# Presentation: 114747

## ARIES-Advanced TokamakPower 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 Parameters

## ARIES-AT Equilibrium

## Use 99% flux surface from free-boundary equilibrium in fixed boundary equilibrium and stability analysis

- Old 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

- ARIES-AT

- ARIES-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  increases the N for ballooning modes, and increases the plasma current for fixed q95, while higher  enhances the effect           N (Ip/aB)

## Increasing the plasma elongation leads to higher  

- Higher  leads to higher N for ballooning modes when  is less than 2.2-2.5 so long at  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 X 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 coils

- For >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  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 mode

- ARIES-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_x0008_

- 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 control

- 120 keV NBI provides plasma rotation and current for >0.6, P(NBI)=44 MW for 1.15 MA
- HHFW at 20ci 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  limits

- A 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*/(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