free boundary simulations of the iter baseline scenario and its variants

F. Koechl (1) IOS ITPA Meeting, Kyoto 19.10.2011

Free boundary simulationsof the ITER baseline scenario

and 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


Objectives / Motivation:

Integrated simulations with the free boundary equilibrium code CREATE-NLand the JET suite of codes JINTRAC of the 15 MA ELMy H-mode scenarioin 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


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.


JINTRAC

• Weakly coupled mode: data exchange after each simulation run, iterative consistency check

• Strongly coupled mode: data exchange at every time step


• 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 c

– Lower c 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 ( heating), NGPS / HPI2 (pellets), FRANTIC (neutrals), …

Transport / source models


Discharge configuration / shape evolution:

ITER Baseline Scenario

Central inboard breakdown with fast expansion to diverted shape

Flat-topRamp-down

Ramp-up


Plasma performance / flux consumption:

ITER Baseline Scenario (2)


PF coil voltages during current ramp-up in limiter phase:

Current ramp-up

Voltage saturationHigh load on PF coil converters to achieve fast plasma expansion


Early L-H transition:

Current ramp-up (2)

L-H @ 80s (15 MA)L-H @ 48s (10 MA)L-H @ 30s (7 MA)

same Pfus when steady-state jz is reached

sharp drop in li(3), because dIpl/dt 0

Confinement deterioration due to enhanced transport at lower s/q


Quasi-steady state profiles:

Flat-top

t = 400s


Comparison flat / prescribed vs. peaked / simulated density:

Flat-top (2)

ne ax/avg/ped

Te ax/avg/ped

Ti ax/avg/ped

ne

Te

Ti

t = 200s


Density / pedestal sensitivity scan:

Flat-top (3)

-- Simulation··· Qfus pped

2.0

··· Qfus pped1.3

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


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

PL-H Martin

PL-H Green

Delayed transition from L-mode to type-I ELMy H-mode with increased flux losses

V. Parail, IOS-JA2

Consideration of type-III ELMs:


Flat-top (5)

dashed: PAUX = 40 MWsolid: PAUX = 0 MW

Pfus ~ 50 MW

Profile stiffness allows arbitrary increase in Qfus (provided that Pnet > PL-H)!

Scan in heating power:


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


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


Current ramp-down

li(3) / Vs consumption at ramp-down (L-mode only):

Solid: Vstot Dotted: Vsind

Dashed: Vsres Dash-dotted: Vssawt.

|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

Strong increase in li(3) for high ramp rates affecting vertical stability control


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) / P (dash-dotted) / PAUX (dotted) / PL-H threshold (dashed):

H-L transition @ 15 MAH-L transition @ 7 MA

Drastic reduction in flux losses with prolonged H-mode phasePnet > PL-H achievable with

reduced P + PAUX


Ramp-down at constant loop voltage:


prescr. IplVloop = -0.2VVloop = 0 VVloop = 0.6 V

li(3)

Ipl

Wth

Vloop Fast reduction in Ipl / inductive flux after the transition helps to increase headroom for PF coil control

li(3) kept at lower level in later phase due to decrease in |dIpl/dt|



Non accessible region

Compatibility of flux at ramp-down with PF coil system:

450s flat-top scenarios:

ITER baselineFast ramp-up/downLate H-L transitionLate transition to

diverted phase at ramp-up

pol = 0.6 limits

pol = 0.1 limits

critical flux level reached at ramp-down for baseline scenario


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


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.


Complementary slides


Current ramp-up scan:

Current ramp-up


Current ramp-up (3)Late transition between limiter / divertor configurationPsep in dependence of PAUX level in limited phase:

ITER baselineLate lim./div. transition,Paux: Ipl>7MALate lim./div. transition,Paux: Ipl>5.4MALate lim./div. transition,Paux: Ipl>4MAslowed down plasma expansion

Maximum allowed PAUX of 3 – 5 MW in limiter phase!


Flat-top (6)

Vertical force on P5 coil and CS separation force close to the limit

Forces on PF coils:


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 MAL-H @ 10 MAL-H @ 15 MA


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

free boundary simulations of the iter baseline scenario and its variants

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