fino e roott r th d s 2 / 1 actually. Witeigenvalue furthee th , hth 1 C < r gives ei choic i T y e n b f eo

2 3 2 1/2 4 n r w ±iqHS(l + 29 ) (10)

The eigenfunction is the Hermite function. The lowest eigenmode is given by

with

(11)

For growing mode (Imfi > 0), nondivergent solution requires the choice of the lower sign equationn i s (10 (11)d growine )an Th . g roo thes i t n givey nb

(12)

Now if 5-* -S,

It is easy to check that Re(p3) > 0 is also satisfied. So, the growing root gets stabilised by the shear reversal. However, thee root (given by the upper sign of equations (10) and (11), whic unphysicas hwa l becaus becoms Re(paf ha eo , 0 ereversae < ) unstablth o t f e o l edu shear unstable Th . e roo gives i t y nb

So, this unstable root is the same as given by equation (12). That Re(ps) > 0 is satisfied for this unstable root can be checked easily. So, the shear reversal simply interchanges the characters of the roots, the growing mode becomes damped while an unphysical root becomes growing. In other words, an unstable mode always persists irrespective of the sign of shear.

However, an interesting point is that in deriving equation (9) from equation (8) we have averaged over the fast scale 0 and as a result the term involving the geodesic curvature sinO)~ ( doe t appeasno leadino rt g order [equatio0 n (9)]. However earliee , Likth n ei r cases,

193 weaa k stabilisin gsheae effecth f ro t reversal exist thin si s case also seee . b Thi n ni n sca nexe th t order calculation eigenfunctioe Th . n correct upt nexe oth twrittee b orde n ca rn as

r.1/2

(13)

wherb = ea

It is clear that from eq (13) that by the shear reversal the mode will be more localised in 6>i and hence is more spreaded radially [20] and is more stable [12, 21-22]. So, to conclude, althoug leadinr hou g order calculation indicates that sheae thereffeco th n f rs eo i t reversa l on the stability of the ITG modes, a weak stabilising effect, like in the earlier cases, is found to exist in the higher order due to the geodesic curvature term.

III. Conclusion

n conclusionI have w , e show a simpl y nb e model equation thae sheath t r reversas ha l a weakly stabilising role on the ITG-like microinslabilities. We have carried out full analytic stability analysi regimel al r swavelengthfo f so ,intermediat, shor1) t> e (4/b /> e havd an e ) com1, thio t e < lond se an conclusiong(b ) 1 > e b . Substantia> (fTi 1 /2 l stabilisatio sheae modG th IT y r e b e reversa th f no possibls li . value1) q ~ ew onlq lo s( r yfo However interestins i t i , noto gt e optimisee th tha r fo t d inverted ^-profiles describen di reference [6] and for the reverse shear experiment in TFTR [9] q is always larger than 2 ove entire rth e plasma cross section. This resul closn i s i te resemblance wit numericae hth l d Wakatanan findingm Ki if o [7]s . Thi s alsi s o supporte e recenth y db t reverse shear experimen TFTn i t wher] R[9 appreciablo en electrod an e n reductionio thermae th n ni l diffusivity is observed by the shear reversal. Our result therefore clearly indicates that all efforts should be made to keep 5 < 0 and at the same time keeping rji < r/f ** in the same regioplasme th f no a interior [6]0 wher .< Since5 toroidae eth l microinstabilities temperaturn io drivee th y nb e gradient (VT{) remai nviabla e candidat accouno et r fo t anomaloue th s heat transpor interioe th n i t r part tokamaksf so r resulou , t shows thaa t scenario of creating a reduced transport level in the plasma core is realisable by inverting the ^-profile and at the same time by keeping 77; < r/f **.

References

[1. Kesse . C ]Phys al t e l . Rev. Lett 1212 7 . 2 (1994) . Hugo[2M ] t al.ne , Nuc . (19923 Fus3 2 3 .) . LazaruA . [3E ] t al.se , Phy. Fliud 3644 , sB 4 (1992) ConferenceS EP . Turner F h . 9t . Syke[4M e ProcA ]n d i ,th , f Oxfordsan o . , 1979 (Culham Lab., Abingdon, UK)p 1-161 [5] M. S. Chance et al., Nuc. Fus., 21, 453 (1981)

194 [6] G. Rewoldt et al., in Proc. of Local Transport Studies in Fusion Plasma, Varenna, Editrice Compositori, Bologna, 337 (1993) [7] J. Y. Kirn and M. Wakatani, Phys. Plasma, 2 1012 (1995) . TanM t . al.g[8e W ] , Phys. Fluids 3719 2 , 5 (1986) . LevintoM [9. . PrincetoF ]al t ne n Plasma Physics Report, PPPL 3117, 1995 (submitted to Phys. Rev. Lett ) [10] W. Horton et al., 24 1077 (1981) [11 . SenS ] , Plasma Phys. ControUe d(19955 Fus.9 7 3 ), [12] S. Sen, Phys. Plasma, 2 2701 (1995) . Romanelli[13F ] , Phys. Fluid , 1011 , s8B (1989) . ConnoW . [14 J t al.] e r , Proc . SocR . . London 365A , (1979,1 ) [15] S. C. Quo et al, Plasma Phys. ControUed Fus. 31 423 (1989) . StraussR . [16H ] , Phys. Fluids 2004 2 , 4 (1984) [17] T C. Hender et al., Phys. Fluids, 24 1439 (1984) [18] D. Lortz et al., Nuc Fus., 191207 (1979) . Kadomtse . B Pogutse P . [19B . 0 ] d ,v an Sov . Phys. JETP 1174 2 . 2 (1967) . . MeissHazeltinD D . . [20J R ,] d Physean . Report (19851 1 )12 . [21] L. D. Perlstein and H. L. Berk, Phy. Rev. Lett. 23 220 (1969) [22] M. Yagi, K. Itoh et al., National Institute for Fusion Science Report, NIFS 246 (1993)

195 PONDEROMOTIVE MODIFICATIO F NO XA9745756 DRIFT TEARING MODES

G. URQUIJO, R. SINGH, A. SEN Institute for Plasma Research, Bhat, Gandhinagar, India

Abstract

e lineaTh r characteristic f drifso t tearing mode e investigatear s e th n i d

presenc a significan f o e t backgroun f radio-frequenco d y (RF) e waveth n i s

cyclotron io n rang frequenciesf eo ponderomotive Th . e force, arising froe mth

radial gradientfielF R de energyth n si founs i , significantlo dt y modif innee yth r

layer solution drife th tf so tearin g modesn havca estabilizina t I . g influence,

even at moderate RF powers, provided the field energy has a decreasing radial

profilmode th t ea e rational surface.

Tearing modes play an important role in the general stability properties of a tokamak

discharge through their involvemen minoe th majod n i an tr r disruption phenomena [1-3]A . large number of studies have therefore been devoted to exploring ways of stabilizing these

modes by external means [4,5]. Numerical simulations as well as analytical studies have also investigated the effect of equilibrium flows on the stability of these modes [6]. Such flows can

aris byproduca s ea heatinf o t plasme gth a through neutral beam injection (NBI). Recently,

there is some experimental evidence that if such NBI heated plasmas are further heated by waveF R s ther significana s ei t increas core th e n eplasmi a temperatur accompanyinn a d ean g

improvemen plasme th n i t a confinement. Such shots, knowmodesH C s na , were observen do

the PBX-M tokamak during Ion-Bernstein Wave (IBW) heating experiments [7]. Motivated

by these results, we have examined the effect of a large background of RF fields on the linear characteristics of the drift tearing modes. In a simple model calculation we have investigated e influencth direce th f eto ponderomotive force, arising fro radiae mth l F gradientR e th n si

field energy, on the stability characteristics of the m = 1 drift tearing mode. The calculation

is restricted to the inner resistive layer, which is most sensitive to non-ideal dynamical effects,

197 d showan s thaponderomotive th t e forc significantln ca e y modif e stabilitth y y propertief so the mode. It can induce stabilization of the mode, for rather modest RF powers, provided mode d\VE\th t ea rationa0 2 /dr< l surface inducee , th where s i th velocit n E do io t eV e ydu external RF field. The basic equations for the drift-tearing mode in the inner resistive layer can be written

down quite simply fro mfluia d description s outlinea , examplr dfo [8]n ei .

O ~* -i p~ + P= -JxB- Vp (1)

(2) en = Q (3)

e oscillatinth whers i E fluin V egio d velocit F wavy R induce e th fiel y db (e.g n Iona . - Bernstein wave (IBW)) and the other symbols have their standard meaning. We adopt the

slab model unde assumptioe rth (r = withi, —n1 x layerr, e (sP1 r 3)/r n= / th « B sFo %. « the perturbed electri describee cb fielelectrie n th dca y db c parallepotentiae th d lan vectol0 r

potential A\\. We also assume that the ponderomotive force is radially symmetric with respect minoe th o t rmeae radiuth nd contributionan s momentun io e th n i ,m e equatioth o t e ndu

RF wave field, can be written as :

2 dr (4)

For u > k\\vthi and fcj_pt « 1, the ions can be treated as collisionless. Taking the parallel component of vorticity equation (i.e. Vx of Eq.(l)), and the parallel component of the Ohms Eq.(2)w La obtaie w , o couplentw d equation: \ A\ r d fo an

-x) (5)

- Slo (6)

l2 wher= kl/(2k\\ O Q e c^L ne magnitud) th —j£— s e i ponderomotivth f eo dre force= 0 , rs .22 (fcnc/o;)^ the normalized electrostatic potential, UG = — i(47r/77c )(u; — u;*e) the normalized conductivity, 77 is the Spitzer resistivity, x\ — u(u ~ u*i} I (k\\2 c\} •, w*e = ~ fl

0 T - the electron and ion drift frequencies respectively. k\\ — ky/ Ls, where ky is

poloidae th l wave vector sheae th s rL , scale length Cd B/-^/4npA an — , e Alfveth s i n velocity. For simplicity, we have neglected the electron and ion temperature perturbations.

198 solvw Wno ee Eqs.(5-6 drif1 ) = t analyticall tearinm e th g r modyfo usiny b ee th g

variational metho s outlineda [9]n di . Combining Eqs.(5-6 eliminatind an ) g <£, give fourtsa h

order differential equation in A\\. We further Fourier transform the resulting equation in x

in orde obtaio t r nsela f adjoint second order differential equatio whicq n ni the s hi n suitable for variational methods. The resulting equation is :

J' 2 1 S(q) ) (7 2 0 = U f ' A .

+ 00

' / -OO and A' is the driving MHD energy. The A' term is introduced in order to correctly take into

account the asymptotic matching condition with the outer solution.

derivee b Equatio n ca d ) fron(8 mvariationaa l principle fro functionale mth :

dq (8) A'

An exact analyti obtainechoosine cy b b for n 5 m r ca appropriatn dgfo a e trial functior nfo drif1 = t tearingm e th choose r w ,exp(—ag = Fo e J . J 2/2) witt ge h o t Re(a) , 0 >

e standarth wher s i Z ed plasma dispersion function e dispersioTh . e b n w relationo n nca JO foun eliminatin y closea b d r Fo dsimultaneously . b for0 g a m = — ; 0 y — solvin S r gfo da solution limie ' —>th (appropriato expanA te o n •I functioZ w , . e 1 e dth |/?r th r nfo « | efo

m = 1 mode) the dispersion relation we finally get is,

/ u ' v rjc2

standare I drifth n , —= L > tfieldF oo • absencd m e R d th e an e , resul th n 0 Dth I r Of = eo fo t tearing mod recovereds ei :

XA^O2 — ~ i31 o3r T = To- /" i i \ (11)

e n tearin^ s l g modR T = wher(~f~~ eR A — T growtT 0 a/c^ed 7 )— an A h T ratd ean 4^a2/(r;c2) are the Alfven and resistive time scales respectively. For the limit when u; ~

= u*i > 7 and Ti « Te (and OQ 0) the above dispersion relation also reduces to the known result:

199 (12) ,u,-eu;I

where w*, = —jf-u>+e. Solvin e dispersiogth n relation perturbativel r finityfo e obtaie f2ow ,e followin nth - ex g

pressio growte th r nfo hdrife ratth tf eo tearin gfield F presencmode R th e :n ei th f eo

We note from Eq.(13) that the RF field can have a stabilizing influence for the m — 1 mode

2 when fZ0 < 0. In physical terms this translates to d\V£\ /dr < 0 at the mode rational surface, fiele i.eth .d energy should hav edecreasina g radial profile. Thieasiln ca s y happee th f ni

energy schemF depositioR e th e primarily nb y occurcentrae th n i s l t somregionge eo T . idea of the RF power necessary to effect a significant stabilization we express the condition

c t ^ C I ^*- I I > I Q """•- GnJ. Lr ± 'u I » -T v VthJ P 1, \LS/ Ln

For typical tokamak plasma parameters we can choose Ln ~ a, Ls ~ qR, q ~ 3, a/R ~

.3, |f ~ 1, and the energy deposition scale length LE ~ -la- Then the value of parameter (jT^-J — 6.10~ 2 indicating that the RF power can be a small fraction of the thermal power

for this stabilization to be effective.

In conclusion, we have investigated the effect of the ponderomotive force on the linear characteristic e drifth tf o stearin tearin1 g mode — r calculation gm Ou mod. e th er fo s show that the presence of the ponderomotive force term leads to a significant modification in

the inner layer dynamics of the mode and could lead to stabilization provided the RF power

densitappropriate th s yha e radial gradient poweF R e r requirementTh . thir sfo s stabilization

to be effective are quite modest. In our model calculations we have ignored the effect of the externae terF th R mn o l solutions assuming that this modification wil smalle b l . Howeven ri

mora e realistic scenario powe,F wheR e rn th leve highs i l , this effec alsn significane ca otb t

should an incorporatee db calculatione th n di relevans i t I . mentioo t n here that such effects

have been investigated earliecontexe th n i rexternaf o t l kink modes [10,11 balloonind an ] g

instabilities [12] in tokamaks and in experimental suppression of the interchange instability

in mirror machines [13]. Another major simplifying assumptio presene th n ni t calculatios ni

200 e neglecth f sidebano t d coupling terms which generalizee aristh n ei d ponderomotive force

expression. These effect e presentlar s y under investigatio d wil reportee nan b l mora n di e

extended communication.

References

Bondeson. 1A . , R.D. Parker . SmeuldersHugo. P M , d nan , Nucl. Fusio (19911 n3 ) 1695.

2. 0. Kluber et a/., Nucl. Fusion 31 (1991) 907.

3. J. Howard and M. Persson, Nucl. Fusion 32 (1992) 361.

4. A.W. Morris et a/., Phys. Rev. Lett. 64 (1990) 1254.

5. A. Sen, A.K. Sen, P.K. Kaw and R. Singh in Procs. IAEA Technical Committee Meeting on Research Using Small Tokamaks (1991) Hefei, China, page 299.

6. R.L. Dewar and M. Persson, Phys. Fluids B 5 (1993) 4273.

, (IAEA-CN-60/A-3-I-7al t e o 7On . , Fifteenth International Conferenc n Plasmo e a Physi Controlled can d Nuclear Fusion Research, Seville Spain 1994)

8. J.F. Drak Y.Cd ean . Lee, Phys. Fluid (8)0 s2 , 1341 (1977)

9. R.D. Hazeltine and D.W. Ross, Phys. Fluids 21, 1140 (1978)

10. D.A. D'lppolito, Phys. Fluid (19881 (2)0 s3 34 , )

11. D.A. D'lppolito, J.R. Myra, S.C. Jardin, M.S. Chance,.and E.J. Valeo , Phys. of

Plasma , 342s2 9 (1995)

12. A. Sen, P.K. Kaw and A.K. Sundaram, (IAEA, Plasma Phys. and Controlled Nuclear Fusion) Vol. 2, 101 (1987)

13. J.R. Ferron . HershkowitzN , , R.A. Breun, S.N. Golovato . GouldingR d an , , Phys. Rev.

Lett. 51, 1955 (1983)

left BI.A" EQUILIBRIU FLUCTUATIOND MAN A N SI XA9745757 CURRENTLESS TOROIDAL PLASMA

R. SINGH, S. MAHAJAN, K. AVINASH Institute for Plasma Research, Bhat, Gandhinagar, India

Abstract

Equilibriu propertied man f fluctuationo s plasma n i s a confine toroidan i d l magnetic field are studied in the context of BETA device [Basic Experiment in Toroidal Assembly Institute th t a ]Plasmr eFo a Research [IPR] shows i t I . n that the fluctuation driven ponderomotive force and ion-neutral collisions can extend the free fall of the plasma due to effective gravity. Further, linear sta- fluctuatione bilitth f yo presence th n si self eo f consistently generated sheared poloidal rotation is studied. It is shown that the shear in the poloidal velocity

V'E has no effect on the instability of the fluctuations. On the other hand s relativeli f e effecVj th f o yt mor fluctuation0 e important< g V e r ar sFo . destabilize fluctuation0 > d g whilV r e stabilizedefo ar s .

It is known that a plasma cannot be confined in a toroidal magnetic field. The reason is that the bad curvature of toroidal field lines gives rise to an effective gravity which leads

to a vertical charge separation, and an outward ExxB drift of the plasma. This loss of equilibrium can be avoided by shorting the space charge field via a rotational transform. The most efficienintroducinf o y wa t rotationae gth l transfor toroidaa a vi ms i l current. However, ther othee ear r though somewhat less efficient way generatinf so g rotational transformr Fo . instance Yoshikawa1 has constructed toroidal equilibrium with finite pressure gradient at the plasma boundary. The resulting space charge field is shorted by a conducting limiter placed at the edge. Niewenhove2 has argued that even a fast poloidal rotation can effectively short circui space th t e charge fielprovidd dan e equilibrium othee th f r O .activ e method hithertn oi vogue the most notables ones are by using a vertical field or through error fields. Rotational transform can also be generated by passive methods e.g., using pondermotive force due to the fluctuations3 or by invoking ion neutral collisions. e reaso Th r studyinfo n e equilibriugth d confinemenman n suci t h currentless devices is two fold. Firstly, the study of interesting physics associated with the above-mentioned way generatinf so g equilibrium, confinemen transpord an t sucn i t h devices. Secondlys i s a , welknown, there are many instabilities in tokamaks where toroidal curvature of magnetic field lines play importann sa t rolepropertiee Th . sucf so h instabilitie resultind san g transporn ca t be better delineated in a simpler toroidal device which doesn't have the other complications of a tokamak. device th R eA IP tBETsuc e on h As i currentles s toroidal device parametere Th . thif so s devic cm.5 4 cm. 5 e,= 1 toroida are R ,= :a l fiel0.2-= , typicaB d 1kG l plasma densityN 10 3 ~ 5xl0 cm" and electron temperature Te ~(5-10) eV. The plasma is surrounded by a conducting limiter which supposedly provides the equilibrium. The typical characteristics fluctuatioe oth f n spectr Thre) (1 ae e observe ar cohoren , G 0 BETAn d20 i t = peak B 4 "t s7a 1.88xl0~ u t a 4,3.14xl0 6.28xl0d an 4 amplitud4e rad/sTh ) (2 . thesf eo e coherent peaks decreases with increasing magneti spectrue c Th field ) (3 .m becomes turbulen t highea t r field.

203 (4) The phase difference between density and potential fluctuations lie between ir/2 & TT. e basiOnth f abov o s e observations Mahajan et.al.8 have identified these fluctuationa s a s general clas flutf so e modes viz., Rayleigh-Taylor (R-T) modes, Modified Simon-Hoh (MSH) mode r Drifo s t typ f modeeo s etcr typicaFo . l BETA plasma parameters T modR- e th , dominates ove othee th r r instabilities. However these fluctuation alse ar so observee th n di good curvature regions. This has been attributed to either MSH modes, drift type of modes or poloidal convection of R-T fluctuations, due to the large poloidal flow, from the outboard to inboard of the device9. In this pape presene w r a self-consistent t stud f equilibriuyo d fluctuationman a n i s currentless toroidal device like BETAfirse th r stud t n ou I par.e analys f yw o t e forceth e balanc plasmf eo a along major radiu . FrosR m this analysi identife sw constrainte yth e th n so characteristics of fluctuations for them to impede the plasma fall and provide equilibrium. One of the important constraint is that if the ponderomotive force due to fluctuations is o oppost e fre e th plasmaee th falf o l , e fluctuationthe e leveth th nf o e outboarl th n o s d must monotonically increase wit e distancehth . Subsequentl stude yw e propertie th y f o s fluctuation detain i s shod an lw that indee fluctuatione dth s have appropriate characteristics to impede the free fall of the plasma. Further, we have shown that Vg introduced a radial asymmetry in the mode structure (or radially propagating mode) which generates poloidal flows through the Reynolds stress. It is found that for the typical BETA parameters, this flow is of the order of ion acoustic speed. This paper is organized as follows. Firstly, we discuss the equilibrium in the presence if ponderomotive force due to fluctuations and ion-neutral collisions. Then, the properties of flucuatione th analysee ar s detailn di . Finally summerise w , discusd ean resulte sth s We begin by analysing the equilibrium of a plasma in a toroidal field in the presence of fluctuation ion-neutrad an s l collisions. Here establise w , conditione hth propertiee th n o s s of fluctuation d ion-neutraan s l collision e plasmth r fo sa equilibriu e majoth mn i r radial direction. Withi e singl equatione th th e D flui motiof no MH d gives ni y nb

mtN~ = ~ - VP - mtNVmV (1) e plasmar J ad numbean P , wherrV" density, N e , velocity, pressur currend an e t density

respectively, i/m represents ion-neutral collision frequency and dV /dt = dV /dt + (V -V)V. From Eq.(l) and the charge continuity equation , we get

c da_ - - c dV - c • V - = L J • V - = — m.NBx— jjBxVP+ —m+ Nv BxV dt t m (2)

We use cylindrical coordinates (R, <^, Z) where R is along the major radius, 4> is the toroidal angle and Z is along the vertical direction. Typically in toroidal devices of this type the

electric field is mainly along Z in which case from the ideal ohm's law VR = —cEz/B while

. The0 ~ Vn using idea Poisson'ld ohm'an w sla s equatio Eq.(2)t n ni ge e w , z

d P d m c lNcdV2 R dV R d B dt 4Trcdt\dZ)B dZ\

s beewherha n V e radiaapproximateeth V) • l V par( f o t s (1/2da ) dv^/dR s ther(a s ei

perturbee Vd th Rs i an ) 0 dtoroida~ o n velocitz V d lradian an yi flo 0 w ~ l i.e. directionv V , .

204 In Eq.(3 have w ) e explicitly fluctuationso separate t ter C e mD. du e dth . Integerating

have Eq.(3w e, 2 Z ) o frot i mZ 1 z 4 - - -dvK dt r i?w < >

wher acousti n = wTio s e c e /mth c s i speed, , C B*/4TrNmA— e Alfveth s i nt speed dan (1/2)(d \ VR I fdR) is the radial ponderomotive preesure due to the background fluctuations. In Eq.(4) it should be noted that VR ~ (c/B)(d/dZ}. Since equilibrium is expected to be symmetric about Z = 0, would be an even function of Z. In this case EZ = 0 at Z = d increase0an s wit . Thu hZ e las th d sigst f dV^/dZan terno 0 ms negativ i > e thereby implying that ion-neutral collisions imped radiae eth plasmae lth falf o l . Thi physicalls si y 2 2 reasonable as ion-neutral collision dissipate plasma momentum. Since c /c 1 >> 1, Eq.(4) can be simplified as

l f i 2 ,, ^T/_ ) d(5 t R 2dR> K-".-VI i l *

wher lase th et ter s beemha n approximate « v inr steads Vda Fo R. y faln li d/dt0 ~ which case VR is given by * 2 ~ ^ r /^~* i 2 (6)

Clearly if dvj^/dR > 0 then the fluctuations driven ponderomotive force opposes the free fall. For equilibrium, the required fluctuation level is

(7)

interestins i t I noto gt e that plasma fall tim eextendee a/V/j(Tb w excitin) n (1 ca ) y db g

e shorteth r wavelength fluctuation goiny b ) higheo gt (2 s r atomi increasinC y b c r maso ) s(3 g neutral pressure. 1 For typical BETA plasma parameters B = 500 G, ky = 0.3 cm' , Te = 5 eV, R = 45 cm., a = 15 cm., we get |£ « 33% •*• e As stated earlie t highea r r magnetic field, spectrum tendturbulene b o st t with significant powe largn i r e k's. This will further reduc e criticaeth l fluctuation level requirer dfo equilibrium at higher field. Now, we study that the properties of fluctuations in detail to show that the magnitude and direction of fluctuations driven ponderomotive force is indeed appropriate for the plasma equilibrium e basith f experimentalln o s O . y observed feature f fluctuationo s BETAn i s 6'7 theoreticad an l investigation identife w , dominane th y t backgroune modth n ei d turbulence to be due to Rayleigh-Taylor instability. In particular, we study the magnetic curvature driven Rayleigh-Taylor (R-T) instabilit presence th n yi experimentallf eo y observed electric field curvature. We start with Braginskii10 two fluid equations for studying the magnetic

curvature driven low frequency (u < flt) R-T modes in a current-less toroidal device. We a sla e b us approximation wit d e R an —Th , »>b y z * . z approximatio s valini d

205 if the wavelength of the mode is smaller than R (the major radius of the torus) and a (the minor radius) plasme Th . places ai vacuun di m curved magnetic field which decrease 1/Rs sa . 2 The terms like V • (BxVfi/B ) are transformed as w (2iky/BR)fi where ky = m/r is the poloidal wave number alse W o. assume that eigen-functio slowls ni y varyin (poloidal6 n gi ) variatio B d thu e an th s f modno e amplitud ignores e i e analysisth n di e anticipat W . e eth

flute eigen-functio) 0 — z (k n

(p exp [—i (u>t ~ kyy)] (8)

We now take the curl of the ion and electron momentum equations and the results doted with B. The sum of the projection of the curl of electron and ion momentum equation along B and the continuity equations give two coupled equations for n and (p. The electron continuity equation,

r\ + VE(X)}n + — -^ = 0 \ot dy dy + e-adyn ( - (9)

and the quasineutrality condition (i.e., V • j = 0) (1 dy -£ °>

where the dependent variables are normalized as n = n/N, (p — e/Te, moreover, N = e equilibriuth d an ar : N <]>=((>n m , N+ ¥ densit+ d potentiayan l functionsd an n , e perturbeth e ar (f d densit d potentiayan l fluctuations = a e e diamagnetiscth sV» /L ,s i n c l drift velocity n = cio y/Ts= /fl-dlnN/dx,, = e a s , ezc th /rm,, s L~ i = T./T^- t K Q

cyclotron frequency, en = 2Ln/J?, V'E and Vg are the first and second derivative of ExB in directionx e —th — cE E xV fB., Her have ew e ignore e electrondth d ionan s s temperature perturbations. 2 We choose Ex = Ex(0) + E'x(Q)x + E'^(0)x' /2 to separate the shear (E'x) and curvature (E") effects aroun eliminatinn locae O dth . l 0 poin = gfrox h t m Eqs.(9 d (10an )) using followine th Eq.(8)t ge e g w ,eigenvalu e equatio weae th kn ni shear limit.

,22 £» (1 - en) K (H)

wher x/a— a/2/3er s+ ,

2a V ' s E ft ' n.v// Cx. *~~ P 5 E VE

a) = — fcV£;(0))/u)»e is the normalized Doppler shifted eigen value , VE = LnVg/cs, VE = (asLn/2)(VE/cs) are the normalized electric-field shear and curvature paramters. Thi standara s si d Hermitradiae th s i l1 ( mode equation0 e= 1 number r Fo . ) casee th , eigen function of Eq.(ll)

exp (12)

206 while the linear dispersion relation is expressed as

2 a /2 s _ = _ _

For Ti/Te < 1 and in the absence of V& and V£, the Eq.(13) gives

l/2/i \l/2 r ,22 • n - n 6 *~ • Kaen 5 7o = n en(l-en) Thistandara s si d dispersion relatioT modeR- f no . This shows thamode th t unstabls ei e for poloidal mode number (m)f i ,

x 1 < n e d } f^—-^.}an (— 2 < m (15) ' j n t \a\ sj tak w effece eno th electrif e o t W c field shea curvaturd an r e perturbativel neglectind yan g the Kelvin-Helmholtz term in Eq.(13). For 1 = 0, the result is

(16)

with the constraint Re(/31/2) > 0 for the growing mode which states that the eigen- functio s spatialli n y bounded e e Eq.(16 e th r.h.th termo e firs th f e tw o Th sn t ar .o )s standar thire dTh reae drespons. 0 th ter T lmodee d — par th mR- g an f f eo V to s wit— g hV last term on the r.h.s contribute to the doppler shifted real frequency while the imaginary part of the last term indicate the destabilization of the modes due to electric-field curvature e electric-fielth , 0 > g dV r curvaturFo . 0 < e ' stabilizefoVg r e mod th sd als ean o affects the real frequency. Here, it is interesting to note that the growth rate is independent of electric-field t affecti shea d an rs onl reae yth l frequency par f u>o t nexe W t examin e behavioueth eigen-functioe th f o r n e effecEq.(12Th f o x —t. >t oo a ) electric-field curvatur takes ei n perturbativel mode th n ey o structure resulte : Th .e sar (1) For Vg' < 0, the mode is unstable, corresponds to a bounded solution i.e., Re/31/2 > 0. 1//2 1/ 4 The typical mode width of the A-s^M^eigen-function envelopre A = (as/(2 /3 ' )) is (n)

Ky 870 Vg J

mode th , stabls e0 i correspondind > ' ean Vg r (2Fo ) g eigen-functio boundeds ni . The foregoing analysis shows that if Vg < 0, then the velocity curvature drive is destabi- lizing while for Vg > 0, it has a stabilizing effect. The observed potential profile in BETA5'6 shows that E" > 0 at an off axis location (6-7 cms.) away from the minor axis and E" < 0 near it. On the basis of the present analysis one would therefore expect the fluctuations which increases fro minoe mth r axis Fig.(l)n I . giv e experimentalle w , eth y observed fluctuation leve BETn i l shoAo t w tha toutboard e thith sn i indee casee w fluctuatioe th th ,No s d.i n level increases wit i.e.hR , requires dvj^/dR a equilibrium e 0 th r > dfo . e samInth e figur a significan alse e w eose t leve f fluctuationo l e inboarth n o sd where the magnetic curvature drive is absent. This can be due to number of reasons. First the convection of Rayleigh-Taylor fluctuations due to poloidal flows from the outboard to

207 60

45 /\ K-" 30

0 -10 -6 -2 10

MINOR RADIUS (cm)

Figur : Radiae1 l distributio densitf no potentiad yan l fluctuation amplitude . G 0 80 ) (b d an G 0 20 = B ) fo(a r

208 inboard9 occurs on a time scale ~ a/V^, while the mode linearly grows on a time scale 1 1 7" ~ (cs/^/RLn}~ . Typically (cry/Ve) ~ a/^/RL^ < 1, hence the fluctuations can be convected from outboar inboaro dt d withi growtna h time othee Th . r reaso fluctuationf no s inboare onth modifieo t de couldu de d b Simon-Ho h instability driven fluctuation simplr so y due to drift modes etc.7 Nevertheless, there is significant fluctuation levels on the inboard with a decreasing profile with R (the explanation of which may be given in a future work). This will giv en outwar a ris o t e d ponderomotiv e inwareth forc n d(o e locationsn a d an ) asymmetr densite th n yi y profile. Thi agais si n consistent wit experimentae hth l observation in BETA6'7. To summarize thin ,i s pape have rw e carrie sela t f dconsistenou t stud equilibriuf yo d man fluctuation currentlesa n i s s toroidal device. Firstly have w , e studie plasme dth a equilibrium e presencith n f fluctuationeo neutran io d lan s collisions r analysiOu . s showe s th tha ) (a t

free fall velocity decreases with increasing z/tn (b) If fluctuation level increases with R, the the ponderomotive force due to the fluctuations can effectively break the fall of the plasma.

The threshold for VR = 0 is e(p/Te « ^2a/R(l/kyas}. Then argue w , e thabackgroune th t d turbulenc currentlesa n ei s toroidal devic pres ei - dominantl o Rayleigh-Taylot e ydu r instability have W e. studie e lineadth r propertief o s this instabilit e presencth equilibriun n yi a f eo m electric field whic s generallhha y bee- nob served in such experiments. Our analysis shows that if E" > 0, then the velocity curvature destabilises the R-T mode while if E" < 0, then the curvature stabilizes the R-T mode. Thus, if electric field profiles is such that E" > 0 at an off axis location in the outboard and E" < 0 near the minor axis then the fluctuation level will increase away from the minor axis as required by the radial equilibrium. We have quoted results from BETA (Fig.(l)) to show that typically such currentless toroidal machines do have an equilibrium electric field with E" > 0 on the off axis location and E" < 0 near the minor axis which give rise to an increasing leve fluctuationf o l s away fro minoe mth r axis. e see b e crucia nth n Aca s l elemen r modee rol ou f electrith e o n s i i tl c fiel r poloidao d l flows. The observed density gradient is steep near the minor axis and become broader as we go away from it. Since Rayleigh-Taylor growth 7 a L~1/2, one would therefore expect fluctuatioe th n level decreasin outboarde gth witn i hthiR n I . s case equilibriun a , m with ponderomotive force due to fluctuations will not be possible. As shown earlier because of nea 0 mino e < th r " rE axisaxif d e radiaof san th , e th l t gradienconditioe a th 0 > f " o t nE the ponderomotive pressure has the right direction to oppose the fall. Apart from this, the poloidal flows play another very interesting role. It has long been recognize contexe th n d i astrophysica f to l sceneri recentld oan Diamony yb co-workersd dan 11 in the context of L-H transition in tokamaks, that if the modes are radially propagating, then the quasilinear Reynold stress (v- V)u in the poloidal direction is non-zero. This torque increases the poloidal rotation. If the radial asymmetry is due to poloidal valocity shear or curvature itself, then ther a spontaneou s ei e plasmth sf o spi a p whernu e fluctuations flowd e sustainean ar s d self consistently. From presene Eq.(12) thae th n se i t e t w ,case e th , mode e radiallar s y asymmetric (a/2/9 term asymmetrd an ) s premarilyi velocito t e ydu y shear Reynole Th . d theso strest e es du modifie modeT dR- s wil selly thef ma finite nb d ean consistently sustai e observenth d poloidal flows orden I . estimato t r e this effect write w , e ^-componene th meae th f no t momentum equation

O e v ) V • r (v = > e V —< (18)

209 where flow is mainly damped due to ion-neutral collisions and generated due to R-T modes through Reynolds stress. The mean poloidal flow in steady state is given by

:<]*>« We.. I ^ 12 (19) 2 A / 2i c

2 1 where we have used v = (c/B )BxVif>, kx ~ A" etc. For typical BETA parameters 3 1 1 . s Thiagreemenn c i s V< si ~ g > t cm"~ 33%(3 p ge 0. e s"t, ~ w ,i/,wit 10 y k ,nh ~ experimental observation sucn i s h devices wher esounn ordee flowio th f f do ro s speed have been measured. This raise interestinn a s g possibilit sucn yi h currentless toroidal devices where equilibrium, fluctuations and flows sustain each other self-consistently.

REFERENCES

[1] S.Yoshikawa, W.L.Harries and R.M.Sinclair, Phys. Fluids 6, 1506 (1963). [2] R.Van Nieuwenhove, Plasma Phys Controlled an . d(1992)3 Fusio87 , .n34 [3] K.Rypdal, E.Gronoll, F.Oynes, A.fredriksen, R.J.Armstrong, J.Trulsen and H.L.Pecseli, Plasma Phys. and Controlled Fusion 36, 1099 (1994). [4] D.Bora , Phys.lett. 139, 308 (1989).

[5] G.Prasad, D.Bor d Y.C.Saxenaan a , Geophys.Res.lett 1 (1992)24 , ; 19 , ibid Geo- phys.Res.lett, 19, 245 (1992) [6] K.K.Jain, Phys.Rev.Lett. 70, 806 (1993). [7] G.Prasad, D.Bora, Y.C.Saxen S.D.Vermad aan , Phys.Plasmas , 1831 , 2 (1994). [8] Sangeeta Mahajan, R.Singh and P.K.Kaw, Part-II of this paper (Communicated 1995). [9] P.K.Kaw (Private Communication)

[10] S.I.Braginskii, in Reviews of Plasma Physics, edited by M.A.Leontiwich (Consultant Bureau, New York, 1965), Vol.1, p.205. [11] P.H.Diamond, Y.-M Liang, B.A.Carreras and P.W.Terry, Phys.Rev.Lett., 72, 2565 (1994).

210 XA9745758 A TRANSPORT BIFURCATION MODE ENHANCER LFO D CONFINEMENT WITH NEGATIVE SHEAR

K. AVINASH, P.K. KAW, R. SINGH Institut r Plasmefo a Research, Bhat, Gandhinagar, India

Abstract

A magnetic shear -Invert trar.5CK>rt bifurcation r xodelfo

transition to enhanced confinement ^egime Kith r-eaatii'e shear and

neutral beams is proovsed. Strong fueling by high potter beams

leads o t sealing f o oressure oronle arid generation f t/ large

bootstrap current. The resulting negative shear reduces the groNth f o fluctuations. e transitionTh - tc enhanced confinement regime

occurs when fluctuations are completely quenched. Relevance of

this recentto results from TFTR brieflyis discussed.

It has long been recognised that the viability cf Tc'r amat as a

fusion reactor depends not only on a good thermal insulation of the

hot plasma but also on an efficient scheme of steady state current

drive. Recent theoretica investigation• l s Q,23 have shown that

these objectives can be simultaneously achieved in tcrt'amaks with

reverse magnetic shear S(r). The shear 5(r) is defined as S = — g-^

where q is the safety factor and r is the radial coordinate. These

tokamak equilibri e characterisear a y b highld y pea d £e pressur e

profiles in the core where a variety of modes believed to be

responsibl r particlfo e d energan e y losses e stabilisear e , th y b d

local negative shear 3 ' O L13. The essential point is that highly

"=?l sd cress'j^s crc^iies generate 3 strong off 5"i= bootstrap

zjr^ent, wnicn selfconsistently sustains the negative shear in the

ZDre and also provides a stsaav state current d»-ive. Experiments

D^oduring such configurations transiently have recently been

carried out on Doublet III-D ar.cJ "^TR L3,4D. In Doublet III-D a

211 strong e pressurpeak-inth f o g e profile within e negativth , e shear

^egion suggest e formatioth sa transpor f o n t barrier. Howeven i r

e experimentth , strong neutral beam induced toroidal flows were

also observed e resultTh . f o experiments n TFTo s R CTotamak Fusion

Research Reactor]e otheth n ro , hande ar mor, e dramatic £43A .

forty to fifty times reduction in particle and ion thermal

diffusivity was observed. The bootstrap current constituted nearly e totath f lo plasm 4 3/ a current e neutraTh . l beam injection (NBIJ

was nearly balanced o s thae odserve, th t de corth flow en i swer e

weat:. It thus appears that, contrary to the situation in D III-D,

in TFTR tna role OT .-nagnetic shear is more predominant.

A begining toward a stheoretica l understandin f o thesg e

orocesses has 35=n made by Diamond ana zo-wori-2^5 E53. It is

suggested e thatransitio th e tenhance th o t n d confinement regime

in these experiments e belonggenerath o t sl clas f o transpors t

bifurcation processes. The main results of their model is that the

bifurcation in the transport is essentially driven by the sheared

radial electric field Er produced by toroidal flows. Reduced

transpor o t negativ e du t ee cor th s sheai assume en i r o t plad y

additional subsidiary role. The relevant 1-d transport equations

are solved and the spatial location of the barrier and threshold

etce calculatedar . .

As stated earlier, in TFTR the role of magnetic shear seems to

be more important. In the present paper therefore, we explore a

moael in which a negative magnetic shear driven transport

bifurcaion dominate e maith e nf ohysicsth o sresult e On f o .thi s s

model is that it brings out the conditions of bifurcation to EF:S

'-egiine very clearly. It is shown that the reversal of shear is only

a necessary condition. In addition, the NB power must exceed a

certain threshold for the transition to ERS regime. These results

are consistent with the observations on TFTR £43. As a

212 simplification n i thi, s pape e considew r a minimar l O— d model where

parameter e spatiallar s ye functionar average e card th an f eo n i sd

time only.

We begi y derivinb n n e a equatiogtemporath r fo n l evolution

of core averaged shear S(r) in the presence of ohmic and Bootstrap f~* t i currents e BootstraTh . p curren t = —=densit • — s , i J —giveV , y PJ b yn b b B@ e mino th d majo an e e r plasmar th r R radiis i L<= aa/Rd = p an , ,a

e th poioida s i l pressurB magneti d an e ce th higTieidD hn I .

Bootstrap current regime, 'the peaking of pressure profile is

f o densite pea th g in t o t ymainl e profil n i whicdu y n eh ^ cas . J e 1/2 —— —— 5 — . Furthe e considew r a cylindricar l geometry (r,€»,z) where Be

': r and z are the radial and toroidal directions ^espectivel>. Using

E_^ = r, U*.^ ~ ^h-.^ ^ ls ^ne plasma resistivity and J, is the •* ^g total plasm = —ae obtai =rw E -current 7x n d followinan ] g equation #t -'at tne evolution of S.

l " • - - • - - ds 1 •~

e cartn wheres i average3 e ^ q- d— shear= c r?/4rra^N = _r _ ,d an ,

is the care averaged dimensianless density gradient and r_ is the

^adius of the region with negative shear. This coupling of pressure

gradiene magnetith o t t c shear throug e excitatioth h f o bootstran p

e currentessentiaon s i s l r elemensheaou n ri t driven bifurcation

Ticdel e h n=-;T . o t modete ste lineath s i e lp th r f growtc ) h

fluctuations responsible for particle loss as a function of shear

. 3 Accordin e presenth o t g t wisdom variet f o moaey s e.g. trapped

electron mode, neoclassical tearing modes, ballooning modes Cl,t3

etc. are held responsible for particle and- or =ne-- i .zs= -- on ^e

rz'-'e. A ie~s =. ~cs"~ ~~ =C~E r ~~ = = e TC2== =•= ~~~~~ '""s, a^e T = =-Di_-gz - .r h; negsn. , i . =^- z 3 == e sr.ear A generi. c, , •'ar f o m

wnich models this dependence on S is

, = ^=- Cl + ta-r, 5'5 D 2X

213 where y is the full growth rate for conventional shear case. The

numerical calculations of growth rate displayed by \ essel et.

al.ClH very dramatically exhibit this behaviour. Typically, in the

conventional e sheath n ri case d corth an e e 5 0. average ~ S d

negative shear case S ^ -0.1 L13. With this in view we choose S ^c

U.u5 o thas r ,conventiona fo tr negativfo d an ey l- sheashea= y rr e temporath r Fo l. O ~ evolutio-0. e ~ 3 a-fluctuatioth 1* v f o n n

level E = T n~Vn~" (n is the density fluctuation level) and N(t) we

the equation of Diamond et al. E53 which are

D N B EN NL A 2 2 n t d r r S 5

where Q is the particle source due NB1 fueling and can be directly

e poweth d an e r ar = rr^ W relate—Q poweB N d s wher a an o r t d P e

energy of the beam, V is the plasma volume and _«,, depends on the

fueling efficiency which usually ranges between '.'.5 to O.B. The

e poloiaath s i C_/p paramete l K ^ <.W 3 s , i

radia e linstability th widt f o h . w frequencTypicalllo D r fo MH yy e neoclassicath e ar E D l d . Furthean c se D turbulencr 0 1 * r , a e 1 NC A and anomolous particle diffusion coefficients) . (4 Equation o t ) (1 s

constitute a coupled set of non-linear differential equations for

e th time evolutio f o S(t>n , o Man ) e .Th y

states admitted by these equations are

State I:

= 1 0 N ' ° € " - 1 <* S2 0 & 1N =

214 State II:

r P ^ s > (6 E = = N S = N ° * 2 0 2 '0 ' * » - ^ 2 0 ' ° * : - 0 2 2<*0 & T NC

e totath 5 l1 wher numbeN e* particleso r e -firsTh . t stats i e characterise a positiv y b d e shear, finite fluctuation a level d an ,

small density gradient N. This can be identified with the L-mode.

The second state with zero fluctuation level ,d negativlargan N e e

S p=an be identified with ERS regime. A simple linearisation around

these stationary point gives

.1/2 — 11 E D N (7) o A o

where d/'dt = ^ To e"a

e notw II e thae righth t t hand sid f Eq.(7o e s i smal )r botfo l h

of them i.e. N ~ 0, E . _, =* U. Hence the stability is decided by t s " y d an _ beaw lo —y m t = A - powe 2j». 1 Ey x | r e = rooth 1 y t o s tnaO L madt e root e stablebeath ms A .powe s i increasedr e th ,

L-mode root merges with ERS root so that at high beam power only

one root characterise y statb dI I existse . This bifurcatios i n

called the inverse pitchfort bifurcation. One can clearly see that

the transition to ERS regime occurs only if the beam power is large

enough to ensure ?• = «\ In case if P is not es large, then there is

a continuum of states between I and II characterised by y > O,

positive or negati.-e shear and large N. This regime can be

identified with the supershot regime frequently seen in TFTR.

Further the threshold for shear reversal is well defined and is

QI v en 3_, P = 2 u E vJN ' s "", '-^ - r *"- ^or TFTR parameters with fj ^ A T y 2 S & •-» *n 3-4, D^E ~ 0.25 m^/sec, W ~ 120 KeV, N = (fta^) 2.n R n, R = 2.6 m,

a = 0.94 m, r_/a ~ 0.3, cs_ = 0.5, n=102°m~3, we obtain P ~ 1 MW

which roughly agrees with the observed value of 5 MW. The threshold •^o'- t^ansirion to ERS, i.e., v = o zn T:he -ther hand, is weldefined

215 only in the limit of small Sc. The threshold for EPS is roughly

given by P (2.O5WN D )/(.-r_). Par T^TR parameters given above

and 3 2: '_>.C5 F ^ 1_ :i z^i-i/ 5--~ = £ = ^..f fs

1 = 2= --5-- = ~l7 rt = =" =- . ai'je cf F ^ 1--I. Mw. The value of 5 is c e a detailefixe-cb oy b d d numerical work modelling shear dependence

f lineao r growt laterp u } h whic n e . r hta wile b l

To summarise we has'e considered a magnetic shear driven

transport bifurcation mode r transitiofo regimS l EP e o t nobserve d

recentl n TFTRo yn I thi. s mode a stronl g fuelin o t higg h poweB N r f o pressura pealead g o in lt s e profilee caree resultinth Th .n i s g

bootstrap current generate a negativs e shear which reducee th s

linear grawtn of a nsjmaer G^ modes ''esponsible Tor particle and

energy losse d improvean s e corth s e confinement. This a lead o t s

furthe e pressurth rf o pea eg in l profile. TransitioregimS ER e o t n

occurs when fluctuation e completelar s y stabilisee Tn . dO = i.e j .

peaking of density profiles stabilises the ITG mode leading to a

reduction in the ion thermal diffusivity as observed in TFTF:C4D.

The model shows that reversa f o sheal s i onl ra necessary y

condition n B poweI additioN . re th nmus t excee a threshold r fo d

transition to ERS regime. These results are consistent with the

observation n TFTRe o presens Th . t mode s i differenl t from thaf o t

Diamond et al. C5] who consider a E-field driven bifurcation. Since

NBn i TFTI s i nearlR y balance e induceth d de th flow e d weaar san k

effec f magnetio t c shea s i morr e dominant.

REFERENCES

1. C.kessel, J.Mamckam, B.Rewolt and W.M.Tang, Phys. Rev. Lett. 72, 1212 (1994).

2. A.D.Turnbull, T.S.Taylor, v.R.Lin-Lin and H.S.John, Phys. Rev. Lett. 74, 718 (1995).

3. E.J.Strait et al. Fhys. Rev. Lett. 75, 4421 (1995).

4. F.M.Levington et al. Phys. Rev. Lett. 75, 4417 (1995).

216 . 5 P.H.Diamon . al Bull t .e d Am.Phys. Sac , (19954O . ) pape. no r 7Q21.

6. Z.Chang et al. Phys. Rev. Lett. 74, 4643 (1995).

. 7 B.A.Carreras, D.Newman, P.H.Diaman d J.M.Liangan d . F'hysf o . Plasmas K12J, 4014 (1994). TOMOGRAPHY USING NEURAL NETWORKS XA9745759

G. DEMETER . ZOLETNIS , K Departmen Plasmf o t a Physics, KFKI, Budapest, Hungary

Abstract

Impurity injection using laser accelarated pellets and the study of the transport of these injected impuritie bees majoe sha nth r fiel f investigationo d recenn si t yearn so the MT-1M tokamak. In some experiments, two 16-channel MCP cameras were placed at various cross sections of the torus, one horizontally, the other vertically. With various filters added, these cameras provided information on the distribution of the injected impurity ions in a cross section of the torus. More precisely, each channel f theso e cameras measured some sor f integrao e radiatiot th e injecte f th o lf o n d impurity ions alon glineaa r domai crose th f sno section. These signals were digitalized d thu an a tims s everu e0 1 yevolutio e clouth f impurito f do n y ions coule b d investigated. The problem is to restore the original distribution of impurity ions in the cross sectio torue th f sno fro m measurement wels s a possible s la . Obviously f informatioo t lo a , s losi n t because each channel integratee th s radiation along a linear domain, and it is impossible to restore the exact distribution. o proceet Ony o makt wa es i d e additional assumptions abou e distributionth t n I . practice, this mean tryine s ar tha fine o gt tw distributioda fore givea th f mo n nn i class of functions. For example, a linear combination of some functions may be used, and the coefficients that minimise some sort of error function may be selected. This is fairly simple and computationally efficient, however it requires a good choice of base functions that resembl e symmetrieth e e distributioth f o s n involve e trulb yo t d meaningful r case assumptioe ou th ,n I . localisea s ni d distribution moving wite hth ablating pellet, and therefore such a linear fit is not possible. A nonlinear fit of a localised distribution with parameter horizontae th r sfo verticad an l l positioe th f no center, the amplitude and the width in both directions would be a good candidate, it is, however computationall e evaluatiotediouo th useto e r b r dyfo fa o s t f masseno f o s experimental data. Neural networks have been use r fasfo dt measurement evaluatio plasmn i n a physics previously, including nonlinear curve fitting to experimental data. Such an approach for the fast evaluation of tomographic measurements was also utilised on the MT-1M tokamak. Since with a given distribution it is straightforward to calculate the values thavarioue th t s channels would measure simpls i t i , builo et gooda d database on whic e neurahth l network e trainedb e networkn e theca sTh b . nn useca s o t d estimate the parameters that a nonlinear curve fitting would produce. This method has the advantage of performing tedious, time consuming calculations only once (during e traininth g phase), after whic e traineth h d networks process data quickld an y efficiently. In our work, we have explored the applicability of neural networks to such nonlinear tomographic problems, and found them a useful tool for the fast processing of large amounts of data. Several test were performed on the reliability of neural network estimates, and on the amount of time needed for training the networks compared to the time needed for conventional nonlinear fittings. The network estimates have proven to be accurate enough to be used without further processing. Alternatively, the estimates provided by the networks can be regarded as preprocessing

219 f experimentao l data used startingpoina s an ,da regular fo t r nonlinear curve fittino gt lessen computational time on data of interest Neural networks have the advantage of being insensitiv noiseo et thid san , robustness makes them tas e ideaf pickinth k o r fo l g out automaticall importane yth t feature f noiso s y experimental trainine datTh a g time require e sam th s requires provea ee ha b d r threo arouny no t b d o e dtw hundre d conventional nonlinear fits, and therefore well worth investing into Even though new networks e trainehavb r o everet fo d y experimental setu o accommodatt p e th o t e geometry of the detector, they have proven to be a useful aid in transport studies

Impurity injection using laser accelerate d e transpor e studpelletth th f yd o f san o t these injected impuritie bees majoe ha s nth r fiel f investigationo d recenn si te yearth n o s MT-1M tokama , 6-10k[1 somn I ] e experiments 16-channeo tw , cameraP MC l s [11] were place t variouda s cross section toruse th horizontallye f on so , othee th , r vertically (Fig1 show e experimentath s l setup) With various filters added, these cameras provided information on the distribution of the injected impurity ions in a cross section of the torus More precisely, each channel of these cameras measured some sort of integral of the radiation of the injected impurity ions along a linear domain of the cross section These signals were digitised every 10|as and thus a time evolution of the cloud of impurity ions coul investigatee db changine camerao channele signaTh tw e d th a th e n f th n sgi o o lf so

Soft^C-ray detector Vertical camera

Soft X-ray detector

Horizontal camera

A Pellet injection Fig. 1. Experimental setup.

signalchannelsthe The Fig.on 2. cameras ofthe typicala m experiment, horizontal camera on the left, vertical camera on the right.

220 typical experimene see b o correspone seent n b Fig.2n nca o n t ca a localiseI t. o t d d distribution around the center of the tokamak cross section in the horizontal direction, and moving e froplasm e edgth m th f o ea toward e e verticacenteth th s n i r l directione Th . proble restoro t ms i originae eth l distributio f impuritno ycrose ionth n si s sectioe th f no torus from these measurement wels s a possibles a l .

Obviously f informatioo t lo a , loss ni t because each channel integrate radiatioe sth n along a linear domain, and it is impossible to restore the original distribution exactly. We e thereforar e tryin o t approximatg e distributioth e n function (x,.y) froe th m measurements on the channels of the cameras

using a given class of functions F(x,y,pjf} containing parameters pjf. The functions a)i(x,y) contain informatio measuremene th n no t setup (geometry assumee , ar etc. d )an d parameterf o t se fino e t dth s i s m thaknowne ai b te o t minimisTh . erroe eth r function

<•>

This means, that we are trying to minimise the measurable difference between the original distribution functionapproximatinn a d an , g function belongin a prescribe o gt d clasf o s functions. The integrals containing the functions a>i(x,y) must be evaluated numerically. If F(x,y;pj() functions are linear functions of their parameters, i.e.

then durin iterativn ga e minimisatio erroe th f rno functio parametere nth takee b t n nou sca of the numerical integrals containing the characteristics of the measurement setup, and these integrals performee havb o et d only onc eacr efo h measurement setu choicd pan f eo base functions e minimisatioTh . n will therefor a quadratie tha b f e o t c functioe th f o n parameters, whic simpls hi computationalld ean y efficient r cas problee ou e th n I . , thamis t there is no set of base functions which corresponds well to a localised distribution moving crose inth s sectiotokamake th f no naturaA . l o dimensionachoictw a t efi woul o t le db gaussian distributio measurementse th o nt ,

F(*,y,Pk) -i———~= exp 2 a/

with parameter e positio e e centedistributioth th r th f f fo s o no e horizontar th n ni d an l vertical directions widthe distributioe o directionsth ,th tw f amplitudn o sa e th d n nan i , e parameter problee Th . , thamis t this function contain parameters sit s nonlinearly thid san , means, that the numerical integrals have to be evaluated in every step of an iterative minimisation of the error function. This makes fitting nonlinear functions to tomographic data extremely inefficien cumbersomed an t .

Neural networks have been use r fasdfo t measurement evaluatio plasmn ni a physics previously, including nonlinear curve fitting to experimental data [12-17]. The basic idea behind neurocomputin larga e eus numbeo t s gi primitivf ro e processor evaluato st e datn ai parallel [2,3] single Th . e processor r neuroo , n would perfor msimpla e tas f multiplyinko g eacnumbea f ho f inpuo r t valueinternan a y sb l weight value correspondin thao gt t input,

221 creatine weigheth f o glineaa d m inputsu r producind an s g some simple functioe th f o n linear sum as its output. With a large number of neurons organised into a network, the networ able whol a b perfor o et s y k a ema m complicate mose th df t o tasksimportan e On . t virtue f sucso h system , tha sis calleo s t d learning strategie utilisee b y chango dt sma e eth weights of individual neurons to adjust the performance of the network. These strategies ca usee nb teaco dt networha solvo kt probleea m using example f desireo s d outpuo t t specific inputs e architecturOn . f neuraeo l networks, thas beeha t n used extensiveld yan with calleo success e dth Backpropagatios si n Network. Thi multilayereda s si , feedforward type network, that realise a mappins g fro n n-dimensionaa m l inpu- m t n spaca o t e dimensional output space. There are mathematical theorems to prove this network to be a universal function approximator under certain conditions [2], and it is simple enough for a straightforward learning strateg r traininyfo g from example e formulatedb o t s . Another important virtue of this network (and many others as well) is its resistance to noise, its ability to perform well in a noisy environment. questioe Th arisesw nno , whether such neural networks coul trainee db o 'guessdt ' parametere th s tha conventionaa t l nonlinear curve fitting would produc tomographin eo c datao fint a n complicateothedI y . tr ry wordsma e dw , mapping, that returne th s parameters tha nonlineaa t r curve fitting would produce, give tomographie nth c dataA . database of samples of the input space with corresponding desired outputs to train the network e obtaineb y takinsma y db o dimensiona gtw l gaussian distributions with various parameters, making these generating parameters the desired output, and calculating the signals thadetectore th t s would measure fro knowledge mth experimentae th f eo l setup. Thus dat r traininfo a s readili g y available thid an s, fact make e applicatioth s e th f o n Backpropagation Network straightforward. This method of data evaluation was tried on the MT-1M tokamak. whicn o network e , hth up databas A t se ss werewa e trained separata d an , e database was used for testing the performance of the networks and to compare it to the performance of a conventional nonlinear curve fitting. A substantial amount of noise was also added to the channel values of the samples to simulate realistic experimental conditions. The results comparisoe th f o seee b Fig.3n no Fig.4n d left-hane nca an .Th . d side figures show who well the neural networks guessed the generating parameters of the samples (the desired output), while the right-hand side figures show the same for a conventional nonlinear curve fitting. Standard deviations from the generating parameters can be seen on top of each of figurese th e see b n no n e figures, th ca thae result t th I t r positio. fo s n parametere ar s slightly neurae worsth r lefo network estimate, whil resulte eth r amplitudsfo e parameters are practically the same for the conventional nonlinear fit and the neural networks. While e standarth d deviation widte th r h sfo parameter e smallee neurasar th r lfo r networkse th , structur e erroth s differentf i r o erelative th s a ,e erro r narrofo r w distribution s muci s h larger for the networks. By changing the circumstances of learning, this can be changed, and it is possible to train the network, so that the relative error of the parameters is constant. The overall performance of the networks thus makes them suitable for fast measurement evaluation, and the values returned by the networks may be used later as a startingpoint for a conventional nonlinear fit if greater accuracy is desired, reducing time needed for convergence. The price to be paid for utilising neural networks lies in training the networks. Trainin Backpropagatioe gth n Network involve nastsa y nonlinear minimisation involvinga large number of parameters. This is, of course extremely time consuming, and training networks only pays off, if there are large numbers of tomographic data to be evaluated. In r setup ou computationae th , l loa traininf do networke gth s equa swa tha o t lcompletinf o t g a few hundreds of conventional nonlinear fits. Since using the networks also involves some

222 Neural network output reliability of Reliabilit f conventionayo l horizontal position SD=4.97mm nonlinear curve fitting SD=3.69mm 80 80 60 60 w 4-* 40 40 08 v 20 20 0 O 0 -20 -20 u -40 o -40 -60 1 -60 t) a, -80 o -8OA0 i __ ^_.._i. . i . i . t __^__j__ . ..i. 2 80 -60 08 -40 06 -20 04 0 2 2 00 40 0 -2 60 0 -4 80 0 -6 0 -8 j j Horizontal position of center Horizontal positio f centeno r Neural network output reliability of Reliability of conventional vertical position SD=5.61mm nonlinear curve-fitting SD=3.93mm 100 100 80 80 2 60 60 cS V 5 40 40 20 sj 20 o 1 0 0 -20 -20 |-40 -40 e -60 1 -60 -80 1 -80 1 -100 », -100 0 10 0 6 0 2 0 -2 0 -6 -100 0 10 0 6 0 2 0 -2 ^ 0 -6 -100 Vertical positio f centeno r > Vertical positio f centeno r Neural network output reliability of Reliability of conventional amplitude parameter SD=3SS units nonlinear curve-fitting SD=346 units 6000 6000 5000 5000 4000 4000 i 3000 •a 3000 o 2000 2000 u o 1000 i1000 0 0 0 2000 4000 6000 o 0 2000 4000 6000 Amplitude parameter Amplitude parameter

Performance3. Fig. neuralofthe networks conventionaland curve fitting positionon and amplitude parameters. On each of the figures the network output or the results of the curve fitting are plotted on the vertical axis against the generating parameters of the samples (desired output). Standard deviation from desiredthe output is'written top.on

223 experimentin whico t s ga h architectur moss ei t suitabl solvo et giveea n problem wiss i t i ,e inveso t t into neural network training amoun e onlth f yi tomographif o t c data exceedw fe sa times that amount.

Neural network output reliabilitf yo Reliability of conventional horizontal width SD=2.54mm nonlinear curve-fitting SD=2.66mm 60 ou

50 50 •',.£ Ul a • .-• *S"" 40 _c 40 ~ ".••^CT^""' "S • vjjS"'**: o 30 30 •J^fi-f- • o jc> AJ^V' ' w 20 20 o "2 •go " u i*?* * 3 10 10 JT ^^» • 3 I V O, n * z 0 10 20 30 40 50 60 0 10 20 30 40 50 60 Horizontal width Horizontal width Neural network output reliability of Reliability of conventional vertical width SD=2.62mm nonlinear curve-fitting SD=3.08mm 60 60

50 50 to w 40 40 ;3 8 30 I 30

O 20 20 w o 10 10

0 0 10 20 30 40 50 60 0 10 20 30 40 50 60 Vertical width Vertical width

Fig.Performance4. neuralofthe networks conventionaland curve fitting -widthon parameters. eachOn figures ofthe network the outputresultsthe or curve ofthe fitting are plotted verticalthe on axis against generatingthe parameters samples ofthe (desired output). Standard deviation from desiredthe output is-written top.on

Real experimental data was also processed by the networks. The results on a series of tomographic data involvin gpellea t injection int plasme e seee b o th b Fig.5 n n nn o ca a ca t I . seen positioe th , n parameters returne networke th y db indeeo sd d correspon pellea o dt t moving into the plasma from bellow, while the amplitude rises sharply as the pellet enters plasmae th furtheA . r tes f reliabilit o talse b oy deviseyma fittiny db verticae gth l position parameter straigha y sb t line, calculatin velocite gpelletth e th comparind f yan o , g that with the velocity obtained by a time of flight calculation. The velocities obtained from the neural network position se slightl tenb o dt y smaller tha timne f th flightha eo f o t t methode th , reasons for which are unclear. It must be mentioned, that the standard deviations of the

224 position parameters correspon o approximatelt d y hale distancth f e between adjacent measuring channels.

20 -30

10 '40

•— ' -5fn0 I o 1 0 -6 •oa S -10 a 8

33 "2Voo 4940 4980 5020 5060 > -81900 4940 4980 5020 5060 Tins]

5000

4000

3000

O, 2000 •Oo «-• 1000 a.

4900 4940 4980 5020 5060

Fig. Results5. of tomographic data processed neuralby networks. The position parameters can seenbe correspondto localiseda to distribution moving into plasmathe from bellow in the vertical direction, while the amplitude parameters correspond to a sudden increase in radiation from impurity ions.

conclusiona s A sayn ca , thae w , t neural network suitable sar fasr efo t processinf go tomographi worts i c t datai t h investinbu , g into training such networks onl f eitheyi r there are large numbers of tomographic data to be processed, (which is, however most often the case reaf i r l)o time evaluatio needes ni somr dfo e reason.

REFERENCES

Zoletnik. S 1. , S.Kalvin, Rev. Sci. Instrum. 1204 6 8 (1993) 2. Robert Hecht-Nielsen: Neurocomputing (Addison-Wesley Publishing Company, New York, 1991) 3. J.A.Freeman, D.M.Skapura: Neural Networks (Addison-Wesley Publishing Company, Yorkw Ne , 1991) 4. Albert Bos: Artificial Neural Network TooA s Chemometricsn I lsA , doctoral thesis (1993)

225 5. D.F.Shanno, Math. Op. Res. 3 244 (1978) 6. S. Zoletnik, S. Kalvin, Int. Conf. on X-Ray and Inner-shell processes, Debrecen, Hungar (19939 11 : yp ) 7. S.Zoletnik, J.S.Bakos, B.Kardon, S.Kalvin, G.Kocsis, G.Veres, 20th EPS Conf on Contr Fusion and Plasma Phys., Lisbon, p:III-l 171 (1993) 8. S.Zoletnik, J.S.Bakos, P.N.Ignacz, B.Kardon, S.Kalvin, G.Kocsis, J.Szigeti, G.Veres, Researcn o IAEM AhTC Using Small Tokamaks, Wiirzburg, p:86 (1992) 9. S.Zoletnik, S.Kalvin, G.Burger, Rev. Sci. Instrum. 65 426 (1994) 10. S.Zoletnik, G.Kocsis, G.Burger, P.N.Ignacz, B.Kardon, S.Kalvin, J.S.Bakos, Rev. Sci. Instrum. 66 2904 (1995) . S.Kalvin11 , J.S.Bakos, G.Burger, B.Kardon, G.Petravich, S.Zoletnik, Rev. Sci. Instrum. 60 2857(1989) 12. C.M.Bishop, C.M.Roach, Fast Curve Fitting Using Neural Networks, AEA Fusion Report (1991) 13. C.Bishop, I.Strachan, J.O'Rourke, et al., Reconstruction of Tokamak Density Profiles using Feedforward Networks 14. J.T.Larsen, W.L.Morgan, W.H.Goldstein, Rev. Sci. Instrum. 63 4775 (1992) 15. C.Bishop, Neural Networks for Real-time Plasma Diagnostics, IAEA TCM on the Avoidanc Controd ean Disruptiof o l Tokamaksn i , Culham Laboratory, (1991) 16. C.M.Bishop, C.M.Roach, M.G.von Hellermann, Plasma Phys. Contr. Fusion5 76 5 3 (1993) 17. L.Alien, C.M.Bishop, PlasmaPhys. Contr. Fusion 34 1291 (1991)

226 RESEARCH USING SMALL TOKAMAKS*

Report on the IAEA Technical Committee Meeting, held at Ahmedabad, India Decembe7 6- r 1995

T.J. DOLAN International Atomic Energy Agency, Vienna

INTRODUCTION — An m = 2 fluctuation could result in a large magnetic island (Dhyani et al.). This meeting was not restricted to tokamaks; it also include ddiscussioa somf no e stellaratorse Th . Multiple-tip Langmuir probes were also useo t d main parameter experimente th f so s discussee th t da obtain information about the low frequency fluctu- meeting are listed in Table I. About 50 people took e plasmationth f o sa densit d potentialyan , results part in the meeting, with 26 presentations made by compared with thos ASDEXn o e a similarit d an , y participants fro 2 countries1 m . Summariee th f o s found between the two experiments. A technique was six sessions follow, based on input from the session developed to visualize the radial structure of fluctua- chairmen. tions with two dimensional plots. It was pointed out thaelectroe th t n temperature effect shoul takee db n into account when discussing the electric field from Session 1 floatine th g potential measurements. (Stoecke t al.le )

SAWTOOTH OSCILLATIONS, FLUCTUATIONS Studie sawtootf so h relaxatio e SINth n Pni toka- AND ENERGY CONFINEMENT mak found an additional mechanism on the top of the Chair: J. Fujita tearin1 = gm mode. Stron s puffingga g stabilized the sawtooth crash authore Th . s believe that, when e energTh y confinement tim n TUMAN-3(Mi e ) islane th d grows sufficiently, independen inversionf to , was deduced from the stored energy measured by some fine scale magnetic turbulence is generated that diamagnetic s probesfounwa dt I . thae oh.rn.ith t c cause crashcase e sstronth th f eo n .I puffings gga , this H mode was realized and that the energy confine- turbulenc dampeds ei mod1 . Thoug = stils ei m l e hth ment time strongly depende plasme th n do a current sawtootcause e th th f t eo active no hs i crasht i , . (Hui and input power, but showed a weak dependence on et al.) the plasma density, in contradiction to neo-Alcator scaling. The authors stated that the energy confinement corresponded well to H modes in JET and Session 2 DIII-D discussione th n I . , direct measuremente th f so LOW-ASPECT-RATIO TOKAMAK, stored energy (besides diamagnetic probes) were rec- MAGNETIC LIMITER, PLASMA ROTATION ommended. (Kornev et al.) AND SUPERCONDUCTING TOKAMAK DESIGN Magnetic fluctuations and their correlation were measure e CASTOth n i d R tokamak with movable Chair: R. Amrollahi magnetic probes and Langmuir probes for each shot. It was found that: A plasma in a pure toroidal magnetic field was biased with electrode produco t s a radiae l electric — Magnetic fluctuations are high in the core. field. It was found that the resulting poloidal rota- — Correlation length is .comparable with the tion, measured by the difference of ion saturation cur- machine radius, 6 to 7 cm. rents upstrea downstreamd man , acte suppreso dt s

*Summary publishe Nuclearn di Fusion, (19960 1 . ) VolNo , .36

227 TABLE I. NOMINAL PARAMETERS OF SOME OF THE EXPERIMENTS DESCRIBED

Device Institute City R(m) a(m) B(T) 7(kA) Remarks

Tokamaks JUST TRINITI Moscow 1.7 1.5 1.34 10000 Design stage HL-1M SWIP Chengdu 1.04 0.26 2.5 320 1 s duration START Culham Abingdon 0.34 0.27 0.1 300 Low aspect ratio SST-1 IPR Bhat 1.05 0.20 3.0 220 Superconducting design TUMAN-S(M) loffe . PetersburSt g 0.55 0.24 2 200 H mode TCA/BR Inst. of Physics Sao Paulo 0.61 0.18 1.2 120 0. 1duratios n ADITYA IPR Bhat 0.75 0.25 2.0 80 New power supply DAMAVAND AEOI Tehran 0.36 0.1 1.2 40 Elongation = 2 EGYTOR AEA Cairo 0.3 0.1 1.4 36 Startup phase SINP SINP Calcutta 0.3 0.075 2.0 25 Sawtooth studies MT-1M CRIP Budapest 0.4 0.09 1 15-30 X ray tomography CASTOR IPP Prague 0.4 0.085 1 12 Fluctuation meas. TBR Inst. of Physics Sao Paulo 0.3 0.08 0.4 12 Magnetic limiter BETA IPR Bhat 0.45 0.15 0.1 low Pure toroidal field Stellarators LHD NIFS Toki 3.9 0.6 3-^ low Under construction URAGAN IPT Khar'kov 1.0 0.125 1.3 low RF studies

fluctuations, perhaps via a velocity shear mechanism, t 1.2a t bu. Operationo t s attainenwa t gda C yi val- which may be relevant to tokamak H Mode theories. ues of nearly 1.0, in contrast to larger-aspect-ratio (Jain) devices. With gas puffing, densities approaching the Greenwald limit were observed. Durin densitw glo y The SST-1 steady state superconducting toka- operation, runaway electrons were not observed. The mak is being designed at the Institute for Plasma operating regime is broader than in conventional Research, Bhat, India, wite maith h n parameters tokamaks, and is bounded by internal reconnection being show s planne i Tabln i nt I . eI o havdt n a e events, rather tha curreny nb t terminating disrup- elongatioo 2.0t 7 a ,triangularit 1. f o t o n 4 0. f o y tions. addeNexe b o t d wil NBIe goolb e .Th d results f LHCDW o M W 2 f ICRH0. o M , 5 W 0. ,M 0.7 1 , of this experiment bode wel further fo l r development of ECRH and 0.8 MW of NBI. The main objectives of low-aspect-ratio tokamaks. (Ribeir t al.oe ) arestabliso et e scientifihth c basi steadr fo s y state operatio tokamaksf no studo t ; physice yth diverf so - Helical perturbations were produced in the TBR tors, radiativ targetses layerga d demonstrato st ;an e plasma by two resonant helical windings (m/n = 2/1 steady state heat remova particld an l e exhaust con- 4/1d an ) plus four ergodic magnetic limiter ringn si trol; and to develop plasma confinement improvement the edge plasma (connecte2 7/ produco dt = n em/ and advanced tokamak configurations. For a hydro- 3 19 effects) e ergodie effecth f Th .o t c magnetic limiter e ar V ke 5 1. = T d gean n 10x plasma 2 m~ = n , was generally to reduce MHD activity, in contrast to expected. (Saxen t al.ae ) effectsresonane e th th f o t helical windings. (Vannucci Smale Th l Tight Aspect Ratio Tokamak (START) et al.) has demonstrated the advantages of low-aspect-ratio plasmas, such as the attainment of high plasma cur- The TCA tokamak was moved from Lausanne, toroidaw rentlo t sa l fields plasmas It . s hav elond eha - Switzerland, to Sao Paulo, Brazil, where it is being o 3.0t gation2 , 1. triangularitie f o s s less than 0.8, reconstructed and diagnostics prepared, with the twice neo-Alcator energy confinement, n ~ 0.5 x objective studyinf so g Alfven wave heatin curd gan - 20 3 10 m~ , Te(0) ~ 1 keV, 0(0) < 20% and (/?) = rent drive, confinement improvement, disruptions and 3.9%. Safety factor limits were simila o predict r - turbulence. The first plasma is expected in the sum- e ERAT tionth cod D f t o sflo/a e O MH = 1.6a , mer of 1996. (Nascimento et al.)

228 Session 3 The EGYTOR tokamak, which was operated previously in Diisseldorf, is beginning operation in PLASMA HEATING AND CURRENT DRIVE Egypt. It will be used to study plasma stability Chair: J. Stoeckel regimes, plasma-limiter interaction diagnosticd san s development. Other plasma experiment Egypn i s t HL-1e Th M tokama modifies kwa 1992-199n di 4o t 'AtonA includM 1 ' plasme eth a focus deviced an 8 0. , obtain better plasma access and better position con- thetm 5 a3. pinches pincheZ , glod san w discharges. feedbaca tro y b l k system. (Den t al.)resulga e s .A t (Masoud) of these modifications new operations have produced the following results: The SINP tokamak has an improved power supply that uses a slow bank system designed to work — Boronization of the vacuum vessel has reduced together with the original fast bank. The main effort the radiative losses dramatically. is to increase the plasma current duration. (Ray — The density fluctuation level is significantly et al.) higheinboare th t a r d midplan eoute thath -t na board midplane. e constructioTh e superconductinth f o n g Large — When lower hybrid current drive (400 kW, Helical Device (LHDe s wel)i th t l a unde y wa r 2.45 GHz) was applied to the ohmic target National Institut Fusiof eo n Science (NIFS Japann )i . plasma, plasma confinement improved. (Wang Comprehensive plasma diagnostics and heating sys- et al.) tems are being developed, and the first plasma is The ohmic capacitor bank of the ADITYA toka- expected in 1998. (Fujita) mak was replaced by converter-based power supplies e maiTh ne Join th goa f to l Upgraded Spheri- to increase the plasma current and discharge dura- cal Tokamak (JUST) will be plasma burn investigation. Efforts are now being made to reduce perpen- tion. Becaus s veraspecit w f ylo eo t ratio cape th , - dicular magnetic fields that appea limitine b o rt e gth ital costs and required power consumption will be plasma discharges. (Atrey et al.) low compared with conventional tokamaks. A pre- The SINP tokama s beeha k n operated wita h liminary study of the JUST tokamak concept design safety factor q — 1.5, by using a fast plasma current has resulte parametere th n di s show Table n ni Th . eI ramp during startup. There were some indications of plasma wil l auxiln hava elongation d ea - an 5 2. f no confinement improvement whe nradiaa l electric field iary heating power of 15 to 20 MW. It is expected was created by a circular electrode inserted into the attai o 1-2.t — 5nQ wit hfusioa n burn duratiof no core plasma and biased at —50 to —350 V relative to abou. (Azizos 0 1 t t al.ve ) wallse th . (Ghos t al.ee ) Production of dense plasma by RF waves at frequencies below the ion cyclotron resonance in the URAGAN-3M torsatron was studied numericall experimentallyd yan . Several antentypeF R f so - s havna e been teste o evaluatt d e their efficien- cies for plasma production. The recently proposed Session 5 'crankshaft' antenna has the best characteristics for plasma buildup e firsTh .t applicatioe th f o n PLASMA THEOR DIAGNOSTICD YAN S crankshaft atenna in URAGAN-3M resulted in a Chair: AzizovE. plasma density of about 1019 m~3. (Plyusnin et al.) Experimental transport data from TFTR and TEXT were analysed usin hypothesie gth s that electron heat diffusivity depend e produco th tw n so f o t functions, one of them depending on local parameters and the other on the global profile. On the basis of Session 4 this model steady state solutions were obtainedd an , their stabilit discusseds ywa . (Da t al.se ) STATUS OF RESEARCH IN VARIOUS COUNTRIES From numericaa l simulatio Raleigh-Tayloe th f no r Chair:DengW. X. instability it was found that the instability saturates by self-consistent generatio sheaf no r flow even under DAMAVANe Th D tokama s achievekha elonn da - a condition where an infinite amount of gravitational gatio 2 with f o n, shaping coild passivan s e loops. energy is available. The final states were coherent (Amrollahi) even though the initial stage was random. (Das et al.)

229 insensitivite Th electro f thermayn o io d nan l trans- The Japan Ministry of Education, Science, Cul- port to shear reversal was discussed in terms of the turSportd ean s (Monbusho guess )ha t professorships, weak stabilizing influenc sheaf eo unstabln ro e modes. academic exchange agreements, grants in aid for visi- The result was found to be consistent with TFTR tors, foreign graduate students and international con- data. (Sen et al.) ferences Internationa,e sucth s ha l Toki Conference Fro ma stud f fluctuationyo e presencth n i sf o e and the International Congress of Plasma Physics self-consistently generated sheared poloidal rotation (9-13 September 1996). The Japan Society for Pro- it was found that the poloidal velocity shear had motion of Sciences sponsors fellowships (from a few n oe instabilit th effec n e fluctuationso th t f yo t bu , months)0 1 weeko t p ,u s Asian seminar- co d an s that the derivative of the poloidal velocity was more operative research programmes. (Fujita) important. For-negative derivative fluctuatione sth s Iran can provide housing to guests and small were destabilized, while for positive derivatives they stipendstudentD Ph o st s coming from other coun- were stabilized. (Singh et al.) trie goinr so otheo gt r countries. (Amrollahi) The effect of RF fields near the ion cyclotron range China has an economic development plan with a of frequencies on the drift tearing mode was studied. goal of 1 kW/person energy availability. They antic- The ponderomotive force from radial gradients of the ipate a generating capacity of 400 GW in the year fielF R d energy modifie innee sth r layer solutiond san 2000. That leave thaW shortagsa G t wil0 e b 80 l f eo havn ca a stabilizine g influenc efielF R whed e nth needed to meet the goal. The China National Nuclear energy has a decreasing radial profile at the mode Power Commission is planning a strategy towards rational surface. (Uriquijo et al.) future needs. Fusion-fission hybrids are under consideration. A transport bifurcatio no reverst modeL r e fo l shear mode transition was studied. Solutions of the (1) China will support Velikhov's ideas, such as a model equations with fluctuation levels and den- joint Asian fusion experiment. sity gradients proportiona bootstrae th o t l p current (2) Chin s conducteaha d fusion research since eth showe equilibrio dtw a states wite on ,h high fluctua- largea s 1950ha , experienced san d team applican A . - tion levels and low bootstrap current (L mode) and IAEe tioth bees Ao nt ha n mad supporr efo foro t t m another with zero fluctuation level and large boot- a fusion technology and tritium technology centre at strap current (R/S mode) r positivFo . e sheae th r Hefei. They would lik havo et e visiting scientists come L mode equilibrium was stable and the R/S mode to China with IAEA support. was unstable, while for negative shear the R/S mode (3) Foreign visitor e welcomesar : there have been was stable. (Avinas t al.he ) pase 50th tn 0i decade . (Deng) Neural networks were use r fasdfo t evaluatiof no In Africa, funds were cut for the South African tomographic measurement MT-1e th n Mso tokamak tokamak, and the Libyan tokamak is not operating. in order to study the transport of injected impurities. The Egyptian tokama abous ki staro t t t operations. A 16-channe P camerMC l a with various filters wa s With Sing Lee from Singapore they are consider- measurementse th use r dfo . Though neural networks ing an African-Asian association for plasma training, trainee havb o eeact r dfo h experimental arrangement and trying to get support for an international centre accommodato t geometrye th o et , theprovidn yca e for plasma focus studies, involving abou coun0 2 t - accurate inversions quickly. (Demeter et al.) tries. A 'small tokamak association' is also needed. The Tll-M tokamak attaine dfourfola d increase They would like to have an IAEA co-ordinated of RF heating efficiency by means of a boronized first research programme to activate work on directed modH wald e an loperation . (Kova t al.ne ) topics, such as edge plasma physics measurements. (Masoud) Session 6 Brazil has a history of international co-operation. In 1978, Simpson from Australia helped build INTERNATIONAL COLLABORATION e firsth t Brazilian tokamak, TBR-1, which began Chair: P.K. Kaw operatio 1980n ni e nexTh .t tokamak, TBR-2s wa , planned in the 1980s with help from four Chinese In this session, participants described international scientists n 1989I . joina , t Brazilian tokamas wa k collaboration activitie varioun si s countrie disd san - planned t fundedno t .bu , The Swise tokanA th s-TC cussed possible future actions. Russia is organizing maobtaines kwa shipped dan Brazilo dt the w yNo . 'Asian a n Foundatio Fusior nfo n Research', basen do have two scientists from Russia. The University of proposaa l from Velikhov. Joint fundin nucleaa r gfo r Campinas has a tokamak from Japan, and INPE has fusion experiment, a sphericasuc s ha l tokamaks i , started buildin low-aspect-ratiga o tokamak. Scien- being considered. Several participants agreed with tists fro USAmthe , Russia, United Kingdom, Italy the desirability of such collaboration. (Azizov)

230 and Germany have visited the plasma group at USP Internationae ing th Moscon si t a d wan l Congresn so on many occasions. (Vannucci) Plasma Physics Internationae Th . l Plasma Research Networ utilizee b y spreao kdma t d information about China, India, Korea and Brazil especially feel the this co-operation. need for fusion power, owinenerge th go t y demandn si these countries. Participation by non-ITER countries CONCLUSIONS difficults i ITEe iA nth RED . They nee develoo dt p personne interestee l skillsar possibilite d th an , n di y A broad spectru f papermo presenteds swa , both of a joint experiment among non-ITER countries. An theoretica d experimentalan l , ranging from recent International Fusion Experiment (IPX) could be built dat carefullo at y analysed results. Some interesting satisfo t y various possible set goalsf so , sucs ha new results (such as from START) and techniques (suc neuras ha l network analysi tomographif so c data) (1) Steady state advanced tokamaks, to study were reported e maiTh . n proble r mosmfo t partici- heat removal, fuel exhaust, current drive, confinement pants was trying to do state of the art research on improvement, densit pressurd yan e limits. low budgets (2) A pure ignition experiment, to study fusion There is a worthwhile role for small tokamaks and burn, tritium control and remote handling. other fusion experimento st (3) Low-aspect-ratio tokamaks, to study the advantage sucf so h configurations with regarr fo , dto — Test theories e effect, th sucf plasmo s a sh a example, beta, stabilit transportd yan . rotation, (4) A 250 MW pilot plant, as proposed by Dean, — Check empirical scalings, suc densits ha y limits; Kadomtse otherd van s — Develop diagnostics, suc tomographys ha ; (5) A fusion materials development project, to — Train personne n i plasml a physicd an s stud activatiow ylo n materials, wall damage, radia- technology skills. tion effect superconductorn o s d insulatorsan s , tri- Centre f excellenco s e developinar e n severai g l tium breeding and hybrid blankets. There are also countries, forming strong technical base r futurfo s e other possibilities (Kaw). Multiple fusion research research projects. Better visibility and public sup- machines coul buile db parallen i t l (Sen) none Th . - pore needear t d Increased international collabora- oil-producing nations hav e moseth t urgent neer dfo tion may be useful to facilitate the exchange of ideas, energy sources like fusion, but their opinion makers joint fundin largf go e experiment increased san d pub- are not well informed. (Arvinash) lic awarenes e importancth f o s f fusioeo n research. A working group was formed to continue the dis- Steps are under way to develop such increased cussio internationaf no l co-operatio t futurna e meet- collaboration.

I !®f t BLANK I r v ±' '^^J»^"T?5 ~^.'7jg ?^T----rTJi 231 LIST OF PARTICIPANTS

Ahalpara, D. Institut Plasmr efo a Research, Bhat, Gandhinagar - 382424, India

Amrollahi, R. Plasma Physics Department, AEOI . KaregaEnN f do r Ave., Tehran, Islamic Republi Iraf co n

Atrey, P.K. Institute for Plasma Research, Bhat, Gandhinaga 382424r- , India

Avinash, K. Institut Plasmr efo a Research, Bhat, Gandhinagar - 382424, India

Azizov. E , TRINITY, Moscow, Russian Federation

Azodi, H. Plasma Physics Department, AEOI En Nortf do h Karegar Ave., Islamic Republic of Iran

Balakrishnan. ,V Institute for Plasma Research, Bhat, Gandhinagar - 382424, India

Bannur Institut Plasmr efo a Research, Bhat, Gandhinaga 382424r- , India

Bhatt, S.B. Institut Plasmr efo a Research, Bhat, Gandhinaga 382424- r , India

Buch, B.N. Institute for Plasma Research, Bhat, Gandhinagar - 382424, India

Chaturvedi. S , Institut Plasmr efo a Research, Bhat, Gandhinaga 382424- r , India

Chavda. C , Institut Plasmr efo a Research, Bhat, Gandhinagar - 382424, India

Chenna Reddy. ,D Institut Plasmr efo a Research, Bhat, Gandhinagar - 382424, India

Das. A , Institute for Plasma Research, Bhat, Gandhinagar - 382424, India

233 Demeter. G , KFKI-Research Institute for Particles, Departmen Plasmf o t a Physics, Budapest, XII, Konkoly Theg 29-33t eu , Budapest-P.O.B. 49, Hungary

Deng, X.W. Southwestern Institute of Physics, 432P.Ox Bo ., Chengdu-610041, China

Deshpande, S. Institute for Plasma Research, Bhat, Gandhinaga 382424r- , India

Dhyani. ,V Institute of Plasma Physics, Czech Academ Sciencesf yo , Za Slovankou 3, P.O. Box 17, Prague-18211, Czech Republic

Dolan, T.J. International Atomic Energy Agency, Wagramerstrass0 10 , P.O e5 x Bo . A-1400 Vienna, Austria

Fujita, J. National Institut Fusior efo n Science, Nagoya 464-01, Japan

Ghose, J. Saha Institut Nucleaf eo r Physics, Block-AF, Sector-I, Bidhannagar, Calcutta-700064, India

Gupta, C.N. Institut Plasmr efo a Research, Bhat, Gandhinaga 382424r- , India

Hariri, A. Laser Research Center, AEOI, P.O11365-8486x Bo . , Tehran, Islamic Republi Iraf co n

Hui, A.K. Saha Institute of Nuclear Physics, Block-AF, Sector-I Bidhannagar, Calcutta-700064, India

Ivanov, N.V. Kurchatov Institute, 1 Kurchatov Sq., Moscow - 123182, Russian Federation

234 Jain, K.K. Institut Plasmr efo a Research, Bhat, Gandhinaga 382424r- , India

. R Jha, Institute for Plasma Research, Bhat, Gandhinaga 382424r- , India

Kartashev, K.B. Kurchatov Institute, 1 Kurchatov Sq., Moscow -123182, Russian Federation

Kaw, P.K. Institute for Plasma Research, Bhat, Gandhinagar - 382424, India

Kornev, V. IOFFE Physico-Technical Institute, 26, Polytechnicheskaya, St. Petersburg -194021, Russian Federation

Kulkarni. S , Institute for Plasma Research, Bhat, Gandhinagar - 382424, India

Kumar, A. Institut Plasmr efo a Research, Bhat, Gandhinaga 382424r- , India

Masoud, M.M. Plasma Physics & Fusion Department, Atomic Energy Authority, 13759x Bo P.O. , Cairo, Egypt

Mattoo, S.K. Institute for Plasma Research, Bhat, Gandhinaga 382424r- , India

Mirhashemi, S.M. Departmen Internationaf to l Affairs, AEOI, North Karegar Ave., Tehran, Islamic Republi Iraf co n

Pathak, H.A. Institut Plasmr efo a Research, Bhat, Gandhinaga r382424- , India

Plyusnin, V.V. Institute for Plasma Physics NSC "KhIPT", Kharko v310108- , Ukraine

235 Prabhakara, H.R. Institut Plasmr efo a Research, Bhat, Gandhinaga 382424- r , India

Pujara, H.D. Institut Plasmr efo a Research, Bhat, Gandhinaga 382424- r , India

Ramachandran. ,H Institute for Plasma Research, Bhat, Gandhinaga 382424r- , India

Ran)an, P. Institut Plasmr efo a Research, Bhat, Gandhinaga 382424r- , India

Rao, C.V.S. Institut Plasmr efo a Research, Bhat, Gandhinagar - 382424, India

Ray, R. Saha Institut Nucleaf eo r Physics, Block-AF, Sector-I, Bidhannagar, Calcutta - 700064, India

Ribeiro, C. Culham Laboratory, Abingdon, Oxfordshire OX 14 3DB, United Kingdom

Saxena, Y.C. Institut Plasmr efo a Research, Bhat, Gandhinaga 382424r- , India

Sen. A , Institute for Plasma Research, Bhat, Gandhinaga 382424- r , India

Sen. S , Institut Plasmr efo a Research, Bhat, Gandhinaga 382424r- , India

Sethia, G.C. Institut Plasmr efo a Research, Bhat, Gandhinaga 382424r- , India

Singh, R. Institut Plasmr efo a Research, Bhat, Gandhinagar - 382424, India

236 Stockel, J. Institut Plasmf eo a Physics, Czech Academy of Sciences, Za Slovankou 3, P.O. Box 17, Pragu e18211- , Czech Republic

Urquijo, G. Institute for Plasma Research, Bhat, Gandhinagar - 382424, India

Vannucci, A. Inst Fisicae .d , Universidad Pauloo Sa e e,d Sao Paulo - C.E.P. 053889-970, CP 66318, S.P Pauloo Sa . , Brazil

Vasu. ,P Institute for Plasma Research, Bhat, Gandhinaga 382424r- , India

Wang, E.Y. Southwestern Institute of Physics, P.O. Box 432, Chengd 610041u- , China

237