Free boundary simulations of the ITER baseline scenario and its variants

1. Free boundary simulationsof the ITER baseline scenarioand its variants F. Koechl, M. Mattei, V. Parail, R. Ambrosino, M. Cavinato, G. Corrigan, L. Garzotti,C. Labate, D. C. McDonald, G. Saibene, R. Sartori

2. Objectives / Motivation: Integrated simulations with the free boundary equilibrium code CREATE-NL and the JET suite of codes JINTRAC of the 15 MA ELMy H-mode scenario in ITER and its variants have been done for the following purposes: • exploration of the operational space and possibilities of scenario optimisation • assessment of the compatibility with machine constraints, in particular with the poloidal field (PF) coil system • evaluation of plasma/control system reaction to transient events and associated risks

3. CREATE-NL FBE solver • Axial-symmetric free boundary code based on numerical solution of Grad-Shafranov equation. • Determination of poloidal flux map by FEM discretisation and Newton based iterative method. • Control of plasma shape (6 gaps) based on a feedforward+feedback control strategy (PF coil nominal current waveform calculated a-priori). Vertical stabilisation using in vessel coils (VS3). • Includes eddy currents, limits in currents and voltages, calculation of forces and max. field in coils.

4. JINTRAC • Weakly coupled mode: data exchange after each simulation run, iterative consistency check • Strongly coupled mode: data exchange at every time step

5. L-mode: Bohm/gyroBohm H-mode, plasma core: GLF23 Bohm/gyroBohm with “GLF23-like” pinch for fast transients Kadomtsev model for sawtooth emulation H-mode, ETB: Continuous ELM model with prescribed ac Lower ac prescription for type-III ELMy H-mode emulation L-H transition model: L-H transition for Pnet > PL-H Martin Transition from type-III to type-I ELMs for Pnet > 1.4·PL-H Martin Source models: PENCIL (NBI), PION / TOMCAT / CYRANO (ICRH), SIMOD (a heating), NGPS / HPI2 (pellets), FRANTIC (neutrals), … Transport / source models

6. Discharge configuration / shape evolution: ITER Baseline Scenario Central inboard breakdown with fast expansion to diverted shape Flat-top Ramp-down Ramp-up

7. Plasma performance / flux consumption: ITER Baseline Scenario (2)

8. PF coil voltages during current ramp-up in limiter phase: Current ramp-up High load on PF coil converters to achieve fast plasma expansion Voltage saturation

9. Early L-H transition: Current ramp-up (2) same Pfus when steady-state jz is reached sharp drop in li(3), because dIpl/dt  0 L-H @ 80s (15 MA) L-H @ 48s (10 MA) L-H @ 30s (7 MA) Confinement deterioration due to enhanced transport at lower s/q

10. Quasi-steady state profiles: Flat-top t = 400s

11. Comparison flat / prescribed vs. peaked / simulated density: Flat-top (2) ne ne ax/avg/ped Te ax/avg/ped Te Ti ax/avg/ped Ti t = 200s

12. Density / pedestal sensitivity scan: Flat-top (3) Non-quadratic Pfus increase with pedestal pressure because of rise in bootstrap current causing lower s/q Constant Pfus for higher ne because of pressure gradient maintenance, but higher flux losses -- Simulation ··· Qfus  pped2.0 ···Qfus  pped1.3

13. Consideration of type-III ELMs: Flat-top (4) Qfus ~ 10 cannot be reached due to back-transition to type-III ELMy H-mode and L-mode if PAUX is reduced Delayed transition from L-mode to type-I ELMy H-mode with increased flux losses PL-H Martin PL-H Green  V. Parail, IOS-JA2

14. Profile stiffness allows arbitrary increase in Qfus (provided that Pnet > PL-H)! Flat-top (5) Scan in heating power: DPfus ~ 50 MW dashed: PAUX = 40 MW solid: PAUX = 0 MW

15. Scan in possible abruptness of pedestal decay after H-L transition (depending on boundary conditions): H-L transition Inner plasma-wall gap safety margins can be temporarily violated for most extreme conceivable transition cases

16. PF coil currents / voltages during H-L transition H-L transition (2) PF6 voltage saturation leading to increasing loss of strike point control Small margins left for required decrease in CS1 current, can only partly be compensated by different CS coil currents

17. Current ramp-down li(3) / Vs consumption at ramp-down (L-mode only): Solid: Vstot Dotted: Vsind Dashed: Vsres Dash-dotted: Vssawt. Strong increase in li(3) for high ramp rates affecting vertical stability control |dIpl/dt| ~ 250 kA/s, 60 s ramp-down |dIpl/dt| ~ 75 kA/s, 200 s ramp-down |dIpl/dt| ~ 38 kA/s, 400 s ramp-down

18. Current ramp-down (2) Comparison early vs. late H-L transition: Vs consumption (total: solid, inductive: dotted, sawtooth-induced: dash-dotted, resistive: dashed) Psep (solid) / Pa (dash-dotted) / PAUX (dotted) / PL-H threshold (dashed): Drastic reduction in flux losses with prolonged H-mode phase Pnet > PL-H achievable with reduced Pa + PAUX H-L transition @ 15 MA H-L transition @ 7 MA

19. Ramp-down at constant loop voltage: Current ramp-down (3) Vloop Fast reduction in Ipl / inductive flux after the transition helps to increase headroom for PF coil control Ipl li(3) kept at lower level in later phase due to decrease in |dIpl/dt| li(3) prescr. Ipl Vloop = -0.2V Vloop = 0 V Vloop = 0.6 V Wth

20. Compatibility of flux at ramp-down with PF coil system: Current ramp-down (4) Non accessible region bpol = 0.1 limits bpol = 0.6 limits critical flux level reached at ramp-down for baseline scenario 450s flat-top scenarios: ITER baseline Fast ramp-up/down Late H-L transition Late transition to diverted phase at ramp-up

21. Flat-top duration limited to 10-20s (up to ~50s with increased current ramp rate), as plasma stays in L-mode: ITER 15 MA hydrogen scenario Critical flux level is already reached because of increased Vsind (L-mode jz) and resistive flux losses (low Te) and strong sawtooth activity

22. General results • According to simulation results, a slow ramp-up phase with late L-H transition gives the optimum fusion performance, whereas a fast ramp-up with early transition to high confinement is preferable in order to save Vs and extend the flat-top duration. • Gap safety margins can be reached for H-L transition at 15 MA. • Trade-off between accumulation of resistive Vs losses for small dIpl/dt and limitations of the PF coil and fuelling systems for high dIpl/dt for current ramp-down (optimum ramp-down period: ~200-250s). Late transition to L-mode during current ramp-down is feasible and advantageous. Constant loop voltage ramp-down is preferable. • An optimisation of the ITER baseline scenario needs to be focused on the reduction of flux consumption to increase flux margins for the PF coil control system and / or increase the flat-top length, trying to avoid at the same time a drop in fusion power in the initial flat-top phase which could occur as a consequence of slowed down current penetration at ramp-up and which could increase the risk to remain in type-III ELMy H-mode conditions at flat-top.

23. Complementary slides

24. Current ramp-up scan: Current ramp-up

25. Late transition between limiter / divertor configuration Psep in dependence of PAUX level in limited phase: Current ramp-up (3) Maximum allowed PAUX of 3 – 5 MW in limiter phase! ITER baseline Late lim./div. transition, Paux: Ipl>7MA Late lim./div. transition, Paux: Ipl>5.4MA Late lim./div. transition, Paux: Ipl>4MA slowed down plasma expansion

26. Vertical force on P5 coil and CS separation force close to the limit Flat-top (6) Forces on PF coils:

27. Transport dependence on s/q Comparison of R/LTi predicted by GLF23 (solid) and by experimentally validated formula with s/q dependence (dotted) for t = 400 s: L-H @ 7 MA L-H @ 10 MA L-H @ 15 MA

28. Long flat-top duration feasible due to small Vsind, plasma can access type-III/I Hmode due to low PL-H: ITER 2.65 T / 7.5 MA H / He4 scenarios