J. Plasma Fusion Res. SERIES, Vol.3 (2000) 51-57

Heating and Current Drive System for Tore Supra
Steady-State Operation
EQUIPE TORE SUPRA presented by BECOULET Alain
Association EUMTOM-CEA sur lafusion contr6l6e, CEA Cadarache, 13108 St Paul les Durance, FRANCE(Received: 18 January 2000 / Accepted: 16 June 2000)

Abstract
A major upgrade of the Tore Supra heat and particle exhaust capability, known as the CIEL project[P. Garin et al.,this conference] and installed during the year 2000 shutdown, will raise the heat exhaust
capability to 25 MW (convected + radiated) steady-state, and the pumping capability to -4 Fa.m3/s, andwill allow Tore Supra to investigate high power steady-state plasma operation, with target discharges up

to 25 MW-1000 s. Consequently an upgrade of the wave based heating and current drive system isstudied (ICRH, LHCD, ECRH/CD), in order to reach this target, in terms of input power and current
drive capability, but also in terms of current and pressure profile control, in order to address relevant"advanced tokamak" physics, including steady-state discharges balancing LHCD and bootstrap current,
and possibly exhibiting internal transport barriers. The optimization of antennas for such high power longpulse operation is also discussed.

Keywords:
Heating and current drive steady-state operation ICRH LHCD ECRH

1. lntroduction
The Tore Supra tokamak has been operating
successfully its superconducting toroidal magnet foralmost 12 years (Ro = 2.37 rl, a = 0.80 m, B < 4 T,

series of 6 gyrotrons is under construction, raising thecircular cross-section). Successful long pulse operation
has been achieved during the past years, using radio-

particle exhaust capability of Tore Supra (CIEL project,plant capability up to the target of 2.5-3 MW for 210 s

in the coming years.

Following the major upgrade of the heat and[2]), the present paper reviews the proposed plan for thewave based heating and current drive (H/CD) upgradedsystems, motivated by physics R&D, including thedefinition of the various steady-state target plasmas (lowfrequency (RF) systems, namely ion cyclotron resonant
heating (CRH) and lower hybrid current drive (LHCD).
Among the performances reached by Tore Supra up tonow, one must underline long pulse discharges reaching
2 minutes, long pulse zero loop-voltage discharges
reaching 75 s, combined heating long pulse discharges
reaching 4 MW coupled during 60 s, and high power

density/high current/L-mode or high density/low
current/improved confinement discharges) and a
discussion on the delicate problem of long term power
coupling issues when using wave FVCD methods.

short pulse discharges (up to l0 MW-2 s ICRH, 5.3

2. Requirements and Constraints for
Scenario Studies

MW-6 s LHCD, and 12 MW-2 s combined) [].
Recently, the new electron cyclotron resonant
heating (ECRH) power plant (working at ll8 GHz)
obtained its first results with the prototype gyrotron,

The H/CD system upgrade and the corresponding

both on dummy loads (up to 400 kWl5 s, setting a newworld record) and on plasma (up to 350 kW2 s). The@2000 by The Japan Society of Plasma
Science and Nuclear Fusion Research

5lBecoulet A., Heating and Cunent Drive System for Tore Supra Steady-State Operation

physics scenario analysis are obviously constrained byboth technological and financial considerations, out of
the scope of this paper. They are also influenced by the
specific CIEL power and particle exhaust structure, as

3. Determination of the Optimum Powerwell as by the strong magnetic ripple of theinsure a zero loop voltage operation at significant
plasma current values. Non inductive current drive

BalanceAmong the possible current sources, only LHCD
and the bootstrap effect can provide enough current tosuperconducting Tore Supra tokamak. The present study
has thus followed the rules listed here:
the plasma loop voltage is zero volt; i.e. the plasma-

effects provided by the ICRH or ECRH systems are thus

only considered as local correction effects. Both LHCD
and bootstrap current are sufficiently well documented
in the Tore Supra database, including zero loop voltage
operation, to allow us to fit their behaviour, together
with the electron and diamagnetic stored energy, in acoupled confinement-cunent prediction code [3]. Notethat the only confinement improvement terms concern

current is fully non inductive. This allows toenvisage steady-state operation. However transient

-

-

-

positive or negative loop voltage can be envisaged.
the steady-state injected power is in the range 20-

25 MW. The regimes presented here thus
correspond to the maximum heat removal

the well-documented enhancement due to the increase ofcapability expected on Tore Supra.
the IVCD methods envisaged are ICRFVCD, LHCD
and ERCH/CD. The magnetic ripple of Tore Supra
prevents to envisage a massive input of neutralbeam injection (NBI) power. Only one neutralbeam injector (1.5 MW, 80-100 keV) is kept toprovide probe fast neutrals for diagnostics (MSE,
NPA, CXRS, ..). Seven ports are devoted to H/CD
systems: one port for the already installed ECRH
antenna bringing 2-3 MW in the plasma, six ports
to be shared by 3 MW-cw ICRH antennas and 4
MW-cw LHCD multijunction grills. Attention mustbe paid on long term coupling issues, as well as on
fast ion and electron ripple losses.
R - 2.4 m, O.69 m < a < O.72 m. The optimisationof heat and particle exhaust by the 360' bottom
limiter gives the new plasma configuration.

the mid-radius magnetic shear resulting from asignificant bootstrap current fraction [3]. No further
improvement due to a possible internal transport barrier(ITB) triggering was considered in the zero-D approach,though observed sometimes [4]. Finally, in cases using

ICRH minority heating, the presence of a fast ion
population is included in the total diamagnetic energy as
well as in the global energy confinement time.
Using this model, the optimum balance between

ICRH and LHCD was investigated, assuming that
ECRH brings 2-3 MW of heating power. The magnetic

field is 3.8 T and the ICRH scheme is minority
Hydrogen in Deuterium. The plasma current is bounded
between Ip = 0.5 MA (below which operation is delicate
in Tore Supra), the edge q-value of 3 and the Greenwald
density limit. Similarly, the volume averaged density isbounded between <n> = 1.5 x 10le m-3 (lower limit foran efficient and reliable ICRH coupling) and theGreenwald limit within the constraint that the edge qvalue is limited to 3. The resulting operation space, in- finally, the system optimization must offer
sufficient flexiblity for current and pressure profile
modifications, in order to study advanced tokamak
physics on long term operation.

To fulfill those constraints, the scenario study has
been performed using two distinct approaches: a zerodimensional (0D) approach based on the present ToreSupra H/CD database, and a more sophisticated one
dimensional (lD) self consistent computation including

terms of plasma current versus volume averaged density

power and current deposition codes on plasma equilibria-

is shown on figure l, in the situation of 3 LHCD grills(12 MW-cw) and 3ICRH antennas (9 MW-cw).
Fig 1. allows to define two categories of non
inductive target plasmas:

together with transport simulations and MHD stability
predictions, in order to access the final steady-state
regimes. The 0D model is used mainly to determine theoperating space versus the power balance between H/
CD systems and guide the hardware choice, though thelD model investigates the inner consistency, including
the present FUCD and transport models.

high current-low density plasmas (quoted as"CIEL1" in Fig. l). These discharges are L-mode
type with low (< 207o) bootstrap current. A volume
averaged electron temperature of 4.6 keV is
expected, Zeff - 2.4 (deuterium plasmas) and Bp -1.2. The relatively low density however willprobably not allow to take full advantage of thelimiter pumping capability, thus possibly leading toa non steady-state operation in terms of density
control.
52

Becoulet A., Heating and current Drive system for Tore supra Steady-State operation-

varying the plasma composition, either in terms ofHydrogen minority concentration, or seeding thedischarge with a low Helium3 concentration, one
can operate in ICRH regimes with low temperature

ion tails, favouring bulk ion heating andconsiderably reducing the ion ripple losses. This isconsidered as a "safe" option with respect to ripple

problem. Mode conversion heating and current-

drive will also be possible.
prescribing the loop voltage to positive or negativevalues allows both to study the influence of theedge current profile on the confinement and MHD
properties. Note that negative loop voltage regimes(at the same given input power) also allow toinvestigate transiently higher fto-values.
- a high field side multiple-pellet injector (10 Hz,
1000 s) is also envisaged to fuel the plasma. This
option would modify the density profile, as well as
the bootstrap current profile.
This situation in terms of power balance appears tobe the best compromise, for less LHCD power would
not allow to open up the whole range of plasma current
at zerc loop voltage ("CIEL1') and more LHCD power(lOtop-l;
Fig. 1 Operating space at 3.8 T, 12 MW LHCD, 9 MW
ICRH, 2 MW ECRH. The grey area refers to the
ICRH coupling low density limit.

current-high density plasmas (quoted as- Low*CLEL2would reduce the performance of the non inductive
operation at high density high bootstrap fraction(,CIEL2"\.

in Fig. 1). These discharges, operating at
the Greenwald density boundary, exhibit a high(-45Vo) bootstrap fraction together with aconfinement enhancement (an H factor of 1.8 withrespect to the ITERLgT -tot predictive law is
expected). A volume averaged electron temperature

4. lmplications of Edge RF Physics
Before starting the lD analysis of the two mainof 2.2 keY is predicted, Zeff - 1.7 (deuterium

the physics R&D exercise generated by the very highwave power involved.
Concerning IRCH, locating the power deposition

target plasmas, one must discuss the extra constraints on

plasmas) and B11 - 2.2. Note that the confinement
times of CIELI and CIEL2 are finally very similar(-0.07 s), though the plasma current of CIEL2 is

imposes to position the (hydrogen) ion cyclotron
resonance layer somewhere in the plasma by tuning thealmost half of CIELI. CIEL2 will allow to study
the stability of "advanced" regimes on long term

proper generator frequency. This frequency tuning has

however two major boundaries, which become realconstraints when dealing with long term high power

operation. The high density is also more compatible

both with the particle pumping capability and ahigher fraction of radiated power (-40Vo) requiredat that high level of injected power.
On the top of those two regimes, several otheroptions are open, mainly due to the flexiblity in terms ofinjection:-

ICRH frequency (40-80 MHz):- lowering the magnetic field down to 3 T or 2 T,
one can access the operating domain of the ICRH
direct fast wave electron heating (FWEH), andstudy regimes free offast ion in the discharge. The
antenna phasing capability and the corresponding
FWCD effect will also allow us to study the role offirst, the ion cyclotron resonance layer must locate
in the so-called "ripple good confinement area", i.e.
ICRH must drive fast ions which banana tips stayin the plasma core region, where no trapping in the
ripple magnetic well is possible. In a machine like

Tore Supra, this mainly means that the ion
cyclotron resonance layer must remain on the highfield side of the magnetic axis to prevent strong

ripple losses. Moreover, low current low
collisionality operation may also enhance the ripple
stochastic diffusion effect. which must be carefullv

the central current densitv.
53

Becoulet A., Heating and Current Drive System for Tore Supra Steady-State Operation

LHCD and ICRH, using the same technology as the
CIEL limiter high flux plasma facing components [13].

estimated.

-second, no ion cyclotron resonance layer, including

high harmonic ones, must exist in the immediate
vicinity of the antenna front []. Though they do
not significantly alter the expected power

4. Self-Consistent 1D Scenario Modelling
The results from the 0D simulations have been used

it, subsequently causing damages in long pulse

as first guess of a self-consistent lD model, whereplasma equilibrium, ICRH deposition profile, LHCD,
transport and bootstrap current are coupled in various
codes run sequentially up to convergence. The following

operation.

codes have been coupled and used:

A careful experimental campaign has been-

deposition balance, those layers are responsible forsignificant heat loads on the antenna Faraday
screens when located a few centimetres in front ofconducted on Tore Supra in 1999, in the exact CIEL
configuration, to build up a database of the best

toroidal geometry equilibrium and transport:
ASTRA [4]. The transport model is the so-called
Mixed Bohm/gyro-Bohm model, including amagnetic "shear function", insuring an electrontransport reduction in the low magnetic shear

compromise for ICRH - minority heating.
A second major issue for high power long term
ICRH coupling is the presence of RF sheath effects [5]-at the antenna front level (Faraday screen, lateral
bumpers, ..). These RF sheaths generate convective cells

region [4].
ICRH deposition: PION [15].
LHCD: DELPHINE (ray tracing + Fokker Planck)tl6l.

in front of the antenna, creating both density depletion-in certain areas and density rise in other regions. The
most damagable effect is the last one as those sheaths
may accelerate ions which hit the antenna front andcause damage, as observed already on the Tore Supra
present antennas [6]. A complete model, including RF
field radiated by the antenna (three dimensional) [7] anda one dimensional RF sheath model in the (poloidal/
radial) two dimensional geometry is underway [8]. The
purpose is to help the new long pulse antenna design tominimize such effects, achieving 9 MW-cw coupled tothe plasma by three double-loop resonant antennas (4G-

Bootstrap current: NCLASS [7].

ECRH/CD and neoclassical tearing modes

stabilization codes I I 8, 19].
Though not fully reported here, MHD analyses are also
coupled, mainly using the MISHKA code [20].

4.1 The CIEL1 case
The low density/high current CIEL1 case is mainly
an L-mode discharge driven by LHCD current, the extra
pressure being given by minority ICRH. As shown on

Fig.2, and due to the high electron temperature, the ray
propagation is single pass (at least for rays with nx> 2.3
in that case). The current profile tayloring is thus easily
doable, using different spectra on different launchers("compound spectrum"). On- and off-axis current drives

80 MHz, phase tunable).

In parallel, the LHCD multijunction grills sufferfrom accelerating edge electrons in front of their mouth.
The high-n1 component of the spectrum was recognized
as responsible for this effect which, in case of highpower density, may cause severe damages to grills andconnected limiters [9-l l]. Minimization of such effects
requires a limitation of the power density at the grill
mouth (< 25 MWlm2), as well as a careful design of thegrill itself (septa shaping, careful location of guard

are accessible mixing power between n1= /.J attd n1' =2.9. Moreover, as the plasma current is essentially alower hybrid non inductive current, driven by fastelectrons, the feedback control of the steady-state
current profile by the hard-X ray measurements can beseriously envisaged. It is performed on Tore Supra by atomographic system of two cameras, using CdTe

limiters). Such improvements have been implemented
and successfully tested on the new Tore Supra [12]launcher installed on Tore Supra in 1999. Two extra
launchers of that type are planned to be installed,insuring 12 MW-cw in the plasma.
Of course, all those high power density elements

detectors, with excellent time, space and energy
resolution [21].
On the question of ICRH deposition, the only point

concerns the H-minority heating, for which PION
predicts a very energetic fast ion tail (fast ion slowing
down time -1.6 s), leading to a significant fraction of
power lost by direct orbit losses. This aspect could force

are equipped with high performance bumper protections,exhausting the conducted-convected power coming fromus to split the ICRH power on several frequency

the plasma, as well as RF losses. Such bumpers havebeen designed and already tested on plasma for both("polychromatism"), and/or to use more extensively the54Becoulet A., Heating and current Drive System for Tore supra steady-state operation

0.8

0.4

0.2

€o
N
-0.2

Fig. 3a CIEL2: Electron temperature profile (keV) versus

r/a. (-l first guess, {--) 2"d iteration, (-x-) 3'd
iteration, {-o-) 4th iteration.

R (n)Fig.2 CIEL1: example of LH rays propagation in theactual Tore Supra equilibrium. The launched
spectrum is centred on nil =2.6 (60 rays).

Helium3 minority scheme.

4.2 The GIEL2 caseOn the contrary, in the high density/low current
CIEL2 case, the LHCD absorption is less direct, andrequires 1.5 to 2 passes and a significant q-upshift totake place. This means that the LHCD deposition ismuch more sensitive to the exact pressure and overall

current profiles, fully justifying the self-consistentprocedure described here above. Fig. 3 shows the 4
iterations performed with ASTRA and DELPHINE onsuch a steady state/high density case. Note that theFig. 3b CIEL2: LHCD current profile (a.u.) vs r/a. (-) first
iteration, (-) 2^d iteration, (-x-) 3'd iteration, (-o-)

ICRH deposition was kept constant for these iterations,as the self-consistency in this case is a minor effect. Fig.
3a and Fig 3b show the electron temperature profile, andthe LHCD current deposition profile, respectively. The
bootstrap fraction is 45Vo. The first guess comes from0D considerations. One can see that, due to the off-axis

4th iteration.

beyond).

The influence of the externally imposed LHCD

LHCD deposition (compound spectrum n1 = 2 and n1 2.9) and bootstrap current, the central magnetic shear isspectrum is illustrated on fig. 4, showing the resulting
converged steady-state q-profile for two quite different

low enough to trigger an electron ITB (known on Tore

injected compound spectra. The multi-pass LHCD

Supra as "hot core lower hybrid enhanced performance"[4]), whose first effect is to strongly enhance the central

damping, the significant (uncontrolled) bootstrap
fraction together with the self consistent process linking
current profile and transport significantly reduces, for
CIELZ, the freedom of tayloring the final state. Fig. 4
indicates that the electron ITB (i.e. the shear inversion
radius) will occur in the range O.3 < rla < 0.4.
Further investigation involving in particular a betterpressure, and thus increase the central bootstrap (znd-3rd

iterations). This current profile reanangement slightly
increases the central shear, tending to weaken the ITB
and to peak the LHCD deposition (smaller nu-upshift).
An equilibrium state is reached rapidly (4s iteration and))Becoulet A., Heating and Current Drive System for Tore Supra Steady-State Operation

wfulssd/a
0.12

0.08

0.04

100 2mFig. 4 CIEL2: safety factor profile vs r/a. {-} compound

spectrum n| = 2 & hrr = 2.9, C-l compound300

Time(ms)spectrum nrl = 2'3 & n,, = 2.6.

Fig. 5 CIEL2-2T: (3,2) mode island width normalized tominor radius vs time. Stabilization by ECCD {-) 0.9

control of the bootstrap current profile using off-axis
ICRH, and/or magnetic shear modification by local
ECCD is envisaged. As an example, 2.5 MW of ECCD
(on a Ip = 0.9 MA, <n> =2lgte --3 plasma) can drive"
up to 120 kA at mid-radius , which is sufficient toMW, (-)2.5 MW, (...)2.5 MW modulated.

pulse discharges.

Next, the beta-limit imposed by the resistiveinterchange mode or the so-called infernal modes [23] isdisplace the minimum q location of Ap = 0.15 - 0.2.

known to be an issue for scenarios where the safety
factor profile is flattened or reversed in the core, when astrong pressure gradient exists close to or inside theminimum of the safety factor profile. The CIEL2 model
equilibria above are found to be stable to such modes.
However, the sensitivity of the infernal modes (lessimportantly, the resistive interchange modes) to thedetails of the profiles imply that such modes may occur
in discharges with similar features. Should this happen,
the current and pressure profile control capability of theheating and current-drive system would allow somefruitful analysis of the associated operational limit, in

4.3 MHD considerat:ons
MHD analyses are performed a posteriori using thelD simulation output. Though not fully performed yet,
one can however give some interesting conclusions.

First, concerning disruptive MHD (linear stabilityof kink, ballooning and interchange (ideal and resistive)modes), the CIELI equilibrium is found to be operating

far from the limits, and only seriously affected by thepossible sawtooth activity (enhanced by the presence offast ions), if the target q-profile contains a g = I surface.
Let us recall that this giant or monster sawtooth activity

also affects the wave power coupling by repetitive
strong perturbation of edge conditions. The ICRH

reproducible steady-state discharges

Concerning "softer" events, CIELI and CIEL2
seem to be mostly immune to neoclassical tearing
modes (NTMs). However, should this occur, the ECRW
CD system is able to stabilize such modes (2,1) or (3,2).
As an example, fig. 5 shows the time evolution of suchan (3,2) island, pre-established in a CIEL2-2T discharge
and submitted to either 0.9 MW, 2.5 MW or 2.5 MW
synchronized with the island rotation. In all cases, thestabilisation of the island is achieved in 300 ms.
Finally, in scenarios with high input power fromthe ICRF system, it is worried that the excitation of
Alfvdn Eigenmodes by high-energy ions could degrade

system is however designed using pre-matching stubs inorder to cope with those sudden transients of coupling.
The CIEL2 scenario is found to operate close toboth the high-n ballooning and low-n kink limits. The
stability to low-n kink modes is actually improved by
the presence of the close fitting wall. Taking the wall
into account moves the stability limit about 5OVo above
the target beta. Then, in the absence on any significant
momentum source, the CIEL2 scenario is found tooperate close to the stability limit of the resistive wallmode (RWM) [22], which could be observed in long56

Becoulet A., Heating and Cunent Drive System for Tore Supra Steady-State Operation

the heating performance. However, it is important tonote that the capability to analyze such high-frequency
modes is limited on Tore Supra, for high-frequency

search (Proc. lSth Int. Conf. Yokohama, Japan,
1998) rAEA.

[2] P. Garin et al.,this conference.
[3] A. B6coulet et al.,PPCF 40-8A (1998) A157.
t4l X. Litaudon et al., in Plasma Physics and con-magnetic measurements are hindered by the existence ofsome upper cut-off frequency (typically 4O kHz)imposed by the iron-shell which protects the pick-up

trolled Nuclear Fusion Research (Proc. lTth Int.
Conf. Montrdal, Canada, 1996) IAEA, Vienna Vollcoils installed in the vacuum chamber.

The installation of new inner-wall components inTore Supra, being able to exhaust up to 25 MW ofpower in steady-state is claiming for a significant(ree7) p 669.
J.R.
Myra, D.A. D'Ippolito and Y.L. Ho, Fusion
t5l
Engineering and Design, 3l (1996),291.
[6] L. Colas et al., to appear in proc. of the 27th EPS
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upgrade of the FVCD capability of the machine, both in[7] S.Pecoul et al.,Proceedings of 26th EPS CCFPP,

5. Conclusion

terms of power, but also in terms of current drive
capability, and reliability at each level. Besides theMaastricht- 1999, Netherlands, P3.01 8, 1025.

t8l M. B6coulet et al., to appear in proc. of the 27th

pluri-annual technological efforts put in the upgrade of

EPS Conf., 2000, Budapest (Hungary).

t9l Y. Peysson, M. Goniche et al., in Plasma Physics

the FVCD subsystems (generators, lines, antennas, ...), aphysics R&D activity is underway to better integrate all
the physics requirements and constraints on such longterm operation plasmas (up to 1000 s). A strong effort
has been put on the edge RF physics to help the antenna

and controlled Nuclear Fusion Research (Proc. 18th

Int. Conf. Yokohama, Japan, 1998) IAEA.
tl0l V. Fuchs er a/., Phys. Plasmas 3 (1196) 4023.
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design activity to optimize the long pulse power
couplers. On the plasma scenario side, the strong zero-

loop voltage constraint, together with Tore Supra

173.

specificities, lead us to base the non inductive current on[2] Ph. Blbet et al., proc. of the 5th International Sym-a mixture of LHCD and bootstrap current, setting two
type of target plasmas: the low density/high current Lmode plasmas, mainly driven by LHCD, and the moreposium on Fusion Nuclear Technology - Rome
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t13l G. Agarici et al., proc. of the 5th International"advanced" high density/low current improved
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Symposium on Fusion Nuclear Technology - Rome
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[4] G.V. Pereverzev e/ a/., Nuclear Fusion 32 (1992)1023.

allow local FVCD effects, helping broadening the current
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complete lD self-consistent set of codes has beenassembled to investigate the final states of such
discharges, including MHD stability. A self-consistenttime dependent model is now under development which
will help us to study the routes to such final states,

ll5l L.E. Eriksson and T. Hellsten, Phys. Scr. 55 (1995)

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57