Transcript Kein Folientitel - uni

The Physics Base for ITER and DEMO
Hartmut Zohm
Max-Planck-Institut für Plasmaphysik, Garching, Germany
EURATOM Association
• main topics in fusion plasma physics
• requirements for ITER and DEMO
• present status of physics research
• summary and outlook
Hauptvortrag given at AKE DPG Spring Meeting, Bonn, 15.03.2010
Fusion Reactor in a Nutshell
Core plasma @ T=25 keV,
n=1020 m-3 produces Pfus:
D+T = He + n + 17.6 MeV
Plasma physics – this talk
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
Main Areas of Fusion Plasma Physics
Transport determines amount of heating needed to obtain required T
tE = Wkin/Ploss (Ploss is the power needed to sustain the plasma)
experiments measured relative to multi-machine scaling: H=tE,exp/tE,scal
Stability determines the limits to kinetic pressure (Pfus ~ n2T2 = p2)
b = pkin/pmag = 2m0 pkin / B2 (dimensionless pressure)
experimental progress measured relative to ideal MHD limit bN=b/(I/(aB))
a-heating should largely compensate Ploss in a reactor
Q=Pfus/Pext, since Pa = Pfus/5, the fraction of a-heating is Pa/Ploss=Q/(Q+5)
Exhaust characterised by the ratio of power in charged particles to the
major radius, P/R (since the power deposition width is roughly constant)
• main topics in fusion plasma physics
• requirements for ITER and DEMO
• present status of physics research
• summary and outlook
Q
5
c2
q 3.1 A3.53
5
1
c1 Hb N 0.1R 2.7 B 3.7
ITER (Q=10)
DEMO (ignited)
Major radius R0 [m]
Fusion Power [MW]
H and bN determine machine size
Pfus  c1
b N 2B4R3
2
q95 A4
ITER (bN=1.8)
DEMO (bN=3)
Major radius R0 [m]
•bN does almost not enter into Q, but strongly into fusion power
• high H helps to achieve ignition, but does not enter in fusion power.
DEMO should have reasonable pulse length
Tokamak
(ASDEX Upgrade, JET, ITER)
Stellarator
(Wendelstein 7-X)
• 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)
bN=3
bN=4
fCD=0.3
fCD=0.2
fCD=0
fCD=0.1
Pulse length [s]
Net el. power [MW]
Recirculating power fraction
Noninductive current drive in a tokamak DEMO
fCD=0.0
fCD=0.1
fCD=0.2
fCD=0.3
Fusion power [MW]
Intrinsic thermoelectric current (‚bootstrap current‘) – needs high b
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
Summary: what is required for ITER / DEMO
ITER
DEMO
H
1-1.2
1.2-1.4
bN
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
Simplest ansatz for heat transport:
• Diffusion due to collisions
c  rL2 / tc  0.005 m2/s
tE  a2/(4 c)
• table top device (a  0.2 m, R  0.6 m)
should ignite!
Experimental result:
• Anomalous transport by turbulence:
c, D  a few m2/s
• Tokamaks: Ignition expected for
R = 7.5 m for H~1




collision
Transport to the edge
The H-mode: a transport barrier in the edge
discharges with
turbulence
Suppression
H-mode edge: turbulence
suppressed 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 = optimised
H-mode scenario (H = 1.2-1.5)
ITB (Internal Transport Barrier)
scenario (H  1.5)
• potential for very long pulses
(‘hybrid scenario’)
• potential for steady state
(‘advanced tokamak scenario’)
The next step: studying a-heating
Core plasma parameters sufficient to generate significant fusion power
• study plasmas with significant self-heating by a-particles in ITER
• needs Pa = 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 a-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 a-particles
Previous D-T experiments
JET, P. Thomas et al., Phys. Rev. Lett. 1998
ITER
First D-T experiments at low Pa/Ptot have demonstrated a-heating
• ‚classical‘ (=collisiional) slowing down would guarantee efficient a-heating
• question: can we expect this also when Pa is the dominant heating?
Excitation of Alfven waves by Fast Particles
Magnetic perturbation
Fast ion loss probe
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 a-particles that should heat the plasma!
Stability: ideal pressure limit
bN=b/(I/aB)=3.5
b
[%]
Ideal instabilities lead to fast large scale deformation of plasma - disruption
• ultimate stability limit, usually around bN  4
Active control possible: nearby conducting structures + internal coils
• may help to extend bN above the ideal ‘no-wall’ limit
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
Bolometry of total radiated power
Discharge with P/R = 13 MW/m (ASDEX Upgrade)
19
No impurity
seeding
With N2
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
Edge Localised Modes (ELMs) in the H-mode edge
Thermography of divertor target plates (ASDEX Upgrade)
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
ELM mitigation needed for ITER
DIII-D Tokamak, USA,
Helical perturbation coils (ASDEX Upgrade)
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
• 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
bN
2
4-5
3-4
Q
10
50
0.6
P/R
20
65
15
Main ITER Q=10 requirements demonstrated today (exception: a-heating)
An attractive DEMO will need substantial progress in plasma physics:
• higher b 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