Proceedings of ITC/ISHW2007

Recent Development in the Operational Regime of Large Helical
Device Toward a Steady-State Helical Fusion Reactor
Y. Takeiri, A. Komori, H. Yamada, K. Kawahata, T. Mutoh, N. Ohyabu, O. Kaneko,
S. Imagawa, Y. Nagayama, K.Y. Watanabe, T. Shimozuma, K. Ida, R. Sakamoto, M. Kobayashi,
A. Sagara, K. Nagaoka, M. Yoshinuma, S. Sakakibara, Y. Suzuki, M. Yokoyama, S. Sudo,
O. Motojima, and LHD experimental Group
National Institute for Fusion Science, 322-6 Oroshi-cho, Toki 509-5292, Japan(Received 29 October 2007 / Accepted 29 October 2007)
Recent progress of the plasma performance and the related physics understanding is overviewed in Large
Helical Device. The volume-averaged beta value is increased with an increase in the NBI heating power, andreached 5.0% of the reactor-relevant value. In the high-β plasmas, the plasma aspect ratio should be controlled sothat the Shafranov shift would be reduced mainly to suppress the transport degradation and the deterioration of NBI
heating efficiency. The operational regime of the high-density plasma with internal diffusion barrier (IDB) has been
extended, and the IDB, which was originally found by use of the local island divertor, has been realized in thehelical divertor configuration. The central density was recorded as high as 1x1021m-3, and the central pressure
reached 130kPa. Based on these high-density plasmas with the IDB, a new ignition scenario has been proposed.
This should be a specific scenario to the helical fusion reactor, in which the helical ripple transport would be
mitigated. A low-energy positive-NBI system was newly installed for an increase in the direct ion heating power. As
a result, the Ti exceeded 5.2keV at a density of 1.2x1019m-3 in the hydrogen plasma. The Ti-increase tends to beaccompanied by a large toroidal rotation velocity of the order of 50km/s in the core region. The plasma properties
in the extended operational regime are discussed with regard to a perspective to the steady-state helical fusion
reactor.
Keywords: Large Helical Device, MHD stability, Shafranov shift, internal diffusion barrier, ion transport, toroidal
rotation, impurity hall, helical fusion reactor, neoclassical transport, anomalous transport, core electron-root
confinement

1. Introduction
The Large Helical Device (LHD) is the world’s
largest superconducting helical device, which started its
operation in 1998 [1-3]. The objective of LHD is todemonstrate high performance of net-current free heliotron
plasmas relevant to the steady-state helical fusion reactor.
During the nine years’ operation, LHD has exploited novel
operational regimes related to plasma confinement, MHD
stability and steady-state operation, including plasma-wallinteraction [4,5].
Recently, much progress of the plasma performancehas been achieved together with the physics understanding,which was mainly brought by the upgrade of the heating
systems. The volume-averaged β value is increased with anincrease in the NB injection power, and reached 5% as aconsequence of the enhancement of the negative-NBI
power to 14MW. A low-energy positive-NBI system has
been installed recently for an increase in the direct ion
author’s e-mail: takeiri@nifs.ac.jp

heating power, and, as a result, the Ti exceeded 5.2keV at adensity of 1.2x1019m-3 in a hydrogen plasma. The
distinguished feature found in LHD is the formation of theinternal diffusion barrier (IDB) [6], which is realized by
combination of efficient pumping with the local island
divertor (LID) [7] and core fuelling with repetitive pellet
injection [8], and it is sustained by the high-power NBI
heating. In a super-dense core (SDC) plasma with the IDB,
the central density was recorded as high as 5x1020m-3 [6].
The IDB has also been observed recently in the helicaldivertor (HD) configuration with a well-conditioned wall.
Based on the SDC/IDB plasmas, a new ignition scenario
has been proposed, and this should be a specific scenario tothe helical fusion reactor, in which the helical ripple
transport would be mitigated [9]. Inherent advantage to
steady-state operation has been demonstrated with the
improvement of the ICRF and ECH heating systems. A
long-pulse plasma was sustained for 54min with ICRF andProceedings of ITC/ISHW2007
ECH, and the input heating energy reached 1.6GJ [10].
One of the specific characteristics in LHD is thedependence of the plasma properties, such as confinement,
transport, MHD stability, and divertor structure, on themagnetic axis position. In the inward-shifted configuration,
the transport including the high-energy particle
confinement is improved while the MHD stability isdegraded due to the enhancement of the magnetic hillregion, and the tendency is reversed in the outer-shifted
configuration. The most significant achievement in the
LHD research is the compatibility of the transport and the
MHD stability, which is realized by the sophisticatedoptimization of the magnetic configuration including themagnetic axis position and the control of the Shafranov
shift due to the plasma pressure [11].
In this overview, we present the recent developmentin the operational regime of the LHD experiments,
highlighting mainly topics for the high-β, high-densitywith IDB, and the high-Ti, and the related physics
understanding is discussed with a perspective of thesteady-state helical fusion reactor.

2. Large Helical Device and Heating Systems
The Large Helical Device (LHD) employs a heliotron
configuration, and consists of a pair of continuous helicalcoils, which has an l=2/m=10 poloidal/toroidal field period,
and three pairs of poloidal coils [3]. These coils are allsuperconducting, and provide a wide variety of themagnetic configuration by controlling the current ratio ofthese coils. The magnetic axis position is changed with thepoloidal coils, and ranges 3.5-4.1m in the major radius in
the experiments. The averaged minor radius is about 0.6m
at the maximum and the corresponding plasma volume isabout 30m3, and these are dependent on the magnetic
configuration. The magnetic field strength on the axis is 3T
at the maximum.
The LHD has the intrinsic helical divertor (HD),which is a sort of helically twisted double-null open
divertor. Additionally, the LHD is equipped with a local
island divertor (LID) system, which utilizes the m/n=1/1
magnetic island induced externally with the perturbationcoils [7]. The LID facilitates the edge plasma control withhighly efficient pumping capability.
The main heating system is a negative-ion-based
neutral beam injection (negative-NBI) system, which
consists of three tangential injectors with the nominalhydrogen-injection energy of 180keV [12,13]. The total
injection power achieved is 14MW. A low-energy
positive-NBI system, the injection direction of which isperpendicular to the magnetic axis, was operational in2005, and 40keV-6MW injection has been achieved [14].
The electron cyclotron resonance heating (ECRH)
system is equipped with 168GHz, 84GHz, and 82.7GHz

gyrotrons, and the total injection power achieved is 2.1MW[15]. Using antenna systems with quasi-optical mirrors,each microwave is injected as a highly focused Gaussianbeam, and the focused location can be changed at 3.5-3.9m
of the major radius on the equatorial plane.
The ion cyclotron range of frequencies (ICRF)
heating system facilitates the long-pulse operation, whichis capable of a steady-state injection of 1MW with fourantennas, while the achieved power is 2.9MW in a shortpulse injection [16].

3. Properties of Reactor-Relevant High-β
PlasmasAs described above, the LHD-plasma performance isstrongly dependent on the magnetic axis position. Since theheliotron configuration has a weak magnetic shear in thecore region and a magnetic hill toward the peripheral
region, the linear theory indicates that the interchange
mode is unstable in the inward-shifted configurations, inwhich the particle orbit and the transport are better. On theother hand, in the outward-shifted configurations, although
the MHD mode is stable due to expansion of the magnetic
well region, the particle orbit and the transport are
degraded. Through the optimization of the magnetic axis
position, it was found that the interchange mode in the coreregion is stabilized and the Mercier criterion does notprohibit the access to higher β, even in the inward-shifted
configuration of Raxvac=3.6m by the spontaneous
generation of a magnetic well due to the finite β effect[11].
However, as the β increases further, the Shafranov
shift becomes larger. As a result, the transport is degradedFig.1 Waveforms in a high-β plasma of <β>=5.0%
achieved with the repetitive pellet injection.

Proceedings of ITC/ISHW2007
and the NBI heat deposition is deteriorated due to theenhancement of the orbit loss by the outward shift of themagnetic axis. Moreover, the large Shafranov shift lowers
the MHD equilibrium beta limit. Then, the plasma aspectratio, Ap, which can be controlled by changing the current
ratio of the three layers of the helical coils, should be
optimized [17]. A large Ap reduces the Shafranov shift, but
an excessive increase in the Ap violates the MHD stability
due to immoderate suppression of the spontaneous
generation of the magnetic well. As a consequence of aseries of experiments for the Ap optimization including themagnetic field strength, we have achieved 5% of thevolume-averaged beta value, <β>, which is areactor-relevant value, at Ap=6.6, B=0.425T, and preset
Raxvac=3.6m with an increased NBI absorbed power of
11MW.
Figure 1 shows discharge waveforms in the high-β
plasma of <β>=5.0% achieved with a pellet injection. The
tangential NBI power is reduced in the repetitive pellet
injection phase for effective penetration of the pellets. At
that time, the Shafranov shift is rapidly reduced, and thenthe perpendicular NBI is injected together with therestoration of the tangential NBI immediately after the endof the pellet injection. Due to the reduction of the
Shafranov shift, the perpendicularly injected beameffectively heats the plasma, and the <β> is rapidly
increased, then reaching 5%. Although <β>=5% wastransiently achieved due to the effective fueling by thepellet injection and the effective heating by theperpendicular NBI during the reduction of the Shafranov
shift, no crucial MHD instability was observed in a periodof 10 times the confinement time, during which the <β>
was over 4.5%.

Fig.2

Waveforms in the stationary sustained high<β> plasma with only the gas puffing.

Fig.3 Volume-averaged beta value, <β>, as afunction of the NBI-absorbed power.
With only the gas puffing, 4.8% of <β> wasstationary sustained without any disruptive phenomena as
shown in Fig. 2. The Shafranov shift normalized by theeffective plasma minor radius is as large as around 40%,
approaching the equilibrium beta limit of 50%. Although
the MHD modes excited in the peripheral region with themagnetic hill are dominantly observed, they become stablespontaneously from the inner region to the outer region as
the <β> increases. According to the linear MHD analysis,
the Mercier criterion is not violated, and the observed
peripheral MHD modes in the magnetic hill region is not
crucial due to the strong magnetic shear there [18].
Even in the highest-β plasma, as described above, theMHD instabilities do not limit the achievable beta, but theheating power does. As shown in Fig. 3, the <β> isincreased linearly to the 0.25 power of the NBI absorption
power without saturation. The deduced absorbed power
dependence of the <β> from the energy confinement
scaling ISS95 is Pabs0.4, and, thus, confinement degradationfrom the scaling law is suggested in the high <β> plasmas.
The peripheral transport is enhanced with increasing the<β>. The dependence of the magnetic fluctuation resonatedwith the peripheral rational surfaces on the magnetic
Reynolds number, S, suggests that the observed MHD
modes are resistive interchange modes [19]. Thus, theresistive g-mode turbulence would cause the transportenhancement. Since the S is proportional to BtTe3/2ne-1/2, theMHD mode amplitude is expected to be reduced forplasmas with higher temperature at higher magnetic field,
relevant to the reactor-class plasma conditions. Anotherpossible cause for the confinement degradation is theincrease in the effective helical ripple due to the Shafranov
shift [20]. The real-time control of the magnetic axis
position during the discharge is planned, and that should
resolve this issue. That should also be effective toProceedings of ITC/ISHW2007
enhancement of the heating power by the perpendicular
NBI, and then the further increase in the <β> is expected.

4. Extension of Operational Regime inHigh-Density Plasmas with Internal Diffusion
BarrierFormation of the super-dense core (SDC) plasma with
the internal diffusion barrier (IDB) is a distinguishedfeature in LHD, which has been never observed intokamaks. The IDB is originally formed in the
LID-controlled plasmas which are fueled directly in thecore region by the repetitive pellet injection. The LID
functions as a strong pumping, which realizes a low
recycling condition through the particle control at theplasma edge. Even in the helical divertor (HD)configuration, intensive wall conditioning should lead to apumping by the wall. In the outward-shifted configuration,
localization of the neutrals is small, and the recycling is
suppressed due to weaker plasma-wall interaction. We
have realized the IDB formation in the HD configuration in
the outward-shifted configuration with such a kind ofwell-conditioned wall.
Figure 4 shows typical waveforms of an SDC plasma
with the IDB realized in the HD configuration. The preset
vacuum-magnetic axis position, Raxvac, is 3.8m. During therepetitive pellet injection into the plasma heated withhigh-power NBI, the electron density is greatly increased
and the central density exceeds 5x1020m-3. After the core

Fig.4

Typical waveforms of the SDC/IDB plasma
formed in the HD configuration.

Fig.5 Maximum central density, ne(0), and pressure,
P(0), as a function of the preset vacuummagnetic axis position, Raxvac.
fueling with the pellet injection, the peripheral density is
rapidly decreased in the density relaxation and temperature
recovery phase, and, then, the SDC plasma with the IDB isformed. The central pressure continues to be still increased
after the plasma stored energy turns to be decreased, and,occasionally, the core density collapse (CDC) occurs with
an abrupt decrease in the central pressure as well as thecentral density. The impurity accumulation is not observed
and the Zeff is as low as around 1.5 during the discharge.
A series of the SDC/IDB experiments were performedin the HD configuration with a variation of the Raxvac, andthe results are shown in Fig. 5. The formation of the IDB isdefinitely observed in the outward-shifted configuration at
Raxvac>3.7m. The achieved central density, ne(0), is
increased as the Raxvac is shifted more outward, and reaches
1x1021m-3 at Raxvac=3.9m, while the peripheral density is
maintained low, independent of the Raxvac. The centralpressure, P(0), jumps up at the IDB formation, and the
maximum P(0) is 130kPa, exceeding an atmosphericpressure, at Raxvac=3.85m.
The IDB formation is closely correlated with theedge-region temperature, which is strongly affected by theneutral density there. An increase in the neutral pressure in
the SOL region causes an increase in the edge density withenhanced recycling, leading to a decrease in the edge
temperature. Then, the heating power is lost by enhanced

Proceedings of ITC/ISHW2007

Fig.6 Comparison of the density, temperature, andpressure profiles of the SDC/IDB plasma withthose of the gas-puffing plasma with no IDB.
Raxvac=3.75m.
radiation in the low temperature region, resulting in theradiative collapse [21]. In the outward-shifted
configuration, the recycling is suppressed compared withthat in the inward-shifted configuration, and, thus, the low
edge-density can be maintained by the wall-pumping effectwithout the LID. With a closed helical divertor systemequipped with a strong pumping function, the stable IDB
discharge would be realized even in the inward-shifted
configurations.
Comparison of the pressure profile of the SDC/IDB
plasma with that of the normal gas-puffing plasma at
Raxvac=3.75m is shown in Fig. 6. The magnetic fieldstrength on the axis and the NBI power are the same forboth plasmas, and 2.539T and 11MW, respectively.
Although the central density is about 2 times higher, theelectron temperature is also higher in the SDC/IDB plasma.
As a result, the central pressure is much higher with a steeppressure profile, indicating the confinement improvementin the SDC/IDB plasma. In the peripheral region, lowdensity is maintained, leading to high Te-gradient toward
the core in the SDC/IDB plasma. On the other hand, since
the peripheral density is high and the Te-gradient region isnarrow, the core Te is low in the normal plasma. The
central beta value, β(0), reaches 4.5% even at the highconfinement magnetic field, and a large Shafranov shift ofabout 0.3m is observed due to the high central pressure

with a peaked profile. The achievable β(0) is higher atlower confinement magnetic field, and 5.8% is observed at
Bt=1.5T. The Shafranov shift is approaching to theequilibrium beta limit of 50% of the effective minor radius.
The CDC event shown in Fig. 4 occurs when the
Shafranov shift is so large that the shifted axis position
would exceed R=4.0m. Vertically elongated modificationby the ellipticity change of the magnetic surface is applied
to suppress the Shafranov shift, and the CDC event is
mitigated. The real-time control of the magnetic axis
position by the dynamic control of the vertical field shouldbe effective to improve the SDC/IDB plasma properties.
The achievable plasma stored energy of the SDC/IDB
plasmas in outward-shifted configurations is linearly
dependent on the plasma volume at the preset vacuum
magnetic axis position, Raxvac, including the non-IDB
plasmas in inward-shifted configurations. The plasma
volume is smaller as the magnetic axis shifts more outward.
Considering that the actual plasma volume is smaller due
to the large Shafranov shift in the SDC/IDB plasmas, theplasma confinement is thought to be improved in the
SDC/IDB plasmas.
Compared with the high-β plasma at the low magnetic
field, the SDC/IDB plasma has a similar central beta, asteeper pressure gradient, and a larger Shafranov shift. The

Fig.7 Profiles of the ion temperature and the toroidal
rotation velocity in the high-Ti plasma.

Proceedings of ITC/ISHW2007

t=1.35st=1.75s4
3

ne=1.6x1019m-3

3.8

4.0 4.2 4.4 4.6
Major Radius (m)0Calculation

0.2

0.4

ρ

0.6

0.8

1.0

4
(d)

Calculation

3

2

10
10

2
1

t=1.35st=1.75st=1.35s (Er=0)

χ

i

-4

(m /s)

10

-3

-6
0.0

1

10

(c)

-2

-5

(b)

t=1.35st=1.75s

0.1
0.0

4.8

75235-1.35s.ambEr 15:01:36 2007/05/10-1

lhd_075235_001366_v001_funaba_001_chi_exp 9:30:05 2007/10/09

χ

2

0
3.6

Radial Electric Field (kV/m)

2
3/2
exp 3/2/T(m /s/keV )
i
i

(a)

5

1

10

lhdcxs3_tir@75235_001 10:14:59 2007/04/10

NC

Ion Temperature (keV)

6t=1.35st=1.75s

0.2

0.4

1

ρ

0.6

0.8

1.0

0.1
0.0

0.2

0.4

ρ

0.6

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1.0

Fig.8 (a) Ti profiles, (b) ion thermal diffusivities divided by the gyro-Bohm factor, χi/Ti3/2, (c) radial electric field, Er,
obtained with the neoclassical ambipolar calculation, and (d) neoclassical ion thermal diffusivities, χi.
finite-β equilibrium in the SDC/IDB plasma at
Raxvac=3.85m is investigated, using the HINT2 equilibrium
code in which the nested magnetic surface is not assumed.
The preliminary results show that due to the large
Shafranov shift the magnetic surfaces are distorted in the
peripheral region surrounding the IDB, and that inside the
IDB the magnetic surfaces are clearly closed. Since theconnection lengths of the magnetic field lines are much
longer than the electron mean-free paths even in the regionof the distorted magnetic surfaces, it is possible for theelectron temperature profile to have a gradient. These
analyses of the MHD equilibrium for the SDC/IDB
plasmas should apply to the reactor-relevant high-β
plasmas.

5. Increase in Ion Temperature and the Related
Ion Transport
The high-energy NBI, which is the main heating
device in LHD, dominantly heats electrons, and, thus, thehigh-Z discharge was utilized for the high-Ti experiments

in LHD to increase effectively the ion heating power. As aresult, the Ti was increased with an increase in the ion
heating power, and reached 13.5keV at 0.3x1019m-3 with
Ar-gas puffing [22]. To apply this result to the hydrogen
discharge, a low-energy NBI system with a radial injection,
which dominantly heats ions, has been installed, and
40keV-6MW injection was achieved in the lastexperimental campaign [14,23]. This radially injectedbeam is also utilized for the Ti-profile measurement with
the CXRS along a toroidal line of sight, which is better forthe measurement in the central region than that along apoloidal line of sight [24].
With combination of the high-energy NBI and thelow-energy NBI, 5.2keV of the ion temperature is obtained
at 1.2x1019m-3 as shown in Fig. 7. The density profile tends
to be peaked by the low-energy NBI and to be flat orhollow by the high-energy NBI. The peaked density profile
seems to be preferable for the Ti rise, suggesting a role ofthe fueling effect and/or the inward pinch effect with the
low-energy NBI. Figure 7 also shows a profile of theProceedings of ITC/ISHW2007
toroidal rotation velocity, Vt. The toroidal rotation isenhanced in the same direction as the dominant directionof the tangential NB injectors. As shown in Fig. 7(b), large
Vt of 50km/s and the Vt shear are observed in the coreregion accompanied by the Ti rise. That suggests a
correlation between the ion transport improvement and thetoroidal rotation.
Figure 8 shows the results of the transport analysis.
The ion thermal diffusivities normalized by Ti3/2 of the
gyro-Bohm factor are shown in Fig. 8(b) for the plasmas inFig. 8(a). It is found that the χi/Ti3/2 is much reduced in theTi rise to 5keV. As shown in Fig. 8(c), the neoclassicalambipolar calculation shows negative Er in the core regionin association with the Ti rise [25], meaning that the Ti is
increased in the neoclassical ion root. The calculated
neoclassical χi is shown in Fig. 8(d). Without considerationof the Er effect, the ripple transport is greatly enhanced by
the Ti rise, and it is found that by the negative Er the ripple
transport is greatly reduced. As a result, the neoclassical χi
is not so changed by the Ti rise. Considering that theexperimental χi/Ti3/2 is much reduced, the experimental
improvement of the ion confinement is due to the reductionof the anomalous transport in the ion root. The negative Er
is induced by the increased ion temperature, and a role ofthe negative Er in the reduction of the anomalous transportshould be investigated.
The CXRS intensity profile of the CVI emission
shows that the carbon impurity profile becomes strongly
hollowed as the Ti is increased. This “impurity hole” isclearly observed in the plasmas with the carbon pellet
injection. After the carbon pellet injection, the Ti is
increased in the density-decay phase, and the carbon
density ratio is rapidly decreased to 0.2% at the plasma
center while it is 10% at the edge. As a result, the Ti
measurement in the core region becomes impossible. The
outward flux of carbon is observed even at a negative
carbon-density gradient. In the neoclassical theory, theoutward flow of impurities is predicted in the electron rootwith positive Er, and the impurity pump-out effect isobserved in the ECRH plasmas. However, the “impurityhole” is observed in the ion-root plasma with negative Er.
The physics of the “impurity hole” in the high-Ti plasmas
should be investigated as an inherent subject in the helicalsystems.
On the other hand, the Ti rise is observed also in theelectron root. When the ECRH is superposed on the
NBI-heated plasma, the Ti is increased accompanied by theformation of the electron-ITB with positive Er [22]. The
transport analysis indicates that both electron and ion
transport is improved in the core region with the reductionof the anomalous transport. The toroidal rotation is driven
in the co-direction by adding the ECRH, and thespontaneous toroidal rotation is suggested to be related toFig.9 Operational regime for the high-density ignition
based on the IDB profile in the helical reactor.
the transport improvement in the electron-root plasma [26].
Although the electron transport improvement is a common
feature in the CERC (core electron-root confinement)
plasmas [27], the ion transport improvement in the CERC
suggests a possible approach to achieve thereactor-relevant high-temperature plasmas.

6. High-Density Ignition Scenario for a Helical
Fusion Reactor
For the steady-state helical fusion reactor, a novel
ignition approach based on the high-density SDC/IDB
plasma is proposed. Figure 9 shows the operational regime
for the self-ignition based on the high-density IDB profilein the helical reactor. In tokamaks, since high-density
operation is limited by the current drive condition and the
MHD stability, a high-temperature path at relatively low
density is the possible ignition scenario. On the other hand,
LHD is capable to achieve high-density plasmas above
5x1020m-3 without any limitation, and, based on thishigh-density plasma, the required temperature for theignition, then, should be reduced below 10keV. This
scenario is advantageous to the helical devices which
require no current drive. The operation at a relatively highcollisionality mitigates the helical ripple transport, and theengineering demand derived from the high-temperature
operation is reduced. In the peripheral region of the
SDC/IDB plasma, a temperature gradient is established
due to the low density there, and the core temperature highenough to realize the ignition is achieved by thistemperature gradient. Maintaining the low density in the
peripheral region is also desirable for avoiding theradiative collapse as well as suppression of the enhanced
synclotron radiation. At present, an LHD-type fusion
reactor is being designed based on this high-density
ignition scenario [9].

Proceedings of ITC/ISHW2007

7. Concluding Remarks
Recent extension of the operational regime in the
LHD experiments is reviewed together with the progress ofthe physics understanding. By the sophisticatedoptimization of the magnetic field configuration withregard to the magnetic axis position and the plasma aspectratio, the volume-averaged beta was increased to 5.0%
with an increase in the NBI heating power. Even in thisfusion-relevant high-β plasma, no crucial MHD instability
is observed in the core region while the resistive
interchange modes degrade the plasma confinement in the
peripheral region. However, these MHD modes should bestabilized in reactor-relevant plasmas with higher
temperature at higher magnetic field.
The SDC/IDB plasmas, which were originallyobserved in the LID configuration, were realized in thehelical divertor configuration. The central density is
increased as the preset vacuum-magnetic axis position isshifted outward, and reaches 1x1021m-3 at Raxvac=3.9m. The
central pressure exceeds an atmospheric pressure andreaches 130kPa. The core plasma confinement is improved
while the particle transport in the peripheral plasma
surrounding the IDB is degraded. This, in turn, realizes the
low density with an electron temperature gradient in the
peripheral region. The Shafranov shift due to the highcentral beta and the steep pressure gradient is approaching
the equilibrium limit, leading to the core density collapse
occasionally. The dynamic control of the magnetic axis
position by the vertical field is required as well as in thehigh-β plasmas.
For an increase in the ion heating power, a low-energy
NB injector was newly installed, and 40keV-6MW
injection was achieved. The ion temperature is raised to
5.2keV at 1.2x1019m-3 with the combined heating of thelow-energy NBI and the high-energy NBI. The iontransport is improved with the reduction of the anomaloustransport in the neoclassical ion root. An increase in thetoroidal rotation is observed to be related to theion-temperature rise. The impurity hall is recognized, inwhich the carbon impurity is drastically decreased in thecore region with an increase in the Ti. The impurity
pump-out effect in the ion root is attractive for the helical
reactor because no impurity accumulation is expected in
the high-Ti core plasma.
The SDC/IDB high-density plasma allows us topropose a high-density ignition scenario. In this ignition
approach the temperature requirement is reduced below
10keV at a high-density above 5x1020m-3, and the helicalripple transport is mitigated together with the reduction of
the engineering demand caused by the high-temperature
operation. Including this scenario, most of the LHD resultspresented here are relevant to the steady-state helicalfusion reactor, and the upgrade of the LHD is now planned

including the deuterium experiments for investigations of
the high-performance plasmas, which should lead to adefinite design of the LHD-type fusion reactor.

Acknowledgements
The authors greatly acknowledge all contributions
from overseas and domestic collaborators for the progressof the LHD achievements. They are also grateful to thetechnical staff in the LHD for the excellent operation of the
LHD and the heating systems.

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