21 st IAEA Fusion Energy Conference Summary Innovative Concepts Confinement & performance

21st IAEA Fusion Energy Conference- Summary Session

21st IAEA Fusion Energy ConferenceSummary

Innovative ConceptsConfinement & performance

Jerome PAMELA, EFDAWith the kind support of A.Becoulet (CEA), D.Borba (EFDA) and R.Kamendje (EFDA)


Innovative Concepts

- Z-pinches- field reversed configurations- spheromac formation by steady helicity injection (HIT-SI) - magnetic dipoles (or Ring Trap): SC rings levitated / several seconds to minutes operation

X-Divertor concept proposed to enhance the divertor thermal capacity

RT-1 / study of Jupiter’s magnetosphere


1

10

100

0 2 4 6 8 10rc (cm)

0 2 4 6 8 10

Ti

(ke

V)

0

1

2

3

4

5

rc (cm)

Mirror experiments

T. ChoT. Cho et al.et al. EX/P7-14EX/P7-14;; Phys.Phys. Rev.Rev. Lett. Lett. 97 (2006) 055001;; Phys.Phys. Rev.Rev. Lett. Lett. 9494 (2005) (2005) 085002085002

Ti Increase

TurbulenceTurbulence

Cylindrical ExB Sheared Flow due to off-axis ECH ControlsControlsRadial Transport Barrier, which improves Core Plasma confinement.

ECH (250 kW) raises Te0=750 eV with Ti0=6.5 keV and Till0=2.5 keV

X-Ray Tomography

SuppressSuppress

GAMMA 10 (Japan)GAMMA 10 (Japan)

Without

With ExB Sheared

Layer (b)

ECH Produced LayerECH Produced Layer Radial Transport Radial Transport BarrierBarrier

i /

i-c

lass

ic

Diffusivity

i

GOL3 (RF):Te ~2-4 keV ne ~ 3 1020 m-3

nT ~1018 m-3 s keVPotential for testingPFCs (ELMs, disrup.simulation) tbc.

GDT (RF):Te ~0.2 keV ne ~ 5 1019 m-3

With 4MW NBAgreement with theory

Hanbit (US):Test divertor stabilising m=-1 MHD flute-instab.


Toroidal Magnetic Fusion DevicesMain recent hardware improvements on Tokamaks

• EAST started• JT60

– FST inserts• JET

– MKII-HD Divertor (high delat, 40MW capability)

– NB heating – Several diagnostics

• DIII-D – NB reconfigured – Lower divertor modification

allowing balanced DN operation with imporved density control

– diagnostics

• AUG– W-coverage extended to 85% of

plasma facing components– New integrated control and data

acquisition system– diagnostics

• MAST– JET-type NB PINI– Pellet injection– Diagnostics and control systems– Divertor upgrade

• NSTX– Modified divertor PF coils (high ) – Diagnostics and control

• C-Mod– boronisation

• FTU– Li Liquid Limiter


ASIPP

EAST first plasma on 26 September 2006

• Largest operating fully SC Divertor tokamak• Nominal parameters Bt=3.5T, Ip= 1MA,

R=1.75m, a=0.4m, single/double null• 4.5 s shots already achieved


With FSTs

Installation of Ferritic Steel Tiles

=> Reduction of fast ion lossesby 1/2~1/3 at 1.6T

JT60-U: ripple reduction by Ferritic Steel Tiles

Without FSTs

Larger Pabs at given Pin

=> smaller required NB units for given N

=> better flexibility in NBI combination=> better flexibility of torque profile

Smaller inward Er

=> less ctr-rotation


AUG : 85% W plasma facing components

● in 2005 “thin” W coating of - 4 guard limiters at LFS (water cooled) - 8 ICRH antenna side limiters - top of bottom PSL - roof baffle

● in 2006 - upper and lower ICRH limiters - W coated bottom target tiles (200 m)

full tungsten machine in 2007


JET: Divertor modified ITER plasma shape accessible at high current (up to 3.5 MA)40 MW power capability

Hei

gh

t

(m

)

R (m)

Up to 32 MW plasma heating achieved recently


DIII-D: Reoriented NB box allows more

relevant balanced injection heating


Transport/Confinement Physics Experiments

Specific Stellarator studies-Comparison of quasi-helically symmetric (QHS) configuration to broken symmetric configurations on HSX / confirmation of improved confinement in QHS configuration

-Extensive Core Electron Root Confinement (CERC) studies in helical devices (CHS, LHD, TJ-II and W7-AS) / collisionality threshold depends on magnetic configuration and ECH power

- Studies of effective ripple on confinement (Heliotron-J, LHD)

Other studies reported below


0

0.05

0.10

0 1 2 3 4

lhd_chieff_grb_gmt_disp_12_r360g122_03_eps_eff_temp2

eps_eff(temp)

eff (

2/3)

<dia

> (%)

LHD: study on the Confinement in High- Regime

Degradation can be attributed to global dependence on effective helical ripple to the neoclassical transport not on MHD effect Degradation in high regime will be improved by dynamic Rax control by vertical field in nearest future

Outward shift of plasma by Shafranov shift causes an increase of the effective helical ripple

0

1

2

0 1 2 3 4

Rax

=3.6m, Apa=6.3

Rax

=3.75m with LID, IDB

E

exp/

E

ISS

95

<dia

> (%)



Rotation and Confinement


Tore Supra: toroidal rotation observed with ICRH in correlation with confinement improvement (L-Mode)

• Suggests sheared rotation• Could explain confinement improvement through ITG and TE modes

stabilization (Kinezero)

TORE SUPRA


JT60-U: Pedestal parameters and confinement improved with co-rotation in H-mode

JT-60U

0.7

0.8

0.9

1

1.1

-300 -200 -100 0 100 200

HH

98(y

,2)

VT(r/a=0.2) (km/s)

ctr-

bal-

co-NBw FSTs

w/o FSTsctr-

bal-co-

co-rotation helps to form stronger Te-ITB


JET: measured poloidal velocity in ITB

• Er and ExB shear much larger with measured V

• Weiland model with measured V (rather than neoclassical) matches experiment better

Measured poloidal velocity in ITB layer (60km/s) highly anomalous, far larger than neoclassical (~5-10km/s)

Ion temperature profiles during ITB formation

Poloidal velocity from charge exchange, during ITB formation

Rmid (m)Rmid (m)

Ti (

keV

)

V

(km

/s)


DIII-D profiles with high / low rotation compared to modelling confirm importance of ExB shear for ITBs


DIII-D: Transport physics sensitive to applied torque / Rotation



Turbulent transport: TEM, ITG, ETG

21st IAEA Fusion Energy Conference- Summary SessionLHD: extensive parametric study of non-local Te rise (cold

pulse propagation) / anomalous transport behaviour similar to that observed on tokamaks / usefully extends experimental data base to test ITG-based models

Larger dTe/dt Stronger edge cooling

simultaneity

Te rise delayed

ne increases


ECH heated plasmas TCV: Influence of plasma triangularity on

transport (L-mode)

Gyro-fluid, gyro-kinetic models

TEM dominant transport (mixing length) predicted to decrease with as observed

O

H


AUG: anomalous transport studies

Pure electron heating: threshold for

TEM at R/LTe3

(Similar observation on JET)

Results reproduced by simulation (GS2)

ITG-TEM transition (Er measured at =0.7): TEM suppressed at high collisionality


ECH heated plasmasT10 : TEM dominates turbulent transport

at low collisionalities


C-Mod: TEM turbulence density fluctuation spectra measured during ITB (ICRH

heating) & reproduced by GS2 simulation


Scattering system measures reduced n/n from upper ITG/TEM to ETG kr ranges during H-mode

Results consistent with modelling

• ITG/TEM stable during H-phase

• ETG modes could be important /Lower growth rate during H-mode

ELMs

NSTX: new high resolution tangential microwave scattering system (280 GHz) allows high resolution of turbulence measurement



ETB & ITB studies


Edge Transport Barrier: Er transitions at plasma edge in tokamaks and helical devices

AUG: Negative Er well observed at

ETB, increases with confinement improvement

- coincides with H-mode barrier gradient

- Doppler reflectometry

AUG

H98(y,2)

CHSCHS

AUG

CHS: Negative radial electric field of Er ~ 10 kV/m observed with ETB formation


FTU: ITBs in L-mode / reduced ion thermal diffusivity


0

20

40

60

3 3.1 3.2 3.3 3.4 3.5 3.6 3.7

0

50

100

150

200Pulse No: 62077

R[m]

P RF

[10

2 M

W/m

3]

Ph

ase

[deg

]

Mode conversion

Fast wave

b)

Am

plit

ud

e [e

V] ITB

JET: Te modulation experiments show ITB as a narrow layer with reduced heat diffusivity

• Modulated RF power deposited either side of ITB (at centre and at R=3.6m)

• Heat wave propagates towards ITB from both sides

• Heat wave amplitude (red) damped strongly when wave reaches ITB

• Phase (blue) rises when heat wave approaches ITB, showing heat wave slows down

• ITB is a narrow layer with reduced heat diffusivity

• Indication of region with turbulence stabilised and loss of stiffness



ITB studies: role of rational q surfaces


0 1Effective radius

SXR profiles

0

0.2

0.4

0.6

0.8

1<

n e> (

1019

m-3

) a)

H (a.u.)

0

0.4

0.8

1.2

1.6

0

20

40

60

80

100

Te (

keV

) Ti (eV

)

-4

-3

-2

-1

0

1000 1050 1100 1150 1200

I p (kA

)

time (ms)

1.9

2

2.1

2.2

0 0.2 0.4 0.6 0.8 1

vacuum1050 ms1100 ms1150 ms

/2

Te

Ti

ne

Ip•CERC triggered by the n=4/m=2 rational

• Changes in both Te and Ti.

• The SXR tomography diagnostic shows a flattening of the profiles localized around ≈ 0.4 with a m=2 poloidal structure. The rational must be inside the plasma to trigger the transition.

TJ-II: role of low order rationals in core transitions

21st IAEA Fusion Energy Conference- Summary SessionJET: ITB forms when qmin exists and approaches (rather than reaches) an integer value

Te at various major radii, R, showing formation of an ITB

• ITB formation starts before q=2 surface enters plasma

ITB formation slightly ahead of Alfvén Cascades (marking qmin= integer)

Case number

t AC-t

ITB (

s)

Pulse No: 61347

Time (s)

Te (

keV

)

Start of ITB formationqmin

reaches 2

• Alfvén cascades seen simultaneously on microwave interferometer, O-mode interferometer, X-mode reflectometer and magnetic probe


M.E.Austin, University of Texas

Zonal flow structures with significant radial extent / ExB shear flow needed

DIII-D: similar observations / explained by GYRO simulation (zonal flows)



Turbulence and Zonal Flows


- a large number of experimental and theoretical contributions- an overview by Pr. Fujisawa

ZONAL FLOWS / A HIGHLIGHT

An example of extremely fruitful interaction between:

Forward looking theoreticians Other scientific Communities

“Smaller” devices of all type (tokamaks, helical devices etc.)flexible and well diagnosed

Larger devices(driving specific diagnosticsimprovements)


Zonal Flow Experiments (Pr. Fujisawa Talk)

- electric field or flow measurements in high temporal and spatial resolution

A challenge to experimentalists

i) zonal structure

symmetry (m=n=0)a finite radial wavelength

ii) nonlinear coupling with turbulence

Discoveries

iii) effects on transport

HL-2A (probes)

JIPPT-IIU (HIBP)

T-10 (HIBP)

ASDEX-U (reflectometry)

JFT-2M(HIBP&probes)

DIIID (BES)

CLD (probes)

CASTOR (probes)

TJ-II(probes)

TEXT-U (HIBP)

H1 (probes)

CSDX (probes)

HT-7 (probes)

CHS(HIBP)

LMD (probes)

TJ-K(probes)

Devices

More than a dozen papers have been published as a PPCF cluster (2006).

21st IAEA Fusion Energy Conference- Summary SessionHL-2A , DIII-D, TJ-II : - Toroidal structure of Geodesic Acoustic Modes (GAMs) observed- GAM interacts non-linearly with ambient turbulence and drives forward cascade of energy to high frequency- energy transfer between global (parallel) flows and turbulence also observed on helical devices

DIII-D

TJ-II

HL-2A


CHS: Energy partition between ZF and Turbulence without/with ITB

A larger fraction of zonal flows contributes to confinement improvement inside

the barrier! Importance of zonal flows on confinement is demonstrated.

CHS two Heavy Ion Beam Probes = powerful core plasma diagnostic

0 1 2 3P

ZF(V

2)

L mode

H modeP

ow

er (

E/

T)

∇~

10-3

No ITB

ITB

At a radius without mean Er-shear

inside the barrier

no ITB

ITB

radius0 1

Pot

entia

l (or

Tem

pera

ture

)

confinement is improvedwithout shear

Common ITB in helical plasmas

Clear difference in energy partition


Extensive studies on several machinesAUG: study of parametric dependence of Geodesic Acoustic

Modes (GAMs)



Scaling


- interplay in H-mode

Matching plasma shape, poloidal p*, p and an/T2 provides a constraint on the exponents in the power law scaling:

MAST-DIII-D comparison

Constraint is consistent with: - gyro-Bohm Scaling (x = -3) - weakly favourable collisionality scaling (as observed in MAST & other devices) - - interplay in accord with that derived from the database analysis

MAST and NSTX: scaling studiesNSTX & MAST in ITPA data base: E ~ 1.03 as compared to 98y,2 ~ 0.58

Dedicated scans on NSTX show E ~ BT0.9 Ip

0.4


Confinement in LHD Improved w.r.t. ISS95 scaling

0

0.05

0.10

0 0.05 0.10

gas puffpellet w/o IDBpellet w. IDB

Ee

xp (

s)

EISS95

Rax=3.75m

95 2.21 0.65 0.59 0.51 0.83 0.42 / 30.079ISS

E ea R P n B

Energy confinement time exceeding the ISS95 scaling


scaling experiments in L mode :favourable weak dependence on as seen in

H-mode (JET & DIII-D 20th FEC)

• Weak degradation. exponent: ~ -0.2– ITER L-mode scaling -1.4

• Supported by density fluctuation measurementsTORE SUPRA



Density peaking


Combined JET-AUG database on density peaking in ITER Baseline ELMy H-modes / reduced colinearities between physics variables

Significant density peaking expected on ITER

• Peaking requires anomalous particle pinch / under investigation by theoreticians

• Scaling of density peaking to ITER ne0/<ne> ~ 1.4

• Impact on impurities requires full assessment

Favourable for fusion power, bootstrap fraction, density limit

eff

n 0/<

n>

vol

TCV: Stationary ELMy H-modes, eITBsDensity profiles are peaked despite pure electron heating and no core source Te/Ti~2 and N~2

ITE

R


H-mode Core impurity peaking can be controlled with central electron heatingTransport of impurities is turbulent

0 0.1 0.2

1

10

100

peak

ing

of c

W

PECRH/Ptot

No

rmal

ised

Ni

pro

file

58144

58149

r/a

Minority (ion) heating (MH)

Mode conversion (electron) heating (MC)

JET AUG

Turbulent transport models show peaking dependence on Z and anomalous behaviour of high Z impurities Transition ITG to TEM could explain Ni peaking on JET


Long pulses, steady state and Real Time Control (performance)


JT60-U : Fully Bootstrap-Driven Discharge

Ip = 510 kA was maintained for 1.3s with fBS ~ 1 (with net INB = -35kA)

Comment: q95 still very high (>10) => higher current demonstration needed

slow decrease of Wp & Ip

JT-60U


NSTX: progress in current sustainment NBCD and p provide up to 65% of Ip

Relative to 2004, High N H89P now sustained 2 longer

116313G12

TRANSP non-inductive current fractions

Long Pulse Operation is a challenging issue for STs


0200400600800

power_66053PICRF [kW]PECH[kW]PNBI[kW]

Po

we

r PICRF

PECH

Shot 66053

[kW

]

00.20.40.60.8

firc@66053ne_bar

[101

9 m-3

]

ne

0

1

Ti_Te_66053 Ti(keV) 3.466(m)

Ti0

Te(ECE:R=3.466m)

Ti,

Te

[keV

]

0100200300

66053_6I_Div1-6Div#1(6I)

Div#6(6.5L)

T

div

[oC

] Tdiv(6.5L-I)

Tdiv(6I-U)

0100200300

66053_7I_Div1-6Div#1(7I)

Div#6(7.5L)

T

div

[oC

]

Tdiv(7.5L-I)

Tdiv(7I-U)

20304050

wall_66053_3points#6_FW1#7_FW3#10_FW9

Tw

all

[oC

] 7I

7.5U

9I

3.63.623.643.663.68

0 500 1000 1500 2000 2500 3000 3500

rax_66053 rax_by_PIV[m]

Rax

[m]

Time[s]

020406080

T_feedthrough_660533.5U7.5U7.5L

0 500 1000 1500 2000 2500 3000 3500T

fee

d t

hro

ug

h

[oC

]

3.5U antenna

7.5U antenna

7.5L antenna

Time[s]

Record of input energy 1.6 GJ achieved on LHD(Tore Supra 1 GJ at 20th FEC)

LHD: 54-Minute plasma operationLHD: 54-Minute plasma operation

ALSO: 31-minute long discharge with 680 kW ICRH power,Te(0) and Ti (0) of 2 keV at ne of 0.81019m-3

Rax= 3.67-3.7m, B=2.75T, PICRF= 600-380 kW, PECH=110 kW


HT-7: steady state operation

Steady-state alternating current (AC) operation Ip=125kA

ne(0)= 1.5-2.51019m-3

Te= 500 eV

30 seconds with LH

53 seconds w.o. LH

ALSO: Steady state “standard” long pulse achieved> 6 minutes Ip=60kA ne(0)= 0.8-1 1019m-3 PLH= 150kW


JT60-U: Real time qmin control with MSE diagnostics and LHCD

JT-60U

qMSE qmin qmin,ref

PLH

LHMSE

Real time qmin control scheme

0

10

20

0

0.8

1.6

0

0.5

1.0

1.5

1

1.5

2

7 8 9 10 11 12 13 14 15Time (s)

q min

PLH

(M

W)

PN

B (

MW

)

N

ref.

command

02468

Te

(keV

)

MSELH

Transport reduction at t=12.4 s

Time delay in response of qmin

jOH or jBS change

r/a~0.20.4

0.6


Tore Supra: Integration of real time profile control (LH power and n//) and InfraRed-avoidance scheme

Ip [MA]

PLH [MW]n//

HXR profile width

0.3

0.4

0.5

0 10 20 30

36194 (BT=3.7T ; <ne>=2.5.1020 m-3)

PICRH [MW]

40 50

qo

1

2

3

2

2.5

0.5

1.0

1

2

1

1.6

1.4

1.2

1000

9000.5

1.0 PQ1 [MW]

TIR [Deg]

Time [s]

VLoop [V]

With « Search optimisation » algorithm

Target: broadest HXR profile

ICRH antenna IR view

Antenna septum

Integrated stationary scenario achieved with:

• Constant Vloop

• qo increases by 0.4

• No MHD detected

HXR width n// and PLH

PLH , PICRH IR temperature

PLH

n//

Start

Optimum found

TORE SUPRA

21st IAEA Fusion Energy Conference- Summary Session JET: Advanced Real Time Control

Real Time Control p(r) and q(r) profile reported at 20th FEC

=> New Dynamic-Model Approach for Simultaneous Control of Distributed Magnetic and Kinetic Parameters

4 actuators (NB, IC, LH & PF) Dedicated experiments to determine matrix coefficients

2 time scale controller

2 feedback loops and one feed-forward (disturbance rejection)


Not mentioned here Real Time Control of MHD modes (NTM, RWM)

DIII-D, AUG, JT60-U, NSTX, RFX….

(see Zohm Summary)


Summary on Plasma Scenarios


MST (Reversed Field Pinch): improved ion confinement

Improved-confinement RFP– apply inductive current profile control reduce MHD activity– reduced stochasticity, x10 confinement improvement, E ~ 10

ms– limitation: inductive technique is transient

=> Improved ion confinement• sustained Ti ≥ 1 keV, E,i ≈ 10 ms

• fast ion confinement time > 20 ms

=> high operation with pellet injectiontot = 26%


t=1.36s

Wp =1.1MJ, Pabs=10MW

nET =4.41019m-3 s keV

(0) =4.4 %, <>=1.5 %Rax =3.75m, B=2.64T

0

1

2

3

4

5

0.0

0.5

1.0

1.5

2.0

-1.0 -0.5 0.0 0.5 1.0

ne10^20[m^-3] Te[keV]

n e (1

020m

-3) T

e (keV

)

ne

Te

t=1.136s

LHD: internal diffusion barrier (IDB) leading LHD: internal diffusion barrier (IDB) leading to high density operation: n(0)~5 10to high density operation: n(0)~5 102020 m m-3-3

Effective Core fueling by pellet Effective Core fueling by pellet injection combined with Local injection combined with Local

Island Divertor (LID) Island Divertor (LID)


0.7 < Ti0/Te0 < 2.5 at high density Extension of data base at low collisionality

0.0

0.5

1.0

1.5

2.0

2.5

Ti0/T

e0

0.2 0.4 0.6 0.8 1.0<ne>/nGW

NBI only

NBI+ICRH

ASDEX Upgrade improved H-modes

0.6

0.8

1

1.2

1.4

1.6

0 5 10 15 20*/*ITER

H9

8(y

,2)

ASDEX Upgrade

Improved H-modes

H-modes (Type I)

ITER

Towards more ITER-relevant conditionsAdvanced H-mode (hybrid mode) on AUG

21st IAEA Fusion Energy Conference- Summary SessionTowards more ITER-relevant conditions

DIII-D: High H-mode performance achieved with reduced plasma rotation (balanced NB injection)


H-ModeC-Mod performance improvement with boronisation

Improved pedestal results in improved global E through profile stiffness

Record pressure in tokamak

<p> = 1.8 atm

at N = 1.74


H-ModeJET type II ELMy H-mode, similar to AUG with QDNalthough still at relatively low current

• ELM behaviour constant over pulse• Very fine scale activity - distinct ELMs almost indistinguishable

Time (s)28 29 30 31 32

(a.u

.)M

J1

019

m-3

D

<ne>

H98 0.95

Wtot

nped also constant

0

10

2.5

67911

Turbulent magnetic

fluctuations coincide with

D bursts

1ms

D

Magnetics


Advanced ModeseITB performance on TCV

• High confinement obtained with high bootstrap current fraction and pol

• In eITB region ne/ne~0.5Te/Te (thermo-diffusive pinch)H

RL

W ~

E/

L-m

od

e

bootstrap current fraction

21st IAEA Fusion Energy Conference- Summary SessionAdvanced Modes

JT60-U: High p ELMy H-mode plasma improved by FSTs sustained for 23 s (~12R)

HH98(y,2)=2.2 fBS=36-45% q95~3.3 Ip = 0.9 MA, BT = 1.6 T

0 5 10 15 20 25 30 35Time (s)

0

3

12

1

0.5

0

0123

151050

4

2

0

I p (

MA

) N

, H

H98

(y,2

)

n e(1

019 m

-3)

PN

BI (

MW

)I D

(

a.u.

)

N

HH98(y,2)

JT-60U

R=0<>a2/12 : D.R. Mikkelsen, Phys. Fluids B 1 (1989) 333.

ITERBaseline

ITERHybrid

N H

H98

(y,2

) after 20th IAEAw FSTs

before 20th IAEA, w/o FSTs

.

1

1.5

2

2.5

3

0 5 10 15 20 25 30 35Sustained duration (s)

q

ITER-SS(I)ITER-hybrid (A C C Sips, et al., PPCF 47 (2005) A19.)w FSTs (45436@18s)w/o FSTs (44092@15s)

HH98(y,2)N

fBS

fCD

Fuel purity

Prad/Pheat

ne/nGW

0.56

0.83

1.32.56

0.5

1

0.77

21st IAEA Fusion Energy Conference- Summary SessionAdvanced Modes

JET : ITB plasmas with active ELM control at >30MW with neon seeding (~60% radiated

power)

ELM control with Neon (4-8s) PNB+LH+IC~30 MW Prad~17MW

B~3.1T, I~1.9MA, q95~5 Wdia~5.6MJ, N~2

JET AT database quasi-stationary (/E>10) pulses at high N and high

3.5

3.0

2.5

2.0

1.5

1.0

0.5

0.0 N

50403020100

PNBI+PICRH+PLHCD+Pohm [MW]

2006 2000-2004

5q95 6, 0.3

Bo>3T, E>10Target for JET-EP2AT regimes with 45MW planned power upgrades

Divertor radiation

Between ELMs (no neon seeding)

With neon seeding

R (m)

Z (

m)

Z (

m)

Time (s)

Pulse No: 67869 1.9MA/3.1T

6

4

2

0

[1019

m-3

]

108642Time [s]

neo

D

20151050

PNBI

PICRH

Prad

[MW]

3210

PLHCD [MW]

Ip [MA]

10

5

0

[keV]

Teo Tio

JET Pulse N° 67869

N


Advanced ModesDIII-D : High performance advanced scenarios with reversed

shear and high non-inductive current fraction


Ip (MA)

PNBI (MW)

PICH (MW)

D

N

H98(y,2)

# 20449

odd n

even n

1.2

08

0 0 2 4 6 time (s)

0

3

1

Hybrid Mode/ towards high performanceAUG: q95 range of advanced H-mode extended

1.2MA/2.0T q95 = 3.17• NBI used with beta feedback,

50% of NBI is off-axis.

Central ICRF heating.

• <ne>=6.4x1019m-3, Ti0/Te0=1.4

<ne>/nGW=0.42, */*ITER=2.

• H98(y,2) rises to 1.4 at

N=2.9

• CW,core = 2.5x10-5 (< 10-4).

Core MHD: (1,1) fishbones.

(4,3) NTMs.

NO sawteeth.

Early (3,2) NTM when N ~2.

21st IAEA Fusion Energy Conference- Summary SessionHybrid Mode/ towards high performance

JET: Hybrid modes at low q95~3 reach N~3

• Improved hybrid performance at low q95~3, slightly better than H-mode with controlled in real time

1.8MA, 1.7T, N~3, q95= 3.2

PLH (MW)PNBI (MW)

N

4xli

qo (MSE)

n/nG

H

Ip (MA)

Time (s)

Pulse No: 67940 1.8MA/1.7T

No sawteet

h

Hybrid and H-mode in ITER-like shape

Maintain q0>1 to avoid

sawteeth

Bootstrap current ~ p

q95~3Hybrid 2006

H-mode 2006

q95~4

Fig

ure

of

mer

itH

89 N

/q95

2

Hybrid 2003


ConclusionPhysics (transport and confinement)

• Density peaking: – significant peaking expected on ITER – a puzzle to theory and modelling

• Weak (inverse) dependence of confinement on beta consolidating : favourable to ITER

• A hot topic: role of rational surfaces on ITB formation– Could be linked to Zonal Flows

• Importance of rotation & rotation shear confirmed / growing interest towards more ITER relevant rotation patterns

• Extensive studies of turbulence (TEM, ITG, ETG)• Dramatic progress on physics of Zonal Flows and their

interplay with turbulence


Conclusion: Plasma scenariosStellarator• LHD ne(0) = 5 1020 m-3

Tokamaks: ITER relevant scenarios• Real Time Control expanding & getting more integrated• Operation at more ITER relevant parameters (rotation, *, Te/Ti, n/nG) • ELMy H-mode

some progress towards type-II ELMy H-mode (low Ip yet / no scaling available)C-Mod: 1.8 Atm in H-mode

• Progress on ITB regimes– G=H89N/q2

95 up to 0.4, above ITER target, in steady state with dominant NI current

• Rapid developments on Hybrid regimes– extension of operational space

in particular demonstration at q95 ~3 (improved fusion performance)– G=H89N/q2

95 > 0.45 in steady state regimes with 0.3-0.5 B.S. current

*, N scans needed to develop scaling to ITER => IN GENEVA !

Documents

21 st IAEA Fusion Energy Conference Summary Innovative Concepts Confinement & performance