the physics base for iter and demo hartmut zohm max-planck-institut für plasmaphysik, garching,...
TRANSCRIPT
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