J. Plasma Fusion Res. SERIES, YoI.5 (2OO2) 22-27

Recent Development of LHD Experiment
MOTOJIMA Osamu, YAMADA Hiroshi, ASHIKAWA Naoko, EMOTO Masahiko, FUNABA Hisamichi,
GOTO Motoshi, IDA Katsumi, IDEI Hiroshi, IKEDA Katsunori, INAGAKI Shigeru, INOUE Noriyuki,
ISOBE Mitsutaka, KANEKO Osamu, KAWAHATA Kazuo, KHLOPENKOV Konstantin, KOBUCHI Takashi,
KOMORI Akio, KOSTRIOUKOV Artem, KUBO Shin, KUMAZAWA Ryuhei, LIANG Yunfeng,
MASUZAKI Suguru, MINAMI Takashi, MIYAZAWA Junichi, MORISAKI Tomohiro, MORITA Shigeru,
MURAKAMI Sadayoshi, MUTO Sadatsugu, MUTO Takashi, NAGAYAMA Yoshio, NAKAMURA Yukio,
NAKANISHI Hideya, NARIHARA Kazumichi, NARUSHIMA Yoshiro, NISHIMURA Kiyohiko,
NODA Nobuaki, NOTAKE Takashi, OHDACHI Satoshi, OHYABU Nobuyoshi, OKA Yoshihide,
OSAKABE Masaki, OZAKI Tetsuo, PETERSON Byron, SAGARA Akio, SAITO Kenji,
SAKAKIBARA Satoru, SAKAMOTO Ryuichi, SASAO Mamiko, SATO Kuninori, SATO Motoyasu,
SEKI Tetsuo, SHIMOZUMA Takashi, SHOJI Mamoru, SUZUKI Hajime, TAKEIRI Yasuhiko,
TANAKA Kenji, TAMURA Naoki, TOI Kazuo, TOKUZAWA Tokihiko, TORII Yuki, TSUMORI Katsuyoshi,
WATANABE Kiyomasa, WATANABE Tsuguhiro, YAMADA Ichihiro, YAMOMOTO Satoshi,
YOKOYAMA Masayuki, YOSHIMURA Yasuo, WATARI Tetsuo, XU Yuhoug, CHIKARAISHI Hirotaka,
HAMAGUCHI Shinji, HISHINUMA Yoshimitsu, IMAGAWA Shinsaku, IWAMOTO Akifumi,
MAEKAWA Ryuji, MITO Toshiyuki, NISHIMURA Arata, TAMURA Hitoshi, YAMADA Shuichi,
YANAGI Nagato, TAKAHATA Kazuya, ITOH Kimitaka, MATSUOKA Keisuke, OHKUBO Kunizo,
SATOW Takashi, SUDO Shigeru, UDA Tatsuhiko and YAMAZAKI Kozo
National Institute for Fusion Science, Toki 509-5292, Japan(Received: 2l December 2001 / Accepted:23 August 2002)

Abstract
Experiments in LHD have been developing steadily since the first plasma in 1998. Many
encouraging physical achievements have emerged during five experimental campaigns. The mostsignificant finding is that MHD stability and good transport are compatible in the inward shifted
configuration. Within the range of beta up to 3Vo and the collisionality down to 0.05, the energy
confinement keeps gyro-Bohm characteristics, and no significant degradation due to the neoclassicaltransport and MHD instabilities have not been observed. The present 5th experimental campaign with
upgraded heating resources has realized the electron temperature exceeding 10 keV which shows a signof an internal transport barrier and the ion temperature of 5 keV. Technology of large scale
superconducting system has been demonstrated successfully.

Keywords:
LHD, heliotron, net current-free plasma, inward shifted configuration, internal transport barrier,long-pulse discharge, high beta, low collisionality

C o r r e sp o ndin

I aut ho r' s

e -@2002 by The Japan Society of Plasma
Science and Nuclear Fusion Researchmail : mo toj ima @ I hd. nifs. ac.jp22Motojima O. et al., Recent Development of LHD Experiment

1. Introduction
The Large Helical Device (LHD) is a large
heliotron type device employing a full superconducting

skillful scheme of start-up enables the operation
independent of the ECRH resonance conditions andfacilitates high-B study in a low magnetic field. In thefirst three and a half years of experiment, a lot ofmagnet system. The nominal major and minor radii are
3.9 m and 0.6 m, respectively. Built-in divertor as wellas superconducting coils boost an intrinsic advantage ofnet current-fre,3 plasmas to steady state operation [-8].
The cross-section of LHD is illustrated in Fig. l, and its
device specifications and heating capability are listed inphysical achievements and findings have been
accumulated, which provide a new perspective of
attractive fusion reactor of net current-free plasmas.
Development of LHD experiments are reviewed withemphasis on the recent results from the 5th experimental

Table l. It is the largest superconducting fusion
experimental device. The major goal of the LHD
experiment is to demonstrate high performance ofhelical plasmas in a reactor relevant plasma regime.
Large dimension and strong magnetic field up to 3 T
can realize plasmas with high performance in the

campaign in 2001 in this article.

I JFT-2M(JAERI,Japan)
tr ASDEX(FRG)
O ASDEX-U(FRG)

absolute physical quantities and also enables exploration

O DIII-D(USA)in extended regimes of non-dimensional physical^ JT-60U(JAERl,Japan)^ JET(EU)o
LHD(NlFS,Japan)

parameters, i.e., low collisionality and high p. Plasma
generation is normally initiated by ECRH and then themain power from NBI and/or ICH is added to heat theplasma. With good wall conditions, however, plasma
initiation or start-up is possible by NBI alone [9]. This

107

o)
L
106

o
c
o

E

Poloidal coils

#

c)

o 105

a

104
100

EEE

t

E
101

102

Plasma volume(m3)107g) 106LocoFig. 1 lllustration of cross-section of the Large HelicalE

Device (LHD)

o)

o 105
L

a
Table 1 LHD device parameters and heating power usedin the experiment.
Major radius

3.9 m

Coil minor radius 0.975 m
Plasma radius-0.6 m

l/m

2l l0

Bo
r (0) /r (a)

2.9 T

NBI
ECRH

Helical ripple

ICRF

106

107

Heating power (W)0.35 / 1.3

Fig. 2 Operational regime of major tokamaks and LHD.
The stored energy as a function of (a) the plasma
volume and (b) the heating power. The tokamak
data is given by ITER Elmy-H mode database. RS
denotes the highest performance in JT-60U withthe reversed shear.

Heating Power (absorbed)

cr (pitch modulation) 0.1
O.2

104

8.5 MW

2.7 MWz3

Motoiima O. et al., Recent Development of LHD Experiment

2. Development of Performancecampaigns have been conducted successfully and the 5th

The LHD experiment has begun in March 1998
after its eight-year construction. Four experimental

campaign is carried out in 2001. The total cryogenic
operation time, coil excitations and plasma shots haveexceeded 20000 hours. 500 times and 30000 shots,
respectively. These experiences have demonstrated
retiability and safety of the large scale superconducting

Table 2 Achieved plasma parametersnHigh Electron Temperature l0 keV

6.0 x 10r8m-3

High Ion Temperature

5.0 keV

7.0 x 1018m-3High Confinement

1.1

kev

system. The plasma parameters are continuously
enhanced along the reinforcement of heating andparticle controlling resources. Nowaday a plasma shotwith several MW of heating power is available every 3
minutes. Achieved plasma parameters so far are listed in

6.5 x 10re m 3

re = 0.36 s, nTre= 2.2 x l}te keV --t , Puo, = 2 MW,

Table 2. Figure 2 shows the achieved stored energy as afunction of a plasma volume and heating power with

Wr= l.l2MI

Maximum Stored Energy<9> =3.2 Vo at B = O.5 T

Highest Beta

major tokamak experiments. The operation regime of
LHD lies close to the major large tokamaks and iscomparable to DIII-D and ASDEX-U. Recently high

1.5 x l02o m-3

Maximum Density

central electron temperature of l0 keV has beena Thomson Scat.
N ECE
a X-ray PHA

108obtained by localizing ECRH power deposition (a totalof 1.2 MW) at the magnetic axis (see Fig.3). The
geometrical optimization to minimize the neoclassicaltransport and well focused ray of ECRH have realized

6ro

this high temperature.

5

3. Global Energy Confinement and MHD
Stability

-t+

Ft4

d

The energy confinement times in the configuration
with the inward shifted magnetic axis (Ru* = 3.6 m) areconsistent with the ISS95 scaling [0] with anenhancement factor of -1.5 and are comparable to thosein ELMy H-mode discharges (tokamaks) [11]. The
characteristics of the energy confinement time is gyroBohm. It is a factor of 2 higher than those of smaller
heliotron devices such as CHS. Heliotron E, ATF. Thisenhancement is attributed to high edge temperature due

0.7

b
0.6

to monotonically decreasing heat conduction coefficient
towards the edge, which has not observed in mediumsized experiments [12,13]. Dependence of the energy
confinement on the magnetic configurations particularly

0.5
CO

0.4

o
r
c

0.3

the position of the magnetic axis was studied [14].

q)

0.2

Numerical calculation shows that the configuration with
the inward shifted magnetic axis (Ru* = 3.5-3.6 m) has
the best particle orbit properties, i.e., deviation of the
deeply trapped particle orbit from the magnetic surface
is minimum and the deviation becomes larger with

0.1

0r
0

0.2 0.4 0.6 0.8 1.0
pincreasing R"- [4,15]. This is important because
suppression of the neoclassical transport as well as theanomalous transport are required for high plasma

Fig. 3 Achieved highest electron temperature. (a) Elec-tron temperature profile measured by Thomson
scattering (solid circles), ECE (open squares), and
X-ray PHA (open triangles). (b) Density profile reconstructed from the line density measurement
by a multi-channel FIR interferometer.

performance. Figure 4(a) shows that the enhancement
factor over the ISS95 scaling is found to be sensitive tothe variation in the magnetic axis (R^,) [8]. The
AA

Motojima O. et al., Recent Development of LHD Experimentconfinement is improved irt Ru* = 3.55-3.6 m anddecreases with increasing Ru*. For Ru* = 3.9 m, it is as
small as 0.6. Furthermore, strong deterioration of theconfinement occurs for Ru, = 3.75 and 3.9 m when the

100

o)

o
a

plasma becomes collisionless (v* < 1, v* is defined bylrJP

the transition point of the banana and plateau regimes.)
This is attributed to the neoclassical transport. For thex R""=3.55m
o R"r=3'6m
a R""=3.65m
r Ra=3.75m^ R*=3.9m

xq)
uJ

P

inward shifted configuration with low neoclassicaltransport, the enhancement factor is independent of v*
and hence the anomalous transport dominates over the(a)neoclassical transport.

10-1
10-1

100

The average <F> of - 3 Vo at B = 0.5 T has beenachieved and the <p> value is limited by the available
heating power in the present experimental condition [8].

101

v*

It is realized in the inward shifted configuration (Ru* 3.60 m), in which magnetic hill exists in the entire
region. With increasing <P>, low n coherent modes (n/m=112, 1/l and 312) appear [16]. The mode (nlm=ll2),o 1ooU)a
IIJ
P

disappears when <B> exceeds 2.3 Vo. The pressure
profile is flat in the core and high-pressure gradiento.

xexists in the edge. The observed pressure gradient (dp/dr) exceeds the Mercier stability limit in nearly a half of

o

I.JJ

P

the plasma-confining region. Figure 5 shows thetheoretical stability diagram. At p - 0.9 (r= 1.0), it

R"*=3.6m

1r.l
0.1

becomes Mercier unstable when <B> exceeds 1.8 7o and1.Bt (%)

is even unstable to low n mode when <p> exceeds 2.1
Va. On the other hand, at p - 0.5 (t= 0.5), the observedpressure profile is unstable up to <B> = 2 Vo. Blt it
becomes stable when <p> increases further (the second
stability). The observed magnetic fluctuations, which
generally increase with <p> do not lead to any serious
MHD phenomenon which degrades the confinement
time (see Fig. a@)). The drop seen in the time trace ofhigh-B discharge in Fig. 4(b) is due to a temporalcooling by pellet injection and not due to any instability.
It is significant and very encouraging that the MHD

Fig. 4 Confinement enhancement factor over lSS95 as afunction of (a) collisionality and (b) volume aver-aged beta. The dotted line is the time trace of asingle discharge reaching <p> of 3%.

stability is demonstrated in the inward shifted
configuration, which has good transport properties.

s

4. Internal Transport Barrier (lTB) Formation
Enhancement of the confinement is the major
research efforts in the LHD program. When the ECRH

power is absorbed in the central region (p < 0.2) of NBI
heated plasmas, ITB-like temperature profile, which is

0 0.2 0.4 0.6 0.8

similar to that in CHS [17] has been observed, as shownin Fig. 6(a). The theoretical calculation based on

1

neoclassical transport suggests that existence ofelectron-root in the core is related to formation of ITB
through enhanced radial electric field [8]. This ITB-p
Fig. 5 Theoretical MHD stability diagram on the plane ofbeta and the minor radius. The parabolic pressure
profile is assumed.

like profile is obtained when the ECRH power exceeds a
25

Motojima O. et al., Recent Development of LHD Experiment

to two keV during the discharge. It should be pointed
out that recycling is not saturated in this 2-minutedischarge. The longer discharge is required to verify
necessity of an active particle control. The radiation
power is less than 25 Vo of the input power and there

threshold value (see Fig. 6(b)). This threshold increases
with increasing density and decreasing NBI power. The

decay time of the central electron temperature after
turning-off of ECRH power is significantly slower in thecase with ITB than without ITB. This suggests thetransport improvement inside ITB, which is an

was no sign of impurity accumulation. As to theoperational density limit in long pulse discharges, highdensity plasma with 6.7 x l0rem-3 is sustained for 10 s

encouraging sign of confinement enhancement.

with a NBI power of 2 MW. The global energy

5. Long Pulse Discharges
Demonstration of steady-state plasma with highconfinement characteristics of long-duration plasmas arealmost the same as those in short-pulse discharges [21].

performance is one of the most challenging issues. Suchinvestigation is particularly appropriate for the research
program in LHD which employs superconducting coils
119,201. The NBI (0.5 MW) heated plasma wassustained up to 80 s (see Fig. 7(a))and the ICRF (0.4

6. Discussions and Future Prospect
The experiments in LHD have greatly extended theenvelope of study on net current-free plasmas in thesethree and a half years. Significant progress can be seenin achievement of the representative plasma parametersMW) heated plasma with the discharge duration of 2
minutes (see Fig. 7(b)) is achieved. In both discharges
with helium gas puffing, the electron densities are kept
to be around 1 x 10rem-3 and the temperatures were one

8

ITB

(a)

(358kw)

6

9+
F

o

2
0

(Wp= 7.12 MJ, Z"(0) = 10 keV, 4(0) = 5keV, B = 3.2
Vo). The energy confinement time of LHD plasmas iscomparable to that of the ELMy H-mode plasmas intokamaks with q=4.5 from the aspect of the scaling law
1221. The most important finding is that good MHD
stability and good transport are compatible in the inward
shifted configuration (Ru* = 3.6m). This finding gives aperspective of the heliotron line a breakthrough. The

NO ITB

inward shifted configuration has magnetic hill geometry
and thus was predicted to be MHD unstable by ideal

t

PJr
rl

.

0

p

-o

# 17311E

0.5

Teo (ECE)

.o
rc

1

^0.
>0.

P

E0.

F

-{

n l=0.3x101em-3

E

-r

tc

5
.3^

xo

o

B

q

0.1 P

PECH'(MW)

40

60

80

.#n
100

t-

120

Time (s)Fig. 6 Observation ITB-like structure. {a) Electron temFig. 7 Typical wave forms of long-pulse discharges. (a)NBI heated plasma. Slow undulation is due to in-perature profile with high and low heating power
across the transition. (b) Dependence of thecentral electron temperature on heating power.
Arrows correspond to the cases shown in (a).

tentional density control not due to spontaneous
behavior of plasma. (b) ICRH plasma.
26

Motojima O. et al., Recent Development of LHD Experiment

final goal is clearly defined and this figure indicates that102

steady state operation is the most challenging issue. The

lst target is possible by maximizing the present
experimental capability. The 2nd rarget is held up withreasonable upgrade of the machine. LHD is surpassing
the existing tokamak experiments in terms of operated
duration with significant plasma parameters.

101

66'
=
EO
or.li
9oE

100
10-1

.3E

guJ
,9 Po
O t--

trc

ooo
,....._..LHD

10-2

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up to p = 3 Vo. The geometrical optimization ofneoclassical transport by the inward shift of themagnetic axis is quite successful, however, themechanism of suppression of anomalous transport due
to the inward shift is still an open question. With central
deposition of EICRH power, the ITB shape profiles areseen, providing a hope for further confinement
improvement. The other important achievements andobservations are (i) a long pulse discharge with duration
time of 2 minutes (80 seconds) has been achieved with
0.4 MW ICRH (0.5 MW NBI) power, respectively. (ii)The external irnposed island (nlm=l/l) is suppressedwith collionless finite beta plasmas. (iii) ICRF heating(minority ion heating) has been very successful and theheating efficiency is comparable to that of NBI heating[23-251. (iv) The transition from ion root to electron roothas been observed [26]. (v) The divertor is very
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A-245 (200 .

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by measuring ttre divertor plasma by electrostatic probes
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energy 1291.

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Figure 8 shows the operational regime on the planeof the fusion triple product and operated duration. The(2001).

t29l R. Sakamoto et al., Nucl. Fusion. 41, 381 (2001).

27