ARIES-Advanced TokamakPower Plant Study

Physics Analysis and Issues

Charles Kessel, for the ARIES Physics Team

Princeton Plasma Physics Laboratory

U.S.-Japan Workshop on Fusion Power Plant

Design, University of Tokyo

March 29-31, 2001

ARIES-AT Physics Team

• UCSD– T. K. Mau

– R. Miller

– F. Najmabadi

• PPPL– S. Jardin

– C. Kessel

• GA– V. Chan

– M. Chu

– R. La Haye

– L. Lao

– T. Petrie

– G. Staebler

Objectives for ARIES-AT Physics

• Extend the physics basis for ARIES-RS to increase and decrease PCD

– Use 99% flux surface from free-boundary equilibrium

– Use more flexible plasma pressure profile description

– Increase plasma triangularity– Increase plasma elongation– Eliminate HHFW current drive

Objectives for ARIES-AT Physics

• Include more detailed analysis of critical physics issues– Kink stabilization, RWM analysis with rotation and

feedback control– NTM assessment– Comparison of kinetic profiles with experiments

and Gyro-Landau-Fluid Model– Edge plasma/SOL/Divertor modelling– Plasma edge assumption, L-mode and H-mode

ARIES-RS and AT ParametersARIES-RS ARIES-AT

Ip(MA) 11.3 12.8BT(T) 7.98 5.86R(m) 5.52 5.20a(m) 1.38 1.30κ# 1.70 2.15δ# 0.50 0.78κ(Xp )t 1.90 2.20δ(Xp )t 0.70 0.90P 2.29 1.98(%) 4.98 9.15*(%) 6.18 11.0N(%) (max) 4.84 (5.35) 5.40 (6.00)qaxis 2.80 3.50qmin 2.45 2.40qedge # 3.52 3.70Ibs(MA) 10.0 11.4Iself/Ip 0.91 0.91ICD(MA) 1.15 1.25qcyl 2.37 1.85li(3) 0.42 0.29n(0)/<n> 1.36 1.34T(0)/<T> 1.98 1.72p(0)/<p> 2.20 1.93(b/ )a kink 0.25 0.33# val ue corresponds to fixe -d boundar yequilibrium

ARIES-AT Equilibrium

Use 99% flux surface from free-boundary equilibrium in fixed boundary

equilibrium and stability analysisOld method used 95% surface and analytic plasma boundary

99% plasma surface has stronger shaping

99% plasma surface contains higher order shaping from realistic PF coils

99% plasma volume is consistent with free-boundary plasma

Expand the plasma pressure profile description

More flexible pressure profile allows better bootstrap alignment and higher ballooning limits simultaneously

The HHFW current drive could then be eliminated

p(ψ ) =p(0)[c1(1−ψ b1 )a1 +c2(1−ψb2 )a2 ]ARIES-AT

p(ψ ) =p(0)(1−ψ α)2ARIES-RS

Minimum CD power does not occur at the highest

Reduction of dp/d near the plasma edge for ballooning stability reduced bootstrap current there

Higher N leads to excessive bootstrap current near plasma center requiring broader density profiles

These lead to larger externally driven current at highest ’s

Increasing the plasma triangularity leads to higher

Higher δ increases the N for ballooning modes, and increases the plasma current for fixed q95, while higher κ enhances the effect N (Ip/aB)

Increasing the plasma elongation leads to higher

Higher κ leads to higher N for ballooning modes when κ is less than 2.2-2.5 so long at δ is high enough, however continues to rise due to the strong increase in Ip

The plasma elongation is limited by the vertical instability

Vertical stability analysis showed that we could have κX up to 2.2-2.3 with tungsten conductors 0.3-0.35a from the plasma boundary which is inside the blanket

ARIES-AT vertical stability and control solution

Partial tungsten shells on the inboard and outboard sides, 4 cm thick, at 1100 degC. Feedback control coils are located behind the sheild. For 30 MVA limit we produce a large range of plasmas

External kink stability is provided by a conducting shell and feedback

coilsFor κ>2.3 the conducting shell must move much closer to the plasma, and higher N leads to a closer shell and higher limiting n (toroidal mode number)

The location of the conducting wall is affected by triangularity

Higher δ has allowed us to place the conducting shell further from the plasma, at the same location as the vertical stabilizer

From theory rotation of the plasma with a conducting shell might

stabilize the external kink modeARIES-AT requires 9% of the Alfven speed, which is difficult to provide, and experiments indicate it may not work

ARIES-AT uses feedback to stabilize the external kink modes

• Modelled after DIII-D C-coil approach

• Tungsten shells on outboard side slow the kink mode growth rate creating a RWM

• Feedback coils are 8 (or 16) window coils on the outboard side

• Feedback coils span 60 deg poloidally about the midplane and are outside the shield

• Estimated power requirements are about 10 MW

• Only n=1 has been addressed

External current drive is required on axis and near the plasma edge

ICRF/FW on-axis CD; 96 MHz, N||=2, PCD=5 MW, =0.032 A/W

LHCD off-axis CD; 3.6 GHz , N||=1.65-3.5, PCD=24.5 MW, =0.024-0.053 A/W; 2.5 GHz, N||=5.0, PCD=12.5 MW, =0.013 A/W

Scaling of the CD power with temperature and Zeff is used to find

the best operating point

CD efficiency includes bootstrap current effect

Higher Zeff reduces divertor heat load

Higher temperature improves CD efficiency and leads to thermally stable operating points

N|| were varied for this scan

Alternate CD sources are examined for plasma rotation and

current profile control120 keV NBI provides plasma rotation and current for >0.6, P(NBI)=44 MW for 1.15 MA

HHFW at 20ci provides current at =0.7-0.9 which is useful for control of qmin and r/a(qmin), P(HHFW)=20 MW for 0.4 MA

Neoclassical tearing modes must be stabilized to reach ideal MHD

limitsA combination of the LHCD current profile modification and ECCD at the 5/2 is expected to stabilize the NTM (plot below is H-mode edge), while 3/1 is in very high collisionality region

Examination of assumed T and n profiles and those theoretically

predicted by GLF23

Density is input

Temperature is determined, found to be close to assumed profile

GLF23 is theory based predictive model including ITG, TEM, and ETG turbulence

Comparison of L-mode and H-mode plasma edges

H-mode edge has higher T, giving more j(bootstrap), 30% lower ICD

H-mode edge is unstable to 10 < n < 25 kink/ballooning/peeling modes leading to ELM’s

H-mode has higher ion T at plasma separatrix which aggravates divertor heating

H-mode has significant j(bootstrap) at 5/2 and 3/1 surfaces leading to NTM’s

H-mode has higher edge density leading to larger radiated power fraction from plasma

Plasma edge / SOL / Divertor solution must satisfy physics and

engineering constraints • P(plasma) = 388 MW• SOL width 0.8-2.1 cm

thick• 8:1 OB/IB power split• Q(first wall) < 0.45

MW/m2• Q(div) < 5-6 MW/m2

Divertor heat loads are too high with no techniques for enhanced radiation

For radiated power fraction of 0.3 from plasma core (bremsstrahlung + cyclotron + line from 0.0018 Ar) :

Q(first wall) < 0.4 MW/m2, Q(OB div) < 14 MW/m2, Q(IB div) < 3.5 MW/m2

Radiating plasma power is required to obtain workable divertor solutions

Radiating mantle, Frad=0.75, Q(fw)>0.9 MW/m2, Q(OB div)>5 MW/m2, Q(IB div)>1.2 MW/m2 --> 0.0035 Ar (core)

Radiating divertor, Frad=0.36, Frad(div)=0.43, Q(fw)<0.45 MW/m2, Q(OB div)<6 MW/m2, Q(IB div)<1.5 MW/m2--> 0.0026 Ar (divertor)

POPCON plot for ARIES-AT

P(fus) = 1760 MW P(alpha) = 351 MW P(brem) = 55 MW P(cyc) = 18 MW P(line) = 67 MW P(aux) = 37 MW H(89p) = 2.65 H(98y) = 1.35 E = 2.0 s p*/E = 10.0 n/nGr = 1.0 Zeff = 1.85

PF coil locations and currents are optimized with maintenance

constraints

ARIES-AT Physics Issues and New Experimental Results

• Neutral particle control can allow the plasma density to exceed the Greenwald limit without confinement degradation (DIII-D, TEXTOR).

• Helium particle control is demonstrated with pumped divertors giving p*/ 315 (DIII-D, JT-60)

• Detachment of inboard strike point plasma allows high triangularity (DIII-D).

• LHCD is shown to stabilize neoclassical tearing modes (COMPASS, ECCD on ASDEX-U,DIII-D).

• Vertical and inboard pellet launch show better penetration with lower speed (ASDEX-U, DIII-D).

ARIES-AT Physics Issues and Experimental Results

• RWM n=1 kink feedback control experiments continue on DIII-D (if successful n=2 will set the -limit)

• Controlled impurity effects on the plasma edge, SOL, and divertor continue on many tokamaks

• Experiments on the behavior of Internal Transport Barriers (ITB) through control of turbulence will lead to control of T and n profiles (JET, JT-60U, DIII-D, ASDEX-U)

A High , High fBS Configuration Have Been Developed as the Physics Basis for Advanced

Tokamak Fusion Power Plants

• High accuracy equilibria

• Large ideal MHD database over profiles, shape and aspect ratio

• RWM stable with wall/rotation or wall/feedback control

• NTM stable with L-mode edge and LHCD

• Bootstrap current consistency using advanced bootstrap models

• External current drive

• Vertically stable and controllable with modest power (reactive)

• Modest core radiation with radiative SOL/divertor

• Accessible fueling

• No ripple losses

• 0-D consistent startup

• Rough kinetic profile consistency with RS /ITB experiments, examining GLF23 model consistency

• Several assumptions based on experimental/theoretical results