1. 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

2. Fusion Reactor in a Nutshell 4/5*P fus escape as neutrons and hit the first wall (Blanket = tritium production and energy conversion) Neutronics – talk by A. Klix 1/5*P fus + P ext 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=10 20 m -3 produces P fus : D+T = He + n + 17.6 MeV Plasma physics – this talk

3. Transport determines amount of heating needed to obtain required T  E = W kin /P loss (P loss 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 (P fus ~ n 2 T 2 = p 2 )  = p kin /p mag = 2  0 p kin / B 2 (dimensionless pressure) experimental progress measured relative to ideal MHD limit  N =  /(I/(aB))  -heating should largely compensate P loss in a reactor Q=P fus /P ext, since P  = P fus /5, the fraction of  -heating is P  /P loss =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 Areas of Fusion Plasma Physics

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

5. H and  N determine machine size ITER (Q=10) DEMO (ignited)  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. ITER (  N =1.8) DEMO (  N =3) Major radius R 0 [m] Fusion Power [MW]

6. 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)

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

8. Summary: what is required for ITER / DEMO ITERDEMO H1-1.21.2-1.4 NN 24-5 Q1050 P/R2065 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. Confinement of plasma core - transport Experimental result: Anomalous transport by turbulence: , D  a few m 2 /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   r L 2 /  c  0.005 m 2 /s  E  a 2 /(4   table top device (a  0.2 m, R  0.6 m) should ignite!

11. discharges with turbulence Suppression The H-mode: a transport barrier in the edge 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  -heating Core plasma parameters sufficient to generate significant fusion power study plasmas with significant self-heating by  -particles in ITER needs P  = 1/5 P fus >> P ext, 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

14. Previous D-T experiments ITER First D-T experiments at low P  /P tot 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

15. 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

16. 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  [%]

17. Wall erosion strongly depends on edge T e Acceptable erosion rates only if edge plasma T e 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 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 impurity seeding With N 2 seeding Bolometry of total radiated power Discharge with P/R = 13 MW/m (ASDEX Upgrade) 19

20. 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

21. 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)

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 ITER (Q=10)DEMOachieved H1-1.21.2-1.4  1.5 NN 24-53-4 Q10500.6 P/R206515 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