Recent Results ofHigh temperature mode, EC/ICRF heatingand Steady state operationin LHD

T. Mutoh
National Institute for Fusion Sciencethe 7th Integrated Operation Scenarios Topical Group Meeting
Kyoto University, Kyoto, Japan, 18th – 21st October 2011

1Outline•Introduction of LHD Project•High temperature mode experiment•ECH experiment•ICRF
ICRF Heating and Steady State Operation•Future plan2

LHD is a unique and excellent confinement device

Heliotron : Net-current free plasma suited to Steady State Reactor
Originally developed by K. Uo ( Kyoto Univ.)
Goals of the LHD Program
1. Clarification of physics of fusion relevant plasmas in heliotron
configuration
2. Demonstration of scientific feasibility
y of steadyy state helical reactor
3. Comprehensive understanding of toroidal plasmas==> complementary approach to tokamaks
Summary of recent achievements in LHD
 Te: 20 keV,  Ti: 7 keV,  ne: 1.2x1021m-3,  nTi: 5.2x1019m-3skeV3

The Large Helical Device (LHD)

World largest heliotron
External diameter 13.5 m
Plasma major radius 3.9 m
Plasma minor radius 0.6 m
Plasma volume30 m3
Magnetic field
3TTotal weight1,500 t

Superconducting coil system
magnetic energy
1 GJCCryogenici mass ((-269269 ddegree C) 850 tTolerance< 2mmConstruction 1990-1998
First plasma 1998(March)

4

Plasma generated in LHD shows steady progress- LHD supports ITER operation by complementary research Quasi-steady state high beta-- LHD results --LongpulseforLl : > 1 keVk V plasmalf 1 hourhplasma kinetic pressurepressure of magnetic fieldLHD<> = 5.1 % at B = 0.425 T<>  5 % is maintained for more than
100 energy
gy confinement time EHigh density
ne(0) = 1.21021m-3
1.5 atmosphericpp
pressure at B = 2.5 T an innovative concept ofignition at T(0) = 6-7 keVg ion temperaturepHighTi = 6.4 keV at ne = 1.61019m-3
accompanied phenomena to expel
impuritiesn E T = 5  1019 m-3 s keV (%)

Tokamaksus/E

Very high density operation
ne exp (10200 /m3)

High beta :LHD

T o kam ak

1 .0

A lca to r C
D IIIPBX

0 .1
0 .1

1 .0scaln en102033) )(1020/m/m(10GW5

High-Ti,Te mode experiments are helpful for comprehensive
understanding
g of toroidal plasma
pin low collision regimeg Ion ITB with expelling of high Z ions•••Ti0 > 6 keV and improved confinement
Low thermal conductivity with low viscosity
Impurity hole formation

 Electron ITB with NC electron root((CERC))••

High-power gyrotron development
Extended high-T
g e regime
g6

Ion ITB is formed in wide area of NBI heated plasmas
and Ti0i > 6 keV is achieved(c) #101911,t = 4.04 s-3m ]76.4 keV

5TiCXS7eTi [[keV]e, T [keV], n [x101964Te

3ne2Ti10-0.6 -0.4 -0.20r

•

••eff

0.2

0.4

0.6[m]

An ion ITB with a steep gradTi and flat
density profile is formed after carbonpellet injection.
Ti0=6.4 keV is achieved. (2010)
The carbon density of nC/ne lessthan1% is achieved (Impurity hole
M. Yoshinuma, NF, 49, 002002 (2009),
formation)
K. Ida, PoP 16, 056111 (2009)7

Impurity Hole is formed with ion ITB
M. Yoshinuma, NF, 49, 062002 (2009), K. Ida, PoP 16, 056111 (2009), T. Ido, PPCF, 52, 124025 (2010)010

SX image (Fe)#97011(He),#97030(C, Ne)

heliumICConcentrationn n /ne-1

10-2

carbon10-3

10neon-4

10

0

0.2

0.4

0.6


0.8

11.2

• Extremely hollowed profile of impurities (Impurity Hole) forms in ion ITB plasmas.
• Heavier impurity ions form steeper impurity hole profile.
• Impurity hole forms after ion ITB formation (enhancement of Ti gradient)• Outward convection of impurity is considered to be driven by turbulence, but still anopen question.

Simultaneous realization of Good energy confinement and Poorimpurity confinement is not obtained in tokamak plasmas yet. 8

Negative Er during ITB was Confirmed by HIBP
K. Nagaoka, NF, 51, (2011)

55Ti [keV]

4352

Er [kkV/m]potentia
al [kV]

4Er (HIBP)

1

30
3.6 3.8 4.0 4.2 4.4 4.6 4.8

R [m]

2Er (CXS)Er (neoclassical )0-5

10-0.2

00.2

0.4

0.6

0.8

1-10-0.2

00.2

0.4

0.6

0.8

1r/ar/a• Weakly hollowed profile of electrostatic potential in the core is observed by HIBP
• Observed negative Er in the ITB region is consistent with neoclassical theory.
• Positive Er appears at periphery, which is marginal in neoclassical calculations.

Ion ITB == > Improved confinement with negative Er without
significant
i ifi
t Er shear
hlikelik observed
bd in
i tokamakt kk ITB.
ITBcf. electron ITB (CERC) forms with the transition to positive Er from negative Er9Peaked Profiles of Ti and V in ion ITB Plasma on LHD
K.
10 Nagaoka, NF, 51, 083022 (2011)10622

0 [[m /s]

i [[m /s]

81

NC t = 2.37 sec
t = 2.01 sec
NC t = 2.0 sec
t = 2.16 sec
t = 2.36 sec

4
20

0.0

0.2

0.4

0.6

r/a

0.8

1.0

reff/a99=0.2602426

i [m /s]

810

• The peaked Ti and V profiles are simultaneously formed in ion ITB plasmas.
• i decreases down to neoclassical level in ion ITB plasma, indicating significantreduction of anomalous transport.
•MomentumMt
transporttt (viscosity)( iit ) was analyzed
ld withith ththe neoclassicalli ld
damping
i effect
ff t(collision with ripple trapped particles) taken into account.
•The viscosity (0) decreases with a good correlation to thermal diffusivity (i).
•Although spontaneous rotation is not taken into account in this analysis
analysis,a spontaneous rotation driven by grad-Ti was also identified in the ion ITB core.
M. Yoshinuma, NF 49, 075036 (2009). K. Ida, NF 50, 064007 (2010)10Contents•Introduction of LHD Project•High temperature mode experiment•ECH experiment•ICRF Heating and Steady State Operation•FutureFuture plan

1111/26

ECH system for LHD (2006->2011)
Heating Equipment Room
3 - 77GHz
U-Antennas
1- 168GHz 1 MW, 5s
500 kW ,1s
1s
1 - 84GHz
O-Antennas
200kW,CW
3-88.9mm
non-evacuatedwaveguides
LHD Hall

LA t
L-Antennas
2- 31.75mm evacuatedwaveguides

3-88.9mm evacuated1 - 84GHz
800kW,3s

2 - 82.7 GHz
450kW,2s

9 gyrotrons, transmission Lines, antennas are operated.
1 168GHz(Toshiba) 1-84GHz
1-168GHz(Toshiba),
1 84GHz CPD (GYCOM)(GYCOM), 1-82.7GHz
1 82 7GHz non CPD (GYCOM)
3-168GHz is replaced by 3- 77 GHz(NIFS-Tsukuba Collaboration)
1-84GHz CPD (GYCOM) 200kW 1000 s (diamond window)
2 evacuated 31
231.75mm
75mm corrugated waveguide system.
system3 non-evacuated 88.9mm corrugated waveguide system.
123- 88.9mm evacuated, corrugated waveguide upgraded for 77 GHz.

Achievement of 1.8 MW/ 1 s with new 77GHz gyrotron
T. Imai, IAEA-FEC 2010

 New technique of anode bias voltage ramp up

100
P [MW]  [%]80‐> StabStablee aandd highg popowere opeoperation!
at o !
1.8 MW/ 1 s
World’s highest output power(>1s)

BeamCurrentPuls
eWidthMOU
OutGyrotr
on Out77.4 A

0.1 s

1.87
MW

1.97
MW

72.2 A

1s

1.77
177
MW

1.86
186
MW

Power [kkW]

Stepwise VA

1Regular V60

A

40

0.5

2000

1020 30 40 50
Beam current [A]060

13electric efficien
ncy  [%]

151.5

Electron Internal Transport Barrier (e-ITB) formation
e-ITB formation was attained by centrally focused ECH.
 The foot point locates near the rational surface of/2 = 1/2.
-> Lower-order rational surface may trigger ITB.
 There is the threshold in ECH power.
-> Local parameter control plays a key role for ITB
formation.

Prospective effect by 77 GHz heating Expansion of central heating source.

gyrotron-> Improvement of core temperature/ pressure.
 Different frequency from those of existing gyrotrons.
-> simultaneous
simultaneous-multi-point
multi point heating of plasma-> Local control of radial electric field/ radialtransport by local collisionality control.
-> Influence of deposition power/ position on theITB formation.
14

Achievement of Te0 = 20 keV by increased ECRH power 3.7 MW of ECRH power was injectedjfor plasma generation
gandsustainment.
 Te0 = 20 keV was successfully achieved using centre focusedECRH.
ECRH
 Electron temperature increases in higher density regime.

15

Contents•Introduction of LHD Project•High temperature mode experiment•ECH experiment•ICRF
ICRF Heating and Steady State Operation•Future plan1616/26

SSO by H minority in He majority ions
Experiment condition is same with ITER non-activated half B, case 4 (4He)• Power control during plasma discharge with watching sparks in vacuumvessel• Rax = 3.64- 3.67 m, ne～ 0.4x1019 m-3, Te ~ Ti ~ 1 keV• Plasma was terminated by sudden increase of density caused by influx ofiron impurity
17

Shot 75606

2PICRF

10
3.68
3.66
3.64
3 62
3.62
3.60
1.5

Te [keV]Ti0 [keV](ECEE@3.318m)

ne [1019m-3]

Rax [m]Powwer [MW]Higher power operation ( P > 1MW ) for SSO
PNBI

PECH

near antenna
f antenna
far

1.005
0.50.02102100100

200300400Time [s]500600

B=2.85T(@R=3.6m)
q = 38.47MHz))PICRF=894kW ((freq(3.5U, 7.5UL antenna)
PNBI=75kW
PECH=90kW
Ptotal=1059kW
Tduration=525sec

18

HAS antenna is installed to suppress edge
heat loss byy controllingg of k//previous
Poloidalantennastrap width：20cm
strap length：75cm
antenna width：54cm
water cooled strap andwater-cooled
backplate• Faraday shield is madeb Cu-SUSbyC SUS cladl d• Side protector is madebyy CFC• Movable in horizontal
direction••••

19

HAS antennaRF power is strongly concentrated in front of theantenna with reverse phasestronghorizontal plane of plasma

center plane of the right strap

S(kW/m2)
300
42
weak
Poynting vector

In‐phase

Antenna loading is approximately same,same and so thePoynting vector is widely scattered.

30.420.003

The Poynting vector is strongly concentrated in front of theantenna in the reverse-phase.
Poynting vector

reverse‐phase20

Higher density and long pulse operation is on-goingHigh performance plasma of steady state is required as mission plan
nevsPRF4.0

ne [1019m-3]+ P ECH =0.37MW

3.0 HAS antenna(0, )

2.0(0,0)10
1.0Poloidalarray antenna

0.0
0.0

0.5

1.0

1.5

2.0

PICH+PECH [MW]2.5

By using four antennas (a pair poloidal array
antenna, a pair of HAS antenna),
ne of 3.7x1019 m‐3 and Ti ~ Te ~ 1 keV isstablyy sustained.
Pulse length can be elongated in thisexperimental campaign.
21Contents•Objectives of LHD Project•High temperature mode experiment•ECH/ECCD/EBW experiment•ICRF Heating and Steady State Operation•Contribution
Contribution to ITER operation•Future Plan in LHD

22

Future Research Plan in LHD
Extension of operation regime
 DD experiments• Confinement improvement due to isotope effects• High ion temperature• Investigation of high energy particle behavior
 Closed divertor operation• Confinement improvement• Steady state operation
 Steady state operation• Increase of heating power of ICRF and ECH
Researches for physics understanding support an ITER operation
 Further understanding of 3-D plasma physics• Transport• MHD• High energy particles•-23