The Physics Base for ITER and DEMO Hartmut Zohm Max-Planck-Institut für Plasmaphysik, Garching, Germany EURATOM Association Hauptvortrag given at AKE DPG.

The Physics Base for ITER and DEMO

Hartmut Zohm

Max-Planck-Institut für Plasmaphysik, Garching, Germany

EURATOM Association

Hauptvortrag given at AKE DPG Spring Meeting, Bonn, 15.03.2010

• main topics in fusion plasma physics

• requirements for ITER and DEMO

• present status of physics research

• summary and outlook

Fusion Reactor in a Nutshell

4/5*Pfus escape as neutrons and hit the first wall(Blanket = tritium production

and energy conversion)Neutronics – talk by A. Klix

1/5*Pfus + Pext escape in charged particles along B-field lines and hit the wall in a narrow band

Plasma wall interaction – talk by B. Unterberg

Core plasma @ T=25 keV,n=1020 m-3 produces Pfus:D+T = He + n + 17.6 MeV

Plasma physics – this talk

Transport determines amount of heating needed to obtain required T

E = Wkin/Ploss (Ploss is the power needed to sustain the plasma)

experiments measured relative to multi-machine scaling: H=E,exp/E,scal

Stability determines the limits to kinetic pressure (Pfus ~ n2T2 = p2)

= pkin/pmag = 20 pkin / B2 (dimensionless pressure)

experimental progress measured relative to ideal MHD limit N=/(I/(aB))

-heating should largely compensate Ploss in a reactor

Q=Pfus/Pext, since P = Pfus/5, the fraction of -heating is P/Ploss=Q/(Q+5)

Exhaust characterised by the ratio of power in charged particles to themajor radius, P/R (since the power deposition width is roughly constant)

Main Areas of Fusion Plasma Physics

• main topics in fusion plasma physics

• requirements for ITER and DEMO

• present status of physics research

• summary and outlook

H and N determine machine size

ITER (Q=10)

DEMO (ignited)

4295

342

1 Aq

RBcP N

fus

4295

342

1 Aq

RBcP N

fus

•N does almost not enter into Q, but strongly into fusion power

• high H helps to achieve ignition, but does not enter in fusion power.

15

5

7.37.21.0

53.31.3

1

2

BRH

Aqcc

Q

N15

5

7.37.21.0

53.31.3

1

2

BRH

Aqcc

Q

N

ITER (N=1.8)

DEMO (N=3)

Major radius R0 [m] Major radius R0 [m]F

usi

on

Po

wer

[M

W]

DEMO should have reasonable pulse length

• Tokamak: poloidal field from plasma current sustained by transfomer: intrinsically pulsed unless clever tricks are played

• Stellarator: all fields from external coils, intrinsically steady state (but at least 1.5 steps behind in evolution)

Tokamak (ASDEX Upgrade, JET, ITER)

Stellarator (Wendelstein 7-X)

Intrinsic thermoelectric current (‚bootstrap current‘) – needs high

External current drive (e.g. by RF waves) consumes additional power

• ‚offset‘ generated by external current drive calls for large unit size

• this in turn aggravates the exhaust problem in terms of P/R

fCD=0.3

fCD=0.2

fCD=0.1fCD=0

N=3

N=4

fCD=0.0fCD=0.1fCD=0.2fCD=0.3

Noninductive current drive in a tokamak DEMO

Fusion power [MW]Pulse length [s]

Net

el.

po

wer

[M

W]

Rec

ircu

lati

ng

po

wer

fra

ctio

n

Summary: what is required for ITER / DEMO

ITER DEMO

H 1-1.2 1.2-1.4

N 2 4-5

Q 10 50

P/R 20 65

Reality check: how does this compare to present experimental data base?

• main topics in fusion plasma physics

• requirements for ITER and DEMO

• present status of physics research

• summary and outlook

Confinement of plasma core - transport

Experimental result:

• Anomalous transport by turbulence: , D a few m2/s

• Tokamaks: Ignition expected for R = 7.5 m for H~1

collision

Transport to the edge

Simplest ansatz for heat transport:

• Diffusion due to collisions

rL2 / c 0.005 m2/s

E a2/(4

• table top device (a 0.2 m, R 0.6 m) should ignite!

discharges withturbulence Suppression

The H-mode: a transport barrier in the edge

H-mode edge: turbulencesuppressed by sheared rotation

• steep edge gradients of T and n

• T higher in whole plasma core (‘profile stiffness’)

H-Mode is standard operational scenario foreseen for ITER (H=1)

Scenarios with improved confinement (H>1)

Improved H-mode = optimisedH-mode scenario (H = 1.2-1.5)

• potential for very long pulses (‘hybrid scenario’)

ITB (Internal Transport Barrier)scenario (H 1.5)

• potential for steady state (‘advanced tokamak scenario’)

The next step: studying -heating

Core plasma parameters sufficient to generate significant fusion power

• study plasmas with significant self-heating by -particles in ITER

• needs P = 1/5 Pfus >> Pext, so it necessarily is closer to a reactor

We expect to see qualitative new physics:

• self-heating nonlinear - interesting dynamics

• suprathermal -particles population can interact with plasma waves

We can have a ‘preview’ in machines of the present generation

• pilot D-T experiments (JET (EU), TFTR (US))

• suprathermal ions generated by heating systems simulate -particles

Previous D-T experiments

ITER

First D-T experiments at low P/Ptot have demonstrated -heating

• ‚classical‘ (=collisiional) slowing down would guarantee efficient -heating

• question: can we expect this also when P is the dominant heating?

JET, P. Thomas et al., Phys. Rev. Lett. 1998

Excitation of Alfven waves by Fast Particles

Suprathermal ions with can excite Alfven waves which expel them

• in present day experiments, these ions come from heating systems

• in future reactors, this could expel -particles that should heat the plasma!

Magnetic perturbation Fast ion loss probe

Ideal instabilities lead to fast large scale deformation of plasma - disruption

• ultimate stability limit, usually around N 4

Active control possible: nearby conducting structures + internal coils

• may help to extend N above the ideal ‘no-wall’ limit

Stability: ideal pressure limit

N=/(I/aB)=3.5

[%]

Wall erosion strongly depends on edge Te

Acceptable erosion rates only if edge plasma Te is in the 10 eV range

• plasma in front of wall has to be 1000 x colder than core plasma (!)

From Limiters to Divertors

• plasma wall interaction in well defined zone further away from core plasma

• possibility to decrease T, increase n along field lines (p=const.)

Additional cooling by impurity seeding

Injecting adequate impurities can significantly reduce divertor heat load

• impurity species has to be ‘tailored’ according to edge temperature

• edge radiation beneficial, but core radiation (and dilution) must be avoided

No impurityseeding

With N2

seeding

Bolometry of total radiated power Discharge with P/R = 13 MW/m (ASDEX Upgrade)

19

Steep edge pressure gradient in H-mode drives periodic relaxation instability

• Edge Localised Modes (ELMs) lead to burst-like energy pulses on first wall

• simple extrapolation indicates that ELMs are not acceptable in ITER

Thermography of divertor target plates (ASDEX Upgrade)

Edge Localised Modes (ELMs) in the H-mode edge

ELM mitigation needed for ITER

Several techniques have been developed to tailor ELMs

• injection of frozen hydrogen pellets increases repetition frequency

• application of helical fields supresses ELMs completely

Have to understand physics better to extrapolate to ITER

DIII-D Tokamak, USA, Helical perturbation coils (ASDEX Upgrade)

• main topics in fusion plasma physics

• requirements for ITER and DEMO

• present status of physics research

• summary and outlook

Summary: what is required for ITER / DEMO

ITER (Q=10) DEMO achieved

H 1-1.2 1.2-1.4 1.5

N 2 4-5 3-4

Q 10 50 0.6

P/R 20 65 15

Main ITER Q=10 requirements demonstrated today (exception: -heating)

An attractive DEMO will need substantial progress in plasma physics:

• higher to increase fusion power and approach long pulse/steady state

• exhaust of power will be a central point for the success of DEMO

Note: another important area (limitation of plasma density) not covered here