J. Plasma Fusion Res. SERIES, Vol. 9 (2010)

Achievements and Continued New Phase of
the Large Helical Device
Hiroshi YAMADA for LHD Experiment Group
National Institute for Fusion Science, 322-6 Oroshi, Toki, Gifu 509-5292, Japan(Received: 17 November 2009 / Accepted: 19 January 2010)

This paper reviews major achievements in the large helical device (LHD) project and discusses the prospects
in the upcoming nearest future. The LHD is the largest magnetic confinement device among diversified
helical systems and employs superconducting coils. These significant features have enabled comprehensive
studies on steady-state net-current free plasmas with plasma parameters comparable to large tokamaks. The
major achievements are highlighted by high beta (5.1%), high density (1.2×1021m-3), high ion temperature(Ti of 5.6 keV at 1.6×1019m-3), and steady-state operation (3200 s with 490 kW). This progress in physicalparameters has elucidated the potential of net-current free helical plasmas for an attractive fusion reactor.
The recent finding of an internal diffusion barrier has extended high density operation beyond 1×1021m-3 at
the moderate magnetic field of 2.5 T. This achievement provides a novel scenario to a super high density
reactor. Based on achievements that have resolved fundamental concerns in helical system, an upgrade plan
to advance the LHD project to a new phase has just started. Major elements are a closed divertor system toimprove particle control under high heating power and steady state, upgrade of heating capability and the useof deuterium gas.
Keywords: heliotron, superconducting system, steady state operation, high ȕ, high density,internal diffusion barrier, impurity hole, 3-D effectoperational experience over 12 years proves that LHD has
built up the engineering base of a large-scale
superconducting and cryogenic system for fusion reactor
development [4].

1. Introduction

The large helical device (LHD) is the world largest
toroidal device for studying the physics of a net
current-free plasmas. LHD is based on the concept of aheliotron magnetic field and employs superconducting
coils [1,2]. The progress of the LHD project is reviewed
putting emphasis on the latest achievements.
The role of LHD can be revealed by recalling thedefinition of the broader approach. The basic activities andfunctions in a broader approach are categorized into threeframeworks which are primarily ITER oriented,
ITER/DEMO oriented and primarily DEMO oriented. Thissecond category, ITER and DEMO oriented, is attributed to
exploration of operational regimes and issues
complementary to those addressed in ITER. Here threemajor subjects have been extracted; steady state operation,
the advanced plasma regime that is high ȕ, and control ofpower fluxes to walls. Although these are identified in theroadmap towards a tokamak DEMO reactor, it should bepointed out that these are indeed what are defined as themain mission of LHD at the beginning of the project.
Therefore the LHD project is aimed at two goals, one is toformulate a systematic scenario to an attractive helical
reactor and one is a comprehensive understanding of
toroidal plasmas including tokamaks.

Fig.1 A bird’s eye view of the LHD magnet system.
LHD is based on the heliotron magnetic configuration[5] and has a pair of helical coils and 3 sets of poloidal
coils (see Fig.1). All these coils are superconducting andgenerate the confining magnetic field without the help of
net plasma current. In addition, LHD has 10 pairs of
perturbation coils which enable the study of resonant
magnetic perturbations [6]. NBI is the primary heating

2. Machine Status and Parametric Progress

LHD has worked very well for 12 years since its
initial operation in 1998 [3]. The accumulation of

author’s e-mail: hyamada@LHD.nifs.ac.jp1©2010 by The Japan Society of Plasma
Science and Nuclear Fusion Research

H. Yamada for LHD Experiment Group, Achievements and Continued New Phase of the Large Helical Device

steady state magnetic field. Heating capability has been
growing, in particular NBI. Due to this firm engineeringfoundation, physical parameters have progressed
dramatically. The steady-state heating facility of ICH [7] aswell as ECH makes an essential contribution to a 1-hourlong discharge. The beta and the central density are plottedin the middle in Fig.2(b). Distinguished development of thecentral density is due to the finding of an internal diffusion
barrier (IDB) [8,9] describe later. The progress oftemperatures is drawn in Fig.2(c). The electron temperature
jumped up by the improved confinement due to radial
electric field, which is referred to a core electron root
confinement (CERC) [10]. The ion temperature has begun
to increase recently due to a perpendicular NBI [11] which
is favorable to ion heating because of a low accelerating
voltage of 40 keV.

source and also ECH as well as ICH play important rolesin the experiment. The major operating specifications of
LHD are summarized in Table I.
Table I Machine specifications of LHD
Major radius
Minor radius of helical coil
Minor radius of plasma
Magnetic field
Magnetic energy
Coil temperature
Heating power
ECH

ICH
NBI

3.9 m
0.975 m0.5 – 0.65 m
2.96 T at R = 3.6 m
0.77 GJ
3.5 K
3.5 MW
3.0 MW
23 MW

3. Steady State Operation

LHD has already achieved a one-hour longsteady-state discharge with the temperatures in the keVrange as shown in Fig.3. The total amount of input energy
reached 1.6 GJ [12]. This steady state operation of LHD ishighlighted by 500 kW which consists of 400 kW of ICH
and 100 kW of ECH. As shown in Fig. 3(d), thetemperature of divertor plates is saturated by a magnetic
axis swing technique [13]. A sweep of only 3cm in major
radius of the magnetic axis position (less than 1% of the
major radius of the LHD) was sufficient to disperse thedivertor heat load.

(a)

0
Ti
Te

-3

(c)

5
e0

(1019m-3)

n

_
ne

IDB

20

T (keV)

10<E>(10 m )

5

(b)

e0

0
5
4
3
2
1
0
10

Ti, Te
(keV)

10

P
(kW)

NBI
ICH
ECH

CERC
40-keV
NBI

Tdiv
( C)

20

n

<E> (%)

Power (MW)Several-month-long operation has been executed 12
times since 1998. The 13th experimental campaign was
conducted in 2009. The operational time of the helium
compressor has amounted to 57,000 hours with the average
duty of 99.4%. The superconducting coils have been
excited more than 1,300 times and LHD has produced
more than 94,000 plasma discharges to date. This
unusually high availability is due to a reliablesuperconducting system and the intrinsic advantages of net
current free plasmas.

01998 2000 2002 2004 2006 2008

0.5
0
300
200
100
0

0

year

Fig.2

800
600
400
200
0
0.6
0.4
0.2
0
1.0

Progress of heating capability andrepresentative plasma parameters. (a)heating power, (b) volume averaged E andthe central electron density, and (c) central
ion and electron temperature.

Fig.3

Figure 2 is an overview of the progress in these 11
years. Diversified achievements have been obtained by asynergetic effect of technological upgrade and new physics
findings. The superconducting coils provide a reliable

Shot 66053

Ti

Te

Inner
Upper
500 1000 1500 2000 2500 3000 3500

Time (s)Waveforms of a typical steady-state
long-pulse discharge in LHD. (a) heating
power, (b) line averaged density, (c)temperatures, and (d) temperature ofdivertor plate. This figure is a reproduction
from Fig.9 in H.Yamada et al., Fusion Eng.
Design 84 (2009) 186.

The RF oscillator for ICH itself has already achieved
1.6 MW operation for 5000 s [7]. Also the steady state2

H. Yamada for LHD Experiment Group, Achievements and Continued New Phase of the Large Helical Device

which is essentially proportional to the square root of
power density. This scaling is applicable to LHD for
diversified density profiles when we take the edge density
at the electron temperature of 100 eV [19]. Figure 5 shows
the dependence of density on the heating power. The casewith the pellet injection is a more peaked density profile
than the case with normal gas-puffing. The offset between
the cases with different fueling schemes as well as thescatter of data converges when the local edge density at theelectron temperature of 100 eV is evaluated (see Fig.5 (b)).

heating capability of ECH is now upgraded to 300 kW.
Together with a new closed divertor system with more
efficient heat removal than the present one, further upgradeof the steady state heating capability will enable 3 MW
operation for one hour which corresponds to an input
energy of 10 GJ. This is close to the ITER regime in termsof handled auxiliary heating.

4. High Beta

The development of beta in the LHD experiment isvery encouraging. The volume averaged ȕ of 5 % has beenachieved without any disastrous instability while MHD
activity is certainly seen as theory predicts [14]. The highbeta ranging up to 5 % is maintained for longer than 100
times the energy confinement time. The LHD experimenthas discovered that the interchange instability in themagnetic hill is benign [15,16]. The interchange mode
analysis based on the linear MHD theory suggests that
LHD plasmas reaching 5% are unstable against theinterchange mode due to a magnetic hill. In spite of thisunfavorable stability condition, the plasma can passthrough the unstable region and evolves to the high ȕ state.
Figure 4 indicates that the achievable beta is limited by theavailable heating power without a hard beta limit.
Geometrical optimization and the upgrade of the heating
power will enable further development.

E (%)

-3
20

e

<n > (102
2.2 n Sudo

Sudo

nc

20

(10

100eV

1

e

n

3

0

5

P (MW)

10

Fig.5 Dependence of operational envelope ofdensity for the cases with different fuelingschemes on heating power. (a) volume
averaged density and (b) edge density at theelectron temperature of 100 eV.

2
B=0.41-0.425T

Pabs (MW)PelletGas puff

-3

m )

(b)

0

5

c

0

E = 2.4Pabs1/3

0

c

1

n Sudo41PelletGas puff

m )

3

6
5

(a)

10

This means that the core density is free fromoperational limit. This advantage over tokamaks is due tothe freedom from current drive and disruption. Then the
chance to investigate the underlying physics of the density
limit, which is not determined by current disruption,
becomes available. Also, since the energy confinement
time of helical systems has significant positive dependence
0.6
on density like WE v ne [20], extension of this favorable
density dependence is a very important issue for theprospects of a fusion reactor.
As a highlight of the high density topic, adistinguished high density plasma in LHD should be noted.
After a sequence of pellet injections to build up the central

Fig.4 Achieved ȕ as a function of the absorbed
heating power.

5. High Density

The assessment of the density limit is extremely
important in the operational design of a reactor. In
tokamaks, the operational density limit is well described by
the Greenwald limit [17]. In contrast, the plasma in LHD
can be operated well beyond this limit when thedependence on the plasma current density is rephrased by
the rotational transform. Another scaling of the density
limit for helical systems has been proposed by Sudo [18],3H. Yamada for LHD Experiment Group, Achievements and Continued New Phase of the Large Helical Device

temperature. While the peripheral temperature remains low
in spite of an increase of heating power, the central
temperature dramatically grows. The power degradation ofheat transport, in other words temperature dependence, issignificantly mitigated in this mode [22].
In addition, it is very interesting that this high ion
temperature mode is accompanied by a so called impurity
hole [23]. Figure 6 (b) shows the typical profiles ofelectron density by open circles and carbon impurity
density by closed squares. It should be noted that the scales
differ by 100 times. The profile of the carbon impurity isextremely hollow compared with the electron density
profile. This trend is emphasized by the increase of the ion
temperature gradient. Therefore even with carbon impurity
pellet injection, carbon is expelled from the core withsignificant outward convection. It should be noted that thisphenomena contradicts the prediction of neoclassicaltransport with an observed negative radial electric field.

density [21], the super-dense-core is formed under thecondition of reduced edge recycling and avoidance of thelocal concentration of neutrals. The plasma exhibits a highpeaked density profile which is referred to an internal
diffusion barrier (IDB) [8,9]. Although the plasma with an
IDB has a much higher density in the core than thegas-fuelled plasma, the electron temperature is even higher.
Then the central pressure is enhanced by a factor of 4. By
this operation the central density has reached 1.2 ×1021m-3
at the moderate magnetic field of 2.5 T. This recent
discovery suggests a potential for the novel scenario of asuper high density reactor with ignition at relatively low
temperatures less than 10 keV.

6. High Ion Temperature ModeFig.6

7. 3-D EffectIt should be pointed out that the 3-D equilibrium has
the distinguished features of a magnetic island andstochastic field. The numerical code HINT [24] that copeswith this 3-D equilibrium appropriately has beendeveloped in NIFS. Figure 7 (a) is the pressure profile in a
typical high-ȕ discharge. The open circles are theexperimental values and the dotted curve is the result fromHINT [25]. HINT reconstructs the plasma equilibriumquite well, nonetheless, controversial observation comesout. It is that a significant pressure gradient exists in theedge stochastic area. It should be pointed out that a 3-D
equilibrium is not specific to helical systems. Resonant
magnetic field coils are considered in tokamaks so as tocontrol ELMs and also stabilize the RWM.

(a)

5
4
3
2
Ion
Electron

1
0
2

1

0

6

(b) 0.02

4

E (%)

Carbon density (1019m-3)

2
0

0.01
CarbonElectron
0

0.2

0.4
reff (m)

0.6

0

Z (m)

Electron density (1019m-3)

Temperature (keV)

Recently a perpendicular NBI with a low accelerating
voltage of 40 keV has become available and ion heating
experiments have progressed. The previous tangential NBI
has a higher accelerating voltage of 180 keV, which
predominantly heats electrons. To date the central ion
temperature of 5.6 keV has been achieved at the density of
1.6×1019m-3 as shown in Fig.6. The ion temperature profile
is peaked and higher than the electron temperature in the
core.
6

0

-0.5
2.5

Typical profiles in the high-ion-temperatureplasma. (a) ion and electron temperaturesand (b) electron and carbon densities.

3.0

3.5

4.0

R (m)

4.5

Fig.7 (a) Pressure (E) profile. Open circles areexperimental values and dotted curve is theresult from HINT. (b) Reconstructedequilibrium by HINT.

A distinguished feature is observed in the ion
temperature in the core relative to the peripheral

There are possibilities that a plasma heals the4

H. Yamada for LHD Experiment Group, Achievements and Continued New Phase of the Large Helical Device

LHD has been providing the “unique” and“complementary” basis in fusion power development andshares the role with tokamaks. Recent highlights ofphysics achievements are reviewed in this article. The
understanding of net current free plasmas has progressed
dramatically through LHD experiments. Highlighted
achievements and accurate knowledge about steady state
operation, high E, density limit and 3-D effect arecomplementary to tokamaks in fusion power
development. The near-term upgrade package including aclosed helical divertor, the upgrade of heating capability,and deuterium will reinforce the role of LHD. In the
coming next decade, LHD is aimed at the challenging
integration of elements towards an attractive fusion
reactor.

stochastic magnetic field spontaneously or that a finite
pressure gradient is determined consistently with thecharacteristics of a stochastic magnetic field.
With regard to heat transport due the stochastic
magnetic field, the Rechester-Rosenbluth model [26] iswell known, but it predicts unusually large radialtransport at high electron temperatures in LHD. Here thediffusion coefficient of the magnetic field line DFL is an
essential parameter. The Rechester-Rosenbluth model
assumes a simple Markov chain process, however thestochastic area in LHD has a memory of the structure of
the magnetic island. In this situation DFL is defined as theisland width squared divided by the parallel lengtharound a circumference in the magnetic island. As
temperature increases, the parallel length increases. And
consequently DFL is reduced. Then a finite pressure
gradient can be expected in the stochastic area [27]. The
3-D edge transport simulation in the stochastic area by
the EMC3-EIRENE code [28,29] supports this scenario.
As the electron temperature increases, the electron
temperature in the island is flattened more clearly and theboundary becomes clearer and extends. This change
corresponds to the fact that the parallel path becomes
predominant.

Acknowledgements

The author is grateful to all contributors to the LHD
project, in particular colleagues in the LHD experimentgroup. The support by the LHD operation group is greatly
appreciated. This work is supported by budget
NIFS09ULGG001 of the National Institute for Fusion
Science.
References

8. Nearest Future Plan and Summary

In 2010, LHD will make a significant step forward.
The new 5th NBI with perpendicular injection and apower of 7MW will be available. The total heating power
by NBI will reach 30 MW. Also the vacuum vessel will
be modified in order to incorporate a baffle structure toform a closed divertor. As the first step, the closed
divertor will be installed at the inboard side of 2 of 10
toroidal sections. This is a proto-type without cryo-pumps.
The capability of neutral compression will be evaluatedand it will be applied to the final design. Together withthis closed divertor, an upgrade of the steady state heating
capability of ICH as well as ECH will enable us toattempt steady state operation with 3MW for 1 hour in acouple of years. The deuterium experiment is also a very
important next step to assess the isotope effect. Reactor
design studies are also progressing based on the resultsfrom LHD.
It is worth mentioning the Asian network, in otherwords, the solidarity of superconducting steady state
toroidalexperiments.
5
world-classmajor
superconducting experiments; LHD, EAST [30],
KSTAR[31], SST-1[32] and JT-60SA [33], are working
or under construction in Asia. It should be realized that asignificant gap exists between these present experiments
and the next generation; ITER. Therefore it is extremely
important to share common critical issues among theseexperiments and collaboration can accelerate the progress.
It should be emphasized that a complementary approachprovides a more mature solution than one line alone.

[1]

A. Iiyoshi et al., Nucl. Fusion 39, 1245 (1999) .

[2]

O. Motojima et al., Nucl. Fusion 40,599 (2000).

[3]

H.Yamada et al., Fusion Eng. Design 84, 186 (2009).

[4]

T. Mito et al., Fusion Eng. Des. 81, 2389 (2006).

[5]

K.Uo, Plasma Phys. 13, 243 (1971)[6]

A. Komori et al., Nucl. Fusion 45, 837 (2005)[7]

T. Seki et al., Fusion Sci. Technol. 40, 253 (2001).

[8]

N. Ohyabu et al. Phys. Rev. Lett. 97, 05502 (2006).

[9]

H.Yamada et al., Plasma Phys. Control. Fusion 49,
B487(2007)[10]

M.Yokoyama et al., Nucl. Fusion 47, 1213 (2007)[11]

Y. Takeiri et al., Nucl. Fusion 47, 1078 (2007)[12]

T. Mutoh et al., Nucl. Fusion 47, 1250 (2007)[13]

H.Ogawa et al., Plasma Fusion Res. 2, 043 (2007)[14]

S.Sakakibara et al., Plasma Phys. Control. Fusion 50,
124014 (2008).

[15]

H.Yamada et al., Plasma Phys. Control. Fusion 41, A55(2001).

[16]

5

K.Y. Watanabe et al., Nucl Fusion 45, 1247 (2005).

[17]

M.Greemwald et al., Nucl. Fusion 28, 2199 (1988)[18]

S.Sudo et al., Nucl. Fusion 30, 11 (1990)[19]

J.Miyazawa et al., Nucl. Fusion 48, 015003 (2008).

[20]

H.Yamada et al., Nucl. Fusion 45, 1684 (2005)[21]

R.Sakamoto et al., Plasma Fusion Res. 2, 047 (2007).

[22]

K.Ida et al., Nucl. Fusion 49, 095024 (2009).

[23]

M.Yoshinuma et al., Nucl. Fusion 49, 062002 (2009)[24]

K.Harafuji et al., J.Comput. Phys. 81, 123 (1989)[25]

K.Y.Watanabe et al., Plasma Phys. Control. Fusion 49,

H. Yamada for LHD Experiment Group, Achievements and Continued New Phase of the Large Helical Device

605 (2007).
[26]

A.B.Rechester and M.N.Rosenbluth, Phys. Rev. Lett. 40,[27]

M.Kobayashi et al., J. Nucl. Mater. 363-365, 294

38 (1971).
(2007).
[28]

Y.Feng et al., Contrib. Plasma Phys. 44, 57 (2004).

[29]

D.Reiter et al., Fusion Sci. Technol. 47, 172 (2005).

[30]

B.Wan et al., Nucl. Fusion 49, 104011 (2009).

[31]

Y.K.Oh et l., Fusion Eng. Design 83, 804 (2008).

[32]

B.Sarkar et al., Fusion Eng. Design 81, 2633 (2006).

[33]

M.Matsukawa et al., Fusion Eng. Design 83, 795(2008).

6