The Physics Base for ITER and DEMO

1. 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

2. 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

3. 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)

4. main topics in fusion plasma physics • requirements for ITER and DEMO • present status of physics research • summary and outlook

5. H and bN determine machine size Fusion Power [MW] ITER (bN=1.8) DEMO (bN=3) ITER (Q=10) DEMO (ignited) Major radius R0 [m] Major radius R0 [m] • bNdoes almost not enter into Q, but strongly into fusion power • high H helps to achieve ignition, but does not enter in fusion power.

6. 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)

7. bN=3 fCD=0.0 fCD=0.1 fCD=0.2 fCD=0.3 fCD=0.3 bN=4 fCD=0.2 fCD=0.1 fCD=0 Noninductive current drive in a tokamak DEMO Net el. power [MW] Recirculating power fraction Fusion power [MW] Pulse length [s] • 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

8. Summary: what is required for ITER / DEMO Reality check: how does this compare to present experimental data base?

9. main topics in fusion plasma physics • requirements for ITER and DEMO • present status of physics research • summary and outlook

10.     collision Transport to the edge Confinement of plasma core - transport • Simplest ansatz for heat transport: • Diffusion due to collisions • c rL2 / tc 0.005 m2/s • tE a2/(4c) • 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

11. 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)

12. Scenarios with improved confinement (H>1) • Improved H-mode = optimised • H-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’)

13. 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

14. 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?

15. 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!

16. N=/(I/aB)=3.5  [%] Stability: ideal pressure limit • 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

17. 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 (!)

18. 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.)

19. 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

20. 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

21. 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

22. main topics in fusion plasma physics • requirements for ITER and DEMO • present status of physics research • summary and outlook

23. Summary: what is required for ITER / DEMO • 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