J. Plasma Fusion Res. SERIES, Vol.4 (2001) 99-105

Formation of Internal Transport Barriers
and lts lmpact on the JT-60U Plasmas
IDE Shunsuke+ and the JT-60 TeamNaka Fusion Research Establishment, Japan Atomic Energy Research Institute,

Ibaraki, j I I -0193, Japan(Received: 5 December 2000 / Accepted:27 October 2001)

Abstract
Formation of internal transport barriers (ITBs) in a plasma is a key to access to improved confinement. In JT-60U, there are two types of plasmas which are accompanied with ITB. One is a high Bo
plasma and the other one is a reversed magnetic shear (RS) plasma. In RS plasmas, interesting features
can be seen in spatial profiles of plasma parameters. For example, the location of the ITB can be strongly
related to the location where the safety factor (4) becomes minimum, and the radial electric field (8,) has

strong spatial variation at the ITB. These features can be related to the physics mechanisms of confinement improvement at the ITB. On JT-60U RS plasmas, experiments to modify the toroidal current
density or the toroidal rotation profiles, which are expected to influence the ITB structure directly or
indirectly, have been carried out. And these modifications were fond to cause chanses in the ITB
structure. The results are shown in this paper.

Keywords:JT-60U, ITB, reversed magnetic shear, LHCD, toroidal rotation

1. Introductionin the high 0o plasmas in JT-60U. Therefore, topics
concerning to the ITB structure and its modification in
the JT-60U RS plasmas are reported in this paper. In thesecond section, the tokamak, the heating system anddiagnostics which are closely related to RS experiments
shown in this paper are briefly described. In the third
section, some general characteristics of the ITBs in the
JT-60U RS plasmas are introduced. Also some typicalThe main objectives of the JT-60U project arestudies on confinement, stability, steady state operation,
heat and particle handling by a divertor configuration
and understanding of underlaying physics in order toestablish physics basis for the international thermonuclear experimental reactor (ITER) and steady state

reactor such as SSTR [1]. For this purpose highperformance plasmas, such as H-mode, high Bo t2J,
reversed magnetic shear (RS) plasmas [3] and the latterachievements in performance so far in the JT-60U RS
plasmas are shown. Following that, in the fourth section,

two combined with the H-mode [4,5], have been developed. What is essential to keep improved confinement inexperimental results on toroidal current density (j)profile modification via lower hybrid current drive

those plasmas is to maintain the internal and/or the edge

(LHCD) and its impact to confinement are presented. In

transport barriers (ITB t3,61 and/or ETB). Especially athe fifth section, experimental results on relation
between the ITB gradient and the radial electric fieldrole played by the ITBs is a key to access highconfinement. The formation of the ITB and its dynamics
can be seen more clearly in the RS plasmas than those

profile modification via external torque input control are@2001 by The Japan Society of Plasma*

C o r r e sp ondin g autho r' s e - mail : ide @ naka.j ae ri. g o.j p

Science and Nuclear Fusion Research

99

IDE S. et al., Formation of Internal Transport Barriers and its Impact on the JT-60U Plasmas

profile, the magnetic shear s = plq dqldp becomes
negative inside pq-in. Here p is the normalized flux

shown. Lastly, the summary will be presented in the lastsection.

radius.

In JT-60U experiments, an RS configuration is

2. The Tokamak and the Heating/Current
Drive and Diagnostics Systems
The JT-60U tokamak is a single null divertor

usually formed by NBI heating at fast plasma current(10) ramp-up to retard current penetration by raising 7"
and with help of the bootstrap current. However itshould be noted here that LHRF injection at the Ip ramp
phase also could form an RS configuration with quite alarger po-6 [7]. Moreover, even at the current flat-top an
RS configuration formation was observed with LHCD

tokamak device, the plasma major radius (Ro) 3 - 3.5 m,
the plasma minor radius (ar) 0.6 - 1.1 m and the

maximum toroidal magnetic field (,B,) <- 4 T at R =3.32 m. On the tokamak a variety of heating and current
drive systems are equipped. They are the neutral beam
injection (NBI), the lower hybrid range of frequencies(LHRF), the electron cyclotron range of frequencies(ECRF) and the ion cyclotron range of frequencies(ICRF) systems. Two out of the four systems, the NBI
and the LHRF systems, were mainly used in theexperiments described in this paper. Detail of the twoor ECCD although pq.rn is not very large.

Usually a moderate to a high performance RS
plasma is formed by injecting adequate NB power into
an RS configuration. Typical profiles in an RS plasma
are shown in Fig. 1. As shown in the top box, the 4

ITB footsystems are:l) The NBI system: Two types of ion sources are5

used, one is based on standard positive ion based ion
source of which acceleration voltage is 3 I 10 keV andthe other one is based on negative ion based ion sourceof which acceleration voltage is < 500 keV. The

q
1

5
4
3

15

purposes of NBIs are heating, toroidal torque input andcurrent drive. Due to higher acceleration voltage, thenegative ion source based NBI (N-NBI) is better indriving current.
2) The LHRF systems: Two types of multi-junction(MJ) LHRF launchers are installed. One which isinstalled in a port at poloidal angle 0 = 45o, consists offour rows of eight MJ modules which have 3 subwaveguides. The other one, installed in a horizontal
port, consists of two rows of four MJ modules which

T10
(keV)

ne
2 ('l01s1n-e1

u

1

0
10

0

0

have I 2 sub-waveguides.
Key diagnostics indispensable to this paper are, themortional Stark effect (MSE) measurement for thesafety factor (q) or j profile measurement, Thomson
scattering measurement using Ruby and YAG lasers forthe electron temperature (7") and the density (n.) profile

-2o1t

E.

evim;

-40
0.5
p

measurement, and charge exchange recombination

spectroscopy (CXRS) measurement for the ion
temperature (4) and the toroidal rotation (Vo) profiles

Fig. 1 Plasma parameter profiles in a typical RS plasma

in JT-60U. The top box; the q profile and theminimum (9.,") in the profile and its location

measurement.

(po.in) are schematically shown. The second box;

3. General Characteristics of Reversed
Magnetic Shear Plasmas in JT-60U
the 4 profile comes to have a minimum inside a plasma.
In this paper, the minimum value in the q profile and its

the 4, I and n. profiles. The bottom box; the 7"
and the /i profiles with the neoclassical thermal
diffusivity profile for a reference, and the E,
profile. The region of the ITB is schematicallyshown with the gray area. Also the ITB "foot"location is also indicated schematically by an

location are denoted oS 4-in and po-1". With this qarrow.

In an RS plasma, theT profile is a hollow one. Then

100

rDE s. et al., Formation of Internal rransport Baniers and its Impact on the JT-60U plasmas

profile is increasing toward the center from pn.;. In thesecond box the 7", the Ti and the n. profiles are plotted.
At a certain position in a plasma the gradient of all these
profiles changes quickly, and the profiles build up
toward the center. The position is referred to as the ITB"foot" (pioo,) in this paper. It should be noted that
although ppool could vary in each profile somehow it isfound to be strongly related to pn^6 in the JT-60U RS
plasmas [8]. However this steep gradient in the profiles

5
cn 4
o
!, 3
lrJ

2

E36639 5.4s

rr
Bp-o

C'

1

usually does not continue up to the center, but becomesmuch less steep, or even flat, at a certain point (referred
to as the ITB shoulder). A region between the ITB footand the shoulder is called the ITB region. As shown inthe bottom box, both the electron and the ion thermal=0

diffusivities (p and 1) show sudden drop at this region.
Actually in this case, 7, is found to be as small as theneo-classical value as shown in the figure. However, as
expected from the less steep temperature and density
profiles, the thermal diffusivities go up inside the ITB
shoulder. Also, the radial electric field (E,) profile isshown in the bottom box and indicates strong spatial

Fig. 2 The MSE polarization angle ly*"t

-1

n
T

-0.4 0.0

o.4

0.8(R- R"t;"/a- BrlBry vs radial

location (distance from the magnetic axis
normalized by the minor radius) of themeasurement, Almost zero yMsE in the central
region suggests almost zero Bo or j.

/s5. This effort has succeeded in producing an ELMy Hmode RS plasma which has a large/", of about 802o,variation at the ITB region.

Formation of the ITB can bring benefits from aview point of steady state advanced tokamak operation,such as high confinement and large bootstrap current
fraction to 1o ("fus). In JT-60U, there have been many
achievement in view of plasma performance owing to
ITB formation. Some of them are brieflv described aswith 1o = 0.8 MA, B,o =3.4 T, q at 95Vo flux (Qs) = 9.3
and the triangularity (d) = 0.43 [9]. In the plasma, theconfinement enhancement factor to the H-mode thermal

confinement scaling (HHgarv,zt) reached 2.1 - 2.3, andthat to the L-mode scaling (F1sqp) was 3.3 - 3.8, with thenormalized p (8") = 1.9 - 2.2. Owning to the large/r.

follow;of 80Vo and the P-NB driving the rest 20Vo of 10, fullcurrent drive was maintained with such a highly
improved confinement state for 2.7 s (about 6 x r").

l) High equivalent D-T fusion gain (081)
The RS plasmas in JT-60U have been proved tohave potential to achieve high fusion plasma performance, and in the campaign in 1998 the value of Q#
has reached up to 1.25 at Ip = 2.6 MA and the toroidal
magnetic field at the plasma center (8,q) = 4.4 T l5l.
The edge of the RS plasma was in the L-mode. In order

In addition to the performance improvement shown
above, a very interesting structure is found in the RS
plasmas, that is a current hole [10,11]. In an RS plasma,almost zero poloidal field (Bo) is observed in the central
region (Fig. 2). This in other words means that j is verysmall or even zero in the region. It should be stressedto obtain high performance, it was found to beadvantageous to form the ITB early in the 1o ramp-up
phase. In order to maintain the ITB at large radial
location which gives better performance, a certain
amount of power needs to be applied. Otherwise, the
ITB shrinks and the performance becomes lower or iseven lost. On the other hand, since the current keeps
penetrating into the plasma center, the minimum of thesafety factor (4) profile (4-1) keeps decreasing.
2) High confinement RS plasma with H-mode edge
One of the key concept in the advanced tokamak
operation is to utilize bootstrap current as much as
possible for non-inductive operation [1]. On JT-60U,
optimization of an RS plasma has been pursued to raisethat this finding suggests a possibility of a stabletokamak operation with very small or even zero toroidal
current in the core region.

4. Modification of the Current Profile andthe ITB Location
As noted in the previous section, the ITB foot isfound to be strongly related to the position of 4-,n in JT60U RS plasmas. It was also found that expansion of
Pq-in (or - Proo,) gives improvement of confinement
performance [12]. Therefore it is an interesting question

what happens when the q-1, location is changed via
some active external scheme. Modification of the ITB

l0t

IDE S. et al., Formation of Internal Transport Barriers and its Impact on the JT-60U Plasmas

location by modifying target q profile by means of

2A,1

LHCD was already reported [7]. Also quasi-steady statesustainment of the RS configuration and ITB by external

on Pua(mhlo's

LHCD was demonstrated [13]. On the other hand, quasisteady sustainment of the RS configuration and ITB

0;-(Mw)

fu1

Pr-n
5 (Mw)

00

with high bootstrap current fraction (- SOVo) was also
reported as described in the previous section [12].
However, in the former cases the bootstrap current
fraction was not as high as expected in the futurev, o'5(v) o
ne46(10tsm-o;20machines, and in the latter case there was small roomfor external NBCD to work and also 4e5 is too high in

p

0.6

5

Therefore, in order to explore current drive and culTent

6.5 7

5.5 6

7.5

Time {s)view of economical operation of a future rector.

Fig. 3 A typical waveforms of LHCD on a high p" RS
plasma at /D = 1 MA and B,= 2'5 T. The LH power

modification effect with high (but not too high)bootstrap current fraction at lower 4e5, LHCD was

was modulated 0 to 100% with the dutv of 90%.

applied to high BN RS plasmas at low qs5rcgime.

The experiment was done at Bto = 2.5 T with
deuterium gas. The waveforms of typical plasmaI6parameters are shown in Fig. 3. As shown in the figure,the target RS plasma was formed by P-NB heating.
Then Pys was reduced and kept constant for clearer

o \

cr4

measurement of the LHCD effect. It is noted that just at

I

I

I
J

t/l

\

./

\

the PNs step-down a mini collapse occurred and a large

'1

I

2

drop in Au at - 6.3 s is attributed to the mini collapse.
After P111 injection, the surface loop voltage (yt ) started
decreasing, which indicated that the inductively driven
plasma current was being replaced with non-inductive
current, that is LH driven current. In the bottom box n"
at p - O.2,0.6 and 0.8 are shown, which correspond tothe central region, near the ITB foot and peripheral
region respectively. The temporal evolution of n"at p 0.6 starts increasing at about 6.7 s, which indicates thisposition is coming into the ITB region, or in other words
the ITB region is expanding. Change in the profiles isshown in Fig. 4. In Fig. 4 (a), the 4 profiles at 6.4 and7.5 s are shown. As shown in the figure, the location of
the 4.6 shifts outward. This shift is attributed to off-axis
LHCD. In Fig. 4 (b),n. profiles at 6.5 and 7.5 s areplotted. As mentioned before, expansion of the ITB footis seen in the change in the profile. It is noted that

(D

4

I

L
=

I

f
a

CD

o
2
F
go

06

A
fl q

7

a

qn
b'!

\ J :r

? 1

0.5

K
'1
1

p

Fig. 4 Profiles of q and n" at 6,4 s/6.5 s and 7'5 s (solid

curve, circles).

reduced to 0.9 MA and also N-NB was injected. The
role of N-NB injection is not only current drive but also
effective central heating. The waveform ofthe discharge
is shown in Fig. 5. Injection scenario of P-NB and LHW
is similar to that shown previously, but N-NB injection
started prior to LHCD by 0.2 s. As shown in the fitgure'V7 dropped to zero and then even negative, showing
full current or even over drive. Similar to the previous

although the central density decreases, the expansion of

the ITB foot compensates to keep high electron
population in the core region. Furthermore it should be
noted that expansion of the ITBs in the Z" and T1 profiles

is also observed.
In the discharge, V t rcached very closed to zerobut not completely, as shown in Fig. 3. Demonstrationof full current drive in this regime is very interestingand important in relevance to the ITER steady state
operation. In order to access full current drive, 1o wascase, temporal evolution of n" near the ITB footsuggests expansion of the ITB foot. As the results of the
to2IDES. et al., Formation of Internal rransport Barriers and its Impact on the JT-60u plasmas

tF

lp

o'u
(MA}

Momentum Source

10 PilBF

0=o

Prt 4l
(Mw)

(Mvv)

2l

(v)

0H
vl 0.5 L

2 Woi"
1 {MJ)

f.
0r-

(v)

E

3

4E

z

C\I

b

0

ne
1019m -3)

e'F.
0tJ
5.5

6

6.5

7

7.5

Time(s)Fig. 5 Wave forms of the discharge in which full currentpdrive was achieved simultaneously with highconfinement (HHsae,rtl - 1.4 at high normalized

Fig. 6 Typical torque (momentum) input profile,electron density (n"lno*1. Also in this shot Pr" wasmodulated 0 to 100% with the duty of 90%.

balanced, co- and counter directional, in the

discharge shown in Figs. 7 and 8.

ITB expansion, the confinement in the plasma was
improved further and some of the plasma parametersu,=)ft,+va Bo -v, Bo .

Q)

reached to the region which is relevant to the ITER with

Here, pi, Zi, ni, Vq and Ve are the pressure, the charge

improvement factor to the H-mode (HHs8e2) is about

number, the density, the toroidal and the poloidal
rotation velocities of an ion species, and Bp is thefull current drive condition. The confinement

1.4 and, the fraction of line averaged density (4) to theGreenwald density (z6yy) is about 0.82. It should be
stressed that simultaneous achievement of full current
drive at high confinement and high density is a new
regime in JT-60U operation. Although ess = 6.9 in the
discharge is still not very low, it is approaching to therange that is expected in the ITER steady state operation

toroidal magnetic field. As noticed in the equation, there

would be a room to modify the first and the second
terms externally. That is, control of the heating tomodify the first term and the control of the toroidal
torque input to modify the second term. On the other
hand, the third term is expected to be determined mostly

by neo-classical force balance. The idea is to utilize fourtangential P-NB lines on JT-60U to control torque input

scenario, that is 4e5 = 4 - 5.

profile to act on the radial electric field via modificationin the second term in the eq. (2).

5. Modification of the Toroidal Rotation
Profile and the ITB Strength

The experiment was carried out with Ip = 1.5 MA
and 8, = 3.5 T and the working gas was deuterium [8].

As described in the second section, a large spatial

variation in the radial electric field profile is often
observed in RS plasmas in JT-60U. TheoreticalIn the discharge, the input toroidal torque profile waschanged by changing combination of tangential beam

predictions show that competition between the shearing

rate (ars,s) and linear growth rate (y) of a micro-

units. Three input momentum profiles that were used inthis experiment are shown in Fig. 6. The combination

turbulence is a key in the turbulence suppression. Here,o)s,s is written [4];which gives co-directional torque is referred to the"CO" injection. On the other hand, the combinationd)o,o=(RB,)'+? E,
dyl RBo
B

which gives counter-directional torque is referred to the"CTR" injection. The combination which gives almost
zero torque relative to the other two cases is referred tothe "BAL" injection.
The target RS plasma was initially formed with the
BAL injection, then switched to the CO injection, then(1)Here, B6is the poloidal magnetic field and y= constantis a magnetic flux surface. Therefore, spatial derivativeof E, (it will be expressed as Ei later on in this paper)
would be a key parameter. On the other hand the radial
electric field is written as;

the CTR injection and finally to the BAL injection
again. Since the interest is on the ITB structure, the103IDE S. el al., Formation of Internal Transport Barriers and its Impact on the JT-60U Plasmascogradient of 4 at the ITB is taken as a measure of the
ITB characteristics. In Fig. 7, the temporal evolution of

CTA

BAL

E

the d gradient at the ITB is shown, and the combinationof the torque input is also shown schematically at the
top of the figure. As indicated in the figure, atjust after
6.25 s the Z; gradient started decreasing, and atjust after

o
L

60
40

f 2A
c

6.65 s the gradient started increasing again, or
recovering. Note that the NB power was kept almost
constant. The confinement improvement (l/sep) was
evaluated as 2.2, 1.6 and 1.8 at 6.25,6.6 and 7.0 s
respectively. The change in the toroidal rotation profileis shown in Fig. 8 (a). As shown in the figure, in thecase of the CO injection Va builds up in the central
region while the profile flattens at the ITB foot justbefore the degradation in the I gradient starts (- 6.25s). On the other hand, in the case of the CTR injection
followed by the BAL injection, a flat V6 profile at the
ITB foot comes to have a positive gradient just before
the recovery of the Ti gradient starts. The gradient of E.

Cg(t, o6'I

6.5

Time (s)7Fig. 7 Temporal evolution of the fi gradient at the ITB

when torque input profile was changed as "CO"+ "CTR" =+ "BAL".

ll

E

at each timing was calculated and shown in Fig. 8 (b).
Together with E',, gradients of the first and the second*Vp" and Vq
terms in eg. Q) are shown, labeled withrespectively. As seen in Fig. 8 (b), towards thedegradation in Ii gradient ,what changes most clearly inth]dE

the Ej profile is a decrease in E', at the ITB foot. On theother hand, when Z1 gradient is about to recover,increase of Ei at the ITB foot is observed. Both ofthem

Fig. 8 (a); The V, profile at 5.9, 6.25, 6.45 and 6.6 s in the

discharge shown in Fig.7. (b); The gradient of E

profile and its terms dominated by Yp and Vr,

are attributed to the change in the "Va" due to change inthe Va profile. These results suggest that the E, gradientlabeled with "Vp" and "Vr" respectively, at 5.9,
6.25,6.45 and 6.6 s in the same discharge.

at the ITB foot is a key parameter to control theconfinement of whole the ITB resion.

lower hybrid waves the location of the ITB footexpanded outward. By combining the off-axis LHCD

6. Summary
General feature of the reversed magnetic shear
plasma in JT-60U are descried;

and the current drive by N-NBI, full current drive of /o =0.9 MA with high confinement of HHss1r21of about 1.4
at high density normalized to the Greenwald density rc"/o the location ofthe ITB foot is strongly related ton61ry of about 0.82 and qes of 6.9.

the location at which q becomes minimum,

The other one is conceming to modification of theE, profile via control of the toroidal torque input inorder to affect the transport characteristics of the ITB. Ito strong spatial variation in the E, profile is often
found to be formed at the ITB region,

was demonstrated that by changing the toroidal torqueinput profile by choosing combination of four tangential
NB lines the I gradient at the ITB changed. As theresult of the modification in the Vr profile, the E. profileand thus the gradient of E, profile was changed. It was
fond that prior to the degradation of the Zi gradient, the. the temperatures and density profiles often
become flat in the central region.

Two topics relating to the first and second feature aredescribed in this paper. The first one is concerning to
the current profile and the ITB location. It was shownthat by extending the location of the minimum 4 bymeans of non-inductive off-axis current drive using

E, gradient at the ITB foot became small due toflattening of the Vr profile at the ITB. On the other
hand, prior to the steepening of the d gradient increasero4

rDE S. et al., Formation of Internal rransport Barriers and its Impact on the JT-60u plasmasof the E, gradient at the ITB foot was observed. These
results suggest that the E, gradient at the ITB foot is a
key parameter to control the confinement of whole the[5] T. Fujita, Y. Kamada, S. Ishida et al.,Ntcl. Fusion
39, 1627 (1999).
t6l Y. Koide, M. Kikuchi, M. Mori er a/., Phys. Rev.

ITB region.

Lett.72,3662 (1994).

[7] S. Ide et al., in the Sixteenth IAEA Fusion Energy

Acknowledgments

Conference, Montrdal, I 996 (IAEA, Vienna, I 997),

The authors would like to express our gratitude toVol.-3, pp.253-264.

all the researchers who have contributed JT-60U under at8l Y. Sakamoto, Y. Kamada, S. Ide er a/., Nucl.

wide range of collaborations for their valuable

Fusion 41, 865 (2001)[9] T. Fujita, S. Ide, Y. Kamada et al.,Phys. Rev. Lett.

discussion, suggestion and comments.

87,085001 (2001).

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105