J. Plasma Fusion Res. SERIES, Vol.3 (2000) 3-9

Progress in LHD Experiment
MOTOJIMA Osamu*, OHYABU Nobuyoshi, KOMORI Akio, KANEKO Osamu, YAMADA Hiroshi,
KAWAHATA Kazuo, ASHIKAWA NaoKoI, CHIKARAISHI HirotaKa, DEVRIES Peter, EMOTO MasahiKo,
FUNABA Hisamichi, GOTO Motoshi, HAMAGUCHI Shinji, IDA Katsumi, IDEI Hiroshi, IMAGAWA Shinsaku,
INAGAKI Shigeru, INOUE Noriyuki, IWAMOTO Akifumi, KADO Shinichiro, KITAGAWA Shiro, KUBO Shin,
KUBOTA Yusuke, KUMAZAWA Ryuhei, MAEKAWA Ryuji, MASUZAKI Suguru, MINAMI Takashi,
MITO Toshiyuki, MIYAZAWA Junichi, MORISAKI Tomohiro, MORITA Shigeru, MURAKAMI Sadayoshi,
MUTO Sadatsugu, MUTOH Takashi, NAGAYAMA Yoshio, NAKAMURA Yukio, NAKANISHI Hideya,
NARIHARA Kazumichi, NISHIMURA Arata, NISHIMURA Kiyohiko, NODA Nobuaki, KOBUCHI Takashir,
OHDACHI Satoshi, OKA Yoshihide, OSAKABE Takeshi, OZAKI Tetsuo, PETERSON Byron, SAGARA Akio,
SAITOH Kenji2, SAKAKIBARA Satoru, SAKAMOTO Ryuichi, SASAO Hajimer, SASAO Mamiko,
SATO Kuninori, SATO Motoyasu, SEKI Tetsuo, SHIMOZUMA Takashi, SHOJI lviamoru, SUZUKI Hajime,
TAKAHATA Kazuya, TAKECHI Manabu2, TAKEIRI Yasuhiko, TAMURA Hitoshi, TAMURA Naoki,
TANAKA Kenji, TOI Kazuo, TOKUZAWA Tokihiko, TSUMORI Katsuyoshi, TSUZUKI Kazuhiro,
YAMADA Ichihiro, YAMADA Syuichi, YAMAGUCHI Satarou, YAMAMOTO Satoshi2, YANAGI Nagato,
YOKOYAMA Masayuki, WATANABE Kiyomasa, WATARI Tetsuo, HAMADA Yasuji, ITOH Kimitaka,
MATSUOKA Keisuke, MURAI Katsuji, OHKUBO Kunizo, SATOH Sadao, SATOW Takashi, SUDO Sigeru,
TANAHASHI Shyugo, YAMAZAKI Kozo, OHTAKE Isao, AKIYAMA Ryuuichi, HABA Kiichiro,
IIMA Masashi, KODAIRA Jun-ichi, TSUZUKI Tetsuya and FUJIWARA Masami
National Institute for Fusion Science, Toki, Gifu-ken, 509-5292, JapanlGraduate University
Advanced Studies, Hayama, 240-0193, Japanfor

2Department of Energy Engineering and Science, Nagoya University 464-8603, Japan(Received: 25 February 2000 / Accepted: 19 June 2000)

Abstract
In the third campaign (O7.13.99-12.14.99), NBI (<4.2 MW), ICRH (<1.4 MW) and ECRH (<0.9MW) were used to heat the plasma in LHD. The carbon tiles were installed as divertor plates, resulting insignificant reduction in metal impurity concentration (Fe). The maximum field used in the experiment is

2.gT.IJpgrading of the key hardware systems led to (i) higher T" [T,(0) = 4.4keY atn"= J.J x 1018m3 and Po6,

= 1.8 Mwl, (ii) higher confinement [%= 0.3 s, Z"(0) = l.l keV at n"(0) = 6.5 x l0re m-3 andPo6, = 2.0 MWl, (iiD higher stored energy W, = 880 kJ, (iv) higher volume averaged B = 2.4Eo (at B = 1.3

T). The temperature pedestal enhances the confinement significantly (50Ea higher than the ISS95
scaling). The ICRH heating was found to be as efficient as NBI heating. Long pulse discharges (80 sec
NBI shot, 68 sec ICRH shot) have been achieved without any technical difficulty.

Keywords:
toroidal confinement device, helical device, plasma heating, divertor, high temperature plasma,confinement time, pedestal*

Corresponding author's e -mail: motojima@ LHD.nifs.ac.jp@2000 by The Japan Society of Plasma
Science and Nuclear Fusion Research

Motojima O. e/ a/., Progress in LHD Experiment

1. Introduction
The Large Helical Device (LHD) is a large
heliotron type superconducting device with divertor [ =2, m = 10, Ro, (position of magnetic axis) = 3.6-3.9 m, a(minor radius) = 0.6 m, B = 3 Tl. The LHD experiment
began in March 1998 after its eight-year construction

Table 1 Plasma parameters achieved during the 3rd
campaign (1999).

T"(0) 4.4keY
T,(0) 2.7 keYHigh Electron Temperature t"[-

31. The major goals of the LHD experiments are: (i)Physics experiment from which break even condition
can be extrapolated (nTr5 > 1020 keVm-3s (Q-0.35), I >10 keV, F > SSo), (ii) demonstration of advanced
toroidal plasma operation, i.e., disruption free steadyT'(0) 3.5 keV
rE 0.09 s
P"o" 3.9 MWn"1.0 x lOre mJ

High Ion Temperaturestate operation of current less helical plasma withstored energy are compared with tokamak H-mode
database in Fig. l. Considering the smaller heating
power and plasma volume, the present data of the LHD
are very promising.. In this paper, we describe the key
physics areas with good progress.

2. ICRF Heating

rt(0)tT"(O) = 2.0 x 10re keVm-3s
Maximum Stored Energr 880 kJ

1.6-2.0keY, Prcpp = 1.25 MW and such a plasma wasmaintained stably for 5 seconds by ICRF power alone
without any impurity problem. Moreover, when ICRF
power of 1.3 MW is added to the NBI (1.4 MW) heatedplasma, the stored energy increased from2TOkJ to 420
kJ, accompanying electron temperature rise of 0.5 keV.
Such heating experiments clearly demonstrated that theheating efficiency was similar to that of NBI heating.

3. Generation of High Temperature Plasmas
To achieve high central temperature, we used RF

heating. Central ECH heating (168 GHz, second
harmonics) requires B = 2.9 T for the inward shifted
configuration (R* = 3.6 m). The temperature profiles ati" = 5 x 1018 m-3 are shown in Fig. 2. The ion
temperature, measured by Doppler shift of the impurity
line was 2.7 keY. The electron resonance heatine

2.4Vo atB,= l.3T

Highest Beta
(volurne averaged)

4

Maximum Density

107

1.1 x l02o m-3

O ASDEXI

tr AUGO JET

helium with hydrogen as a minority species. When
ICRF power was applied to ECRH generated target
while the density was kept nearly constant. The plasma
parameters were: i" (line averaged electron density) = 5
x 1018 m-3, Wo (plasma stored energy) = 200 kJ, Z,(0) =T"(0) l.l keV
xs 0.3 s
P"* 2.0 MW
r1(0) 6.5 x 10re m3

High Confinement

Up to now, NBI has been the main power source toheat the LHD plasmas. But in the 3rd campaign, we
executed successfully ICRF heating experiment (1.3
MW, 38.47 MHz). The gas used in the experiment is

plasma, the stored energy and the temperature increased

1.8 MW

5.3 x 1018 mr

T"(0) 3.3 keV

divertor.

In Table I we summarize the plasma parameters
achieved during the 3rd campaign of the LHD
experiment in 1999. Achieved data points of the plasma-0.06 s

P""
n

106

a

&

" JT6OU" Dlll-D

48 I

.

4

^-&

Ar

N JFT-2M
LHD

=

*gi

10-

@
1Qr

ffi
lon L
100

H-mode Database

1or

102

ve (m3)Fig. 1 Achieved plasma energy of LHD in comparison
with Tokamak H-mode database,resulted in more peaked electron temperature profile
compared with the profiles of the NBI heated discharges
and the central value was as high as 4.4 keV. Figure 3
shows the central electron temperature achieved withpresently available RF power as a function of thedensity. At very low density (n" < 2 x l0l8 m-3), thetemperature became even higher. With combination of
NBI and ICRF heating, we achieved the central ion

Motojima O. et a/., Progress in LHD Experiment#16112#16149t= 0.7 s

t= 1.0 s

aT.

oT.

o

I

o
.Y

-Y

F

oT

3

oT

4

o

z
o

F

2

^T.

1

0rr

-1.5

-1

-0.5 0

1

0.5

0

'l.5

-1.5

.,|

p

oo

4

8fu^
.f,&:p|F

2

1.5

Fig.4 lon and electron temperature profiles ofa

and ICRF.

0.,

0.5 1

p

Fig. 2 Temperature profile of a discharge heated by ECH-v-0.5 0discharge heated by NBI and ICRF.

or*(}\ -" ooqnOO{n*a\l$r'1t9{ \J\t \bq
O O-

O

O

O

O

o

10-1

o"

xo

IJJ

0L
0.0

.t.0

0.5

P
1.5

1o'2
n

e

1101e m-3;

Fig.3 Density dependence of the central electron
temperature of the discharges heated by ECH and
ICRF.

10-3
1

temperature of 3.5 keV and the profiles are shown in
Fig. 4. The electron temperature was found to be nearly
equal to that of the ions. The edge temperatures at p
(normalized radius) = 0.85 were as high as 1.3 keV[4,5]. Such high edge temperatures are believed to result

0-3

1o-2

1ol

r.rsses ts)
Fig. 5 Comparison between experimental values ofenergy confinement and those predicted fromlSS95 scaling. (tokamak data from ITER H-mode
Database, Ver.5)

in improvement in confinement over the stellarator
scaling (ISS95) t4,61.

4. Energy Gonfinement Scalingdischarge with the inward shifted configuration (R- =3.6 m) exhibits a factor of 1.6 improvements over theISS95 scaling and the confinement is comparable to

Early experimental results have shown that theenergy confinement time of the LHD is significantly
better than the value predicted by the scaling based ondata set from the small heliotron typed devices I2,3,61.

those of ELMy H-mode tokamaks. The observed highedge temperature, not seen in the smaller heliotron

We have accumulated the confinement data andcomparison with other devices is given in Fig. 5. The

experimental goals is to maintain such a level of
improvement factor or even to enhance improvement

devices resulted in higher stored energy and hence
higher energy confinement. One of the main

Motojima 0. et a/., Progress in LHD Experimentfactor in the near future experiment with high heating.tl. ..

3

5. Edge Transport Barrier
Edge transport barrier was observed in the LHD
discharges [4,5]. The barrier or pedestal forms during
the plasma pressure increasing phase, not through a'!. tt

o
Jz

. atl

F
O^n

"bo oo R^
vun

rapid transition as in tokamak H-mode. With beamI
vc)

heating on, the hot plasma region expands radially andeventually reaches the last closed magnetic surface(LCMS) and divertor plates. During this process, I"r"d

t1o
o r"*

(a)

power.

o6(the electron temperature at the shoulder of the pedestal,

a

Oo

ooob

360

24

n ltottm-t;

P (MW)(p = O.gS-0.9) increases naturally, forming an edge
temperature pedestal. A clear pedestal can be seen in
Fig. 4. The estimated total thermal conductivity (n"X)
there is fairly low, typically l.O-2.0 x lOre m-rs-r andthus transport in this edge region can be called an edge

Fig.6 Parametric dependences of T"0 and T"n"othermal transport barrier. Unlike tokamak H-mode
barriers, the particle (including the impurity ions)

Ll2lt = | surface or its associated island could play amajor role in formation of the edge pedestal since thetl2tt = 1 surface was always located within or justoutside the pedestal region for the configurations tested
so far. It is also important to notice that the inherent

confinement does not appear to improve andfurthermore there is no sign of the ELM.
For the low-density discharges, the temperature atpedestalthebecomes as high as 1.3 keV. The width ofthe pedestal (A) averaged over the flux surface is foundto be 6 cm, which is much wider than that of thecomparable tokamak. One of the amazing features of the
LHD edge barrier is that the pedestal temperature isfound to be close to the averaged temperature (<7">)[the temperature ratio (T"p'dl<T">) can be as high as
0.81. Thus l7o is almost proportional to n"T"p"d and thus
the edge confinement almost determines rs (the global
energy confinement time). The density dependences ofL(0) and Teped (at the fixed input power) are depicted inFig. 6(a). Both central and pedestal temperatures
decrease gradually with increasing density, leading tohigher stored energy at higher density. When n" is below
3 x l0re m'3, T"p"d exceeds I keV. Fig. 6(b) shows power
dependences of Z"(0) and Teped for a fixed density(n-4.4 x l0re m-3). Both temperatures increase rapidly

with input power when P is below 2 MW. In higher
power regime, however, the increment of T"o"d *i17
power is modest. This is the main issue, which we haveto solve experimentally. Optimistically, further
increased power in the near future experiment could
naturally lead to the better confinement regime or mode.
With modification of the divertor from open to closed

configuration, we are planning a more drastic edge
control (combination of the pellet or beam fueling andhigh efficient pumping) for confinement enhancement.
A possible scenario of pedestal formation is that the(a) Density dependence (P = 3.3 MW, I = 2.75T1,(b) Power dependency (n"= 4.4 x 101s m'3, B = 2.75

T).

features of the LHD magnetic geometry such as"reverse" shear, high shear at the edge and shortconnection length between good and bad curvature
regions may be key factors for the barrier mechanism.
Better understanding of the barrier certainly requires

more detailed experimental study together withtheoretical works.

6. Radial Electric Field
The radial electric field is the key parameter for
confinement of the collisionless helical plasmas [7]. The
transition from ion root (negative radial electric field) toelectron root (positive radial electric field) was observedin plasmas with NBI heating at the low density of 0.4 1.0 x 10re m-3 in LHD. The radial electric field isderived from the poloidal and toroidal rotation velocity
and pressure gradient of Neon impurity measured withcharge exchange recombination spectroscopy using

radial force balance. Since the toroidal rotation wasdamped to less than 10 km/s due to the toroidal
viscosity and the effect of diamagnetic drift velocity was
also small due to the highZ of Neon, the radial electricfield was mainly determined by the poloidal rotation

velocity. The radial electric field was negative (ion root)at the electron density larger than 1.0 x 10re m-3, while it
became positive especially near the plasma edge at the
low density below 1.0 x l0re m-3. Figure 7(a) shows thedensity dependence of radial electric field near theMotojima O. el a/., Progress in LHD Experiment

o2.4

s
5

0

- -fr--

5

-l

o
o

II

d)

I

o

-5- T-

tr

ct)

(U
L

o

n (10''m'')'x,"*:;;'"
crstron r@t

n.(0){.6r10 '' . '1:n.(0H.9r10 " .\-_

t{

O t.a

E

NU I

(b)

r

,

Magnetic Field (T)p0Fig. 8 Maximum achieved B value as a function ofmagnetic field.

-10

0.0

0.2

0.4 0.6p

1.0

t.2

Fig. 7 Density dependence of radial electric field near
the plasma edge (p = 0.9) and (b) radial profiles ofradial electric in the density region at thetransition from ion root to electron root for theplasmas.

plasma edge (p = 0.9) for the plasma heated by NBI

with the magnetic axis of 3.75 m. In these plasmas, thecentral ion temperature was close to the central electron
temperature (TVIOOVo of central electron temperature).
The edge radial electric field increased up to 15 kV/m inthe electron root as the electron density was decreased
to 0.4 x l0re m-3, while it was almost constant and to be-5 kV/m in the ion root in the wide range of electron
density of 1.0 - 3.0 x lOre m-3. The transition from ion
root to electron root was observed at p > 0.8 and there
was no large radial electric field observed in the plasmacore (see Fig.7(b)). Since anomalous heat transport

dominated over the neoclassical transport in ourdischarges, the transition was not accompanied by any
observable change in confinement.

7. Achievement of High p Plasma and MHD
Stability
In the 2nd campaign, the target plasmas weregenerated by second harmonics of ECH (84 GH2,82.6

GHz) and then heated by NBI power. The magnetic
field is strictly restricted to be 1.5 T. But we found thatwith well-conditioned wall NBI alone can initiate
plasma with appropriate gas puffing and heat the plasma
even without ECH [8,91. This is a new plasma initiation
scheme and allowed us to operate the discharges at any

magnetic field (above B = 0.5 T). In the 3rd campaign,high p plasmas were achieved in the low field
discharges, as depicted in Fig. 8 where the achieved pvalues at the various fields are plotted. At B = 1.3'1. B
value estimated by diamagnetic signal reached 2.4Vo
with help of pellet injection. With gas puffing, slightly
lower value of beta, 2.l%o was maintained for I second.
The achieved value was limited by heating power, notby any kind of MHD instability. The inward shifted
configuration exhibited higher confinement. In the LHD

design phase, it has been believed that such aconfiguration is unstable to the ideal interchange modeeven though it has better particle orbit properties andlarger plasma volume compared with the standard
configuration (R*= 3.75 m). We did not observe thatany MHD activity influenced the plasma transport eventhe maximum p value of 2.4Vo was achieved. However,small magnetic fluctuations with various modes
appeared as the F value increased (Fig. 9). Even though

the amplitudes were small, but their appearance seems
to be consistent with linear analysis of the interchange

modes. Non-linear analyses of these modes aredefinitely needed for understanding of the modes.

8. Deposition Profile of the Divertor Heatand Particle FlowsThe LHD discharges are equipped with open
helical divertor configuration. In the lst and 2nd
campaigns in 1998, all plasma facing materials were
stainless steel including divertor plates In the 3rd
campaign, the carbon tiles were used as divertor plates,
resulting in significant reduction in metal impurity (Fe)Motdima O. et al., Progress in LHD Experiment':--.- #t5o4,2 (9>2%'--+#r 1708 (p<r%)

(,
.N

6
c
c

20 40 6080probe position (mm)

^ft&

e*tr
^
&?

?6'

60

940

..fat*
0

0.5

-t'

1.5

1.0

2.O

20

2.5
bU

Fet"l

9ao
-l'

Fig. 9 Change in mode amplitude as a function of p.

20

Various MHD mode has been observed and theamplitude approaches about 10-0. Strong
instabilities which affect plasma confinement
have not been observed so far.

and its associated radiative power. Mainly electrostatic
probes diagnose the divertor plasma. Since the LHD

d (mm)

Fig. 10 Heat deposition profiles on the divertor plate forlow p and high P /,z2%l discharges. (b), (c) The
connection length on the divertor plate for zeroand 2.27o p- configurations. (Calculated by field
line tracing code).

divertor magnetic configuration has a 3-dimensional
structure, the divertor/SOl structure is not a simple
layer as in tokamak. In some cases, the split layers witha few cm gaps strike the divertor plates. An example isshown in Fig. 10. The connection length, defined as afield line length from a position on the divertor plate tothe other divertor plates is calculated by field line trace

R=3.6m B=2.757
I

E t.o

proportional to the line averaged density and hence highrecycling divertor plasma observed in the tokamak(strong non-linear dependence of the divertor density on

the line averaged density) was not observed. It isinteresting to see whether such divertor plasma behavior
persist even in the high power experiment.

o

0.0

0
0.4

o.4

3

0.3

=
c!

o.2

many layers for high p configuration. This is confirmedby the probe measurements (Fig. l0(a)). The divertor

at the high-density regime, the divertor density is almost4(

T.o (TS)'-

code (Fig. l0(b, c)). It predicts that two split layer
structure for low p- configuration becomes one withtemperature and density measured by probes on thedivertor plates were typically 5-40 eV and 0.1 - 5.0 xl018 m-3, respectively. The maximum operational
densities were 1.2 x 1020 m-3 for the pellet fuelleddischarge and 9 x l0re m-3 with He gas puff alone. Event'17170
n

0.1

0.0

P

:d

- OV

-

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0.2

0
!

0.0

0204060

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e

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0

0.0
0.3

a
I

E

0.1

o

P

o.2

OV

0.1

0.0

2;
o
1<

J0

om406080e100

Time (s)Fig. 11 (a) Long pulsed discharge heated andmaintained by ICRF heating alone, (b) Long
pulsed discharge heated and maintained by NBI
heating alone.

9. Long Pulse Discharges
Demonstration of steady-state helical plasma withhigh quality is one of the major goals in the LHD
experiment, In the 3rd campaign, we installed carbon

tile divertor plates, which made long pulse discharge
possible. Figure I I shows the time evolutions of two
discharges, heated by ICRF (l MW) and NBI (0.5 Mw)

Motojima O. et a/., hogress in LHD Experimentrespectively. They lasted more than I minute. In bothdischarges, the electron densities were l-2 y 16te --3
and were controlled by manual gas puffing. The electron
temperature was kept to be around l-2 keV. The
radiation power is less than 20Vo of the input power andno sign of the impurity accumulation. These discharges
were terminated by minor hardware problems of theheating systems. In near future, we plan to operate evenlonger discharges with higher input power.

10. Summary
LHD experiment is progressing well with improved
plasma parameters in Table I and Fig. I and many

fruitful physics results. The rapid progress of the plasma

References

tll A. Iiyoshi, A. Komori, A. Ejiri er a/., Nucl. Fusion
39, r24s (1999).

t2l M. Fujiwara, H. Yamada, M. Emoto et a/., Nucl.
Fusion 39, 1659 (1999).
t3l O. Motojima, H. Yamada. A. Komori et al.,Phys.
Plasma 6, 1843 (1999).
14)t5l N. Ohyabu, A. Fujisawa, N. Ashikawa et al., toappear in Phys. Plasma, (2000).
t6l H. Yamada, K.Y. Watanabe, S. Sakakibara et al.,Phys. Rev. Lett.84, 1216 (2000).
17l

the LHD magnetic configuration and technicalplasma physics and controlled fusion research.

O. Motojima, T. Mutoh, M. Sato e/ al., Nuclear
Fusion Supplement I (1989) 551 (IAEA-CN-50/Er-4).

parameters is realized by the disruption free property ofadvantages of superconducting device. The LHD
experiment is now exploring a new regime for toroidal
magnetic confinement with current less plasma, which is
the least free energy configuration. It will provide
valuable database with the wide dynamic range to the

N. Ohyabu, K. Narihara, H. Funaba et al., Phys.
Rev. Leu.84, 103 (2000).

t81

M. Fujiwara, O. Kaneko, A. Komori, et al., toappear in Plasma Physics and Controlled Fusion 4l

Supplement l2B, 8157 (1999).
O.
Kaneko, Y. Takeiri, K. Tsumori et al., Nucl.
tel
Fusion 39, 1087 (1999).