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NCSX & Compact Stellarators

in the

World Fusion Program

M.C. Zarnstorff

PPPL

Briefing for Prof. R. Fonck

Associate Director, Office of Fusion Energy Science

1 May 2007

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Outline

• What is needed for DEMO?

• Stellarator solutions

• Role & goals of NCSX

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What must the World Program developfor DEMO?

This requires development of plasmas with:• Higher pressure, by at least factor of 1.7• Less external current drive, more bootstrap current; no inductive current• No external rotation drive• Essentially no disruptions

• And: high heat flux PFCs, T-breeding cycle, long-lived materials.

For tokamaks, many of these needed improvements conflict with each other.

ITER (~ 2016)

ITER: 500 MW for 10 minutes, gain > 10

DEMO: ~2500 MW, continuous, gain > 25, ~ same size and field.

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Tokamak Advanced Scenarios for Steady-State DEMOAre Very Challenging

• Most reactor studies aim for fbs ~ 80-99%

requires feedback controlled stability with > no-wall

and NTM stabilization

• Results in non-linear coupling between -heating profile, current profile, transport, and MHD stability

requires feedback controlled pressure/current profile

• Only feasible solution for divertor power handling appears to be radiating mantle, with Prad ~ 80-90% of loss power impurity seeding

• High density operation needed for radiating divertor solution.– EU designs all have n > nGreenwald!! Yet need HH > 1.

– Reduces current drive efficiency.See D. Campbell et al.IAEA 2000, Chengdu“Critical Issues for Tokamak Power Plants”

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“Disruptions Will Essentially Have to be Eliminated”

D. Campbell et al.IAEA 2006, Chendu“Critical Issues for Tokamak Power Plants”

Aries studies: disruptions must be < 1 per year

• Breeding blanket designs have first-wall thickness of 1.5 – 5 mm insufficient thermal mass to avoid melting, doesn’t stop runaways.• How to guarantee no disruptions at high beta in bootstrap sustained configurations?

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Compact Stellarators provide Solutions A steady-state toroidal reactor with

Quiescent steady state at high-beta No current drive intrinsically high Q, can operate ignited Not limited by macroscopic instabilities. No need to control profiles. No need

for feedback or rotation to control instabilities, or nearby conducting structure No disruptions allows thin PFCs, allowing breeding blanket Very high density limit easier plasma solutions for divertor

reduced fast-ion instability drive• High power density (similar to ARIES-RS and –AT)• Good alpha-particle confinement

= Already demonstrated in high-aspect ratio, non-symmetric stellarators

Most of these are available in all stellarators (except high beta)

• Greatly simplifies many aspects of a projected DEMO.• Need to demonstrate these properties at low aspect ratio, to project

to high power density.

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Stellarator Operating Range is Stellarator Operating Range is much larger than for Tokamaks much larger than for Tokamaks

Using equivalent toroidal current that produces same edge iota in Greenwald evaluation.

Limits are not due to MHD instabilities. No disruptions.

High density favorable:– Lower plasma edge temperature, Eases edge design– Reduces energetic particle instability drive

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NCSX Strategy: Combine Advantages of Stellarators and Tokamaks

Tokamaks:• Compact cost-effective, project to high power density • Importance of flows ( including self-generated) for turbulence stabilization• ‘Reversed shear’ to reduce turbulence, increase stability, suppress islands

Stellarators:• Externally-generated helical fields

– Plasma current not required. No current drive. Steady-state easy.– Robust stability. Generally, disruption-free.

• Numerical optimization of 3D field (shape) to obtain desired

physics properties, including – Increased stability, good flux surfaces at high-beta – Quasi-toroidal symmetry: If |B| is symmetric in flux coordinates, orbits are like

tokamak neoclassical transport very similar to tokamaks, undamped rotation

Goal: Steady-state high-, good confinement without disruptions

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NCSX Designed to Integrate Attractive PropertiesAt Low Aspect Ratio & Address Needs

• 3 periods, R/a=4.4, ~1.8 , ~1

• Quasi-axisymmetric: transport similar to tokamaks

ripple thermal transport insignificant.

• Passively stable at =4.1% to kink, ballooning, vertical,

Mercier, neoclassical-tearing modes, …

(steady-state tokamak limit ~ 2.7% without feedback)

• Stable for > 6.5% by adjusting coil currents

• Passive disruption stability: equilibrium maintained even with total loss of or IP

• Flexible configuration: 9 independent coil currents

by adjusting currents can control stability, transport, shape: iota, shear

The US is the World Leader in Compact Stellarator Design

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ARIES-CS: a Competitive, Attractive ReactorReference parameters for baseline:

NCSX-like config.

R = 7.75 m

a = 1.72 m

n = 4.0 x 1020 m–3

T = 6.6 keV

Baxis = 5.7 T

= 6.4%

H(ISS95) = 2.0

H(ISS04) = 1.1

Iplasma = 3.5 MA

(bootstrap)

P(fusion) = 2.364 GW

P(electric) = 1 GWBased on NCSX design

Aries- -I -RS -CS -AT -CS

Blanket LiPb/FS LiPb/SiC LiPb/SiC

COE(92) 99.7 75.8 61.3 47.5 48.

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NCSX Research MissionAcquire the physics data needed to assess the attractiveness of

compact stellarators; advance understanding of 3D fusion science.

Understand…• Pressure limits and limiting mechanisms in a low-A optimized stellarator

• Effect of 3D magnetic fields on disruptions. Operating limits.

• Reduction of and anomalous neoclassical transport by quasi-axisymmetric design.

• Confinement scaling; reduction of turbulent transport by flow shear control.

• Equilibrium islands and tearing-mode stabilization by design of magnetic shear.

• Compatibility between power and particle exhaust methods and good core performance in a compact stellarator.

• Energetic-ion stability and confinement in compact stellarators

Demonstrate…• Conditions for high , disruption-free operation• High pressure, good confinement, compatible with steady state

Can we achieve the promise of Aries-CS?

Can the design be simplified through improved understanding?

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The World’s Stellarator Programs Focused on Superconducting, Steady State

Large Helical Device (Japan)Enhanced confinement, high pressure;

A = 6-7, R=3.9 m, B=3T

Wendelstein 7-X (Germany) (2012)Non-symmetric optimized design:

A = 11, R=5.4 m, B=3T

• Significant technology programs on 3D superconducting coils• LHD has achieved 54-minute pulses.• Large aspect ratio. Neither optimized for symmetry flows strongly damped• Both are very interested in NCSX, collaborating & contributing hardware ( ECH from Germany; HIBP from Japan under discussion )

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Confinement Degrades with Ripple eff

NCSX Has Lowest

• New global confinement scaling study for stellarators (ISS04v3) found strong dependence on ripple magnitude (eff).

• NCSX designed for the lowest ripple of all configurations.• HSX has demonstrated advantages of quasi-symmetry: increased

confinement and decreased flow damping

~eff–0.4?

10-5

0.0001

0.001

0.01

0.1

1

0 0.2 0.4 0.6 0.8 1

eff

3/2

r/a

ATF

NCSX

QPS

W7-AS

LHDshifted in

HSX

1/ transport ~ eff3/2

NCSX

ARIES-CS

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‘Reversed Shear’ Key to Enhanced Stability in NCSX

0.0

0.1

0.2

0.3

0.4

0.5

0.6

0.7

0.0 0.2 0.4 0.6 0.8 1.0

total

externallygenerated

(rota

tional tr

ansfo

rm)

S (normalized tor. flux)

• Quasi-axisymmetry tokamak like bootstrap current

• ~3/4 of transform (poloidal-B) from external coils externally controllable

• Rotational transform rising to edge key for stabilizing trapped particle and neoclassical tearing instabilities

Explored locally on tokamaks, but cannot be achieved across whole plasma using current.

10

5

2

3

Safety fac to )r (q)

Radial Coordinate2

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Turbulence Growth Decreases for Higher p Similar to Reversed Shear Tokamak

Designed for ‘reversed shear’ to help stabilize turbulent transport

Linear ion-scale turbulence growth rates calculated by FULL-code: Electron-drive stabilized by reversed

shear Ion-drive strongly reduced with Similar to reversed shear tokamak

Very low effective helical ripple gives low flow-damping allows efficient flow-shear stabilization, control of Er

.

Allows strong zonal flows (Mynick).

- G. Rewoldt

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LHD and W7AS both achieved high- without being designed for it!

• LHD recently achieved =5% • In both cases, well above theoretical linear stability limit < 2%. Quiescent.• MHD activity not limiting. No disruptions observed. Sustained without CD.

Germany Japan

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S. Sakaibara, Y. SuzukiM.C. Zarnstorff, A. Reiman

NCSX is designed to keep good flux surfaces and linear stability at high

-limit Correlates with Destruction of Outer Flux Surfaces

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NCSX Designed to Produce Good Flux Surfaces at High-

Poincare: PIES, free boundary

without pressure flattening

< 3% flux loss, including effects of reversed shear and || vs. transport.

• Explicit numerical design to eliminate resonant field perturbations• ‘Reversed shear’ configuration pressure-driven plasma currents heal

equilibrium islands (not included in figure)

Computationboundary

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Initial Non-Linear Kinetic CalculationsIndicate Possible Higher Pressure-Limit for NCSX

MagneticFlux Surfaces

ExB FlowSurfaces

NCSX

• Preliminary M3D calculations. Fixed boundary.

• Finite gyro-radius and self-generated flows stabilize equilibrium

• Does not include neoclassical effects yet. Should further increase stabilization.

• What will the pressure limit be for NCSX?

= 7%

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Intrinsic Divertors in Bean-tipsdivertor

vacuum vessel

• Divertors already operated successfully in LHD and W7-AS. Controlled exhaust and impurities.

• Strong flux-expansion always observed in NCSX bean-shaped cross-section. Allows isolation of PFC interaction.

• Similar to expanded boundary shaped-tokamak configurations

• SOL connection length can be ~100m. Long enough to ensure low temperature divertor plasma.

pumps

MFBE field-line tracing

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• Modular Coils + Toroidal Solenoid + Poloidal Coils for shaping control & flexibility

• Useful for testing understanding of 3D effects in theory & determining role of iota-profile

• E.G., can use coils to vary– effective ripple by factor > 10.– Avg. magnetic shear by factor > 5– Edge rotational transform by factor of 2

• Reduce kink-instability threshold down to 1% by modifying plasma shape

– either at fixed shear or fixed edge-iota !

• These types of experiments will be key for

developing and validating our understanding

NCSX Coils Designed for Flexibility

0.0 0.2 0.4 0.6 1.00.8

Rotational Transform Profiles

Ro

tatio

na

l Tra

nsf

orm

(io

ta)

0.0

0.2

0.4

0.6

1.0

0.8

Ele

vatio

n

–1

.0–

0.5

0.0

0.5

1.0

0.5 1.0 1.5 2.0 2.5Major radius

0.5 1.0 1.5 2.0 2.5Major radius

Ele

vatio

n

–1

.0–

0.5

0.0

0.5

1.0

0.5 1.0 1.5 2.0 2.5Major radius

Ele

vatio

n

–1

.0–

0.5

0.0

0.5

1.0

Rel. Toroidal Flux

Shear

Rotational Transform

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Conclusions• Compact stellarators provide solutions for DEMO

– ARIES-CS competitive with best AT designs

– No disruptions, no current-drive, increased pressure, no profile feedback.

– Relatively small step from achieved stellarator characteristics

• NCSX is an exciting opportunity for unique fusion-science research.

– High-, compatible with steady state, good flux surfaces, and stability

– Sustainment without external current drive

– Stability without rotation drive or active feedback on instabilities

– No disruptions

– Tokamak-like transport using quasi-axisymmetry: low rotation damping, good

alpha confinement.

• NCSX and compact stellarators provide an opportunity for US Leadership

and a route to a more practical DEMO.

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Supplemental

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Disruptions are a Big IssueMust design to protect facility & environment in case of disruption.

ITER:• PFCs use 10mm of Be or W, backed by 22mm of Cu

to have sufficient thermal mass to prevent bulk melting-- incompatible with breeding blankets --

• Calculations of disruption PFC erosion give0.25 mm of W for a VDE-down onto baffle0.15 mm of Be for a VDE-up0.10 mm of Be for a major disruption(last weeks WG 1 meeting; Sugihara, NF 47 (2007) 337)

• Electromagnetic loads are <10% from allowable, strongly dependent on width of halo currents

• Even mitigated disruptions are calculated to melt the PFC surfaces, may produce cracking

• Disruptions in advanced scenarios (reversed-shear or above no-wall limit) are often prompt, with shortest quench times. (draft ITER Physics Basis, Ch. 3) -- may not be possible to predict or mitigate --

ITER (~ 2016)

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Wide Range of and * Accessible B = 1.2 T, 3MW

* =2.7%, *I =0.25 with HISS95=2.9; HISS04=1.5

HITER-97P=0.8

* =2.7%, *I =2.5 with

HISS95=2.0; HISS04=1.0

* =1.4%, collisional with HISS95=1.0, ; HISS04=0.5

sufficient to test stability theory

Contours of HISS95, HITER-97P, and min *i

LHD and W7-AS have achieved HISS95 ~ 2.5PBX-M obtained = 6.8% with HITER-97P = 1.7 and HISS95 ~ 3.9

*

*

*

ne (1019 m-3)

<>

(%

)

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High- low * Plasmas Accessible (6MW)

B = 1.2 T, 6MW =4%, *I =0.25 requires

HISS95=2.9, HISS04=1.5

HITER-97P=0.9

=4% at Sudo-density HISS95=1.8, HISS04=0.9

HISS95=1.0 gives =2.2%

at high collisionality

Contours of HISS95, HITER-97P, and min *i

LHD and W7-AS have achieved HISS95 ~ 2.5PBX-M obtained = 6.8% with HITER-97P = 1.7 and HISS95 ~ 3.9

<>

(%

)

ne (1019 m-3)

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How to Achieve Orbit Confinement in 3D??Quasi-symmetry

3D shape of standard stellarators

No symmetry no conserved canonical momenta

orbits can have resonant perturbations, become stochastic lost

B is bumpy every direction rotation is strongly damped

‘Quasi-symmetry’ (Boozer,1983) Orbits & collisional transport depends on variation of |B|

within flux surface, not the vector components of B ! Hamiltonian for particle drift-motion only depends on |B| in flux coordinates (Nührenberg) If |B| is symmetric in flux coordinates, get confined orbits like

tokamak

neoclassical transport very similar to tokamaks (theoretically),

undamped rotation

• Recently tested in HSX, at Univ. Wisconsin– quasi-Helical-Symmetry reduces electron transport, flow damping

– Too small to study high pressure or ion transport

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NCSX Design: Detailed Plasma Modeling & Optimization

Vary Coil Shapes ~ 200–400 free parameters

Coil Characteristics

3D Equilibrium Calc. (VMEC)

Macroscopic StabilityHigh- & low-n

Orbit Confinement

Flux Surface Quality

Ripple Transport

. . .

Adjust Coil Shapes & Currents(Levenberg-Marquardt,

Differential Evolution, Genetic)

Direct design of coil shapes to achieve desired physics properties

Only possible using high-capacity advanced computingand advances in theoretical understanding and modeling

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0

1

2

3

4

0

5

10

15

20

250

1

2

0

1

2

3

4

0.1 0.2 0.3 0.4Time (s)

ne

(102

0 m

-3)

PNB

Prad

Mirnov B.

P

ow

er

(M

W)

(A.U

.) ≈ 3.4 % : Quiescent, Quasi-stationary ≈ 3.4 % : Quiescent, Quasi-stationary

B = 0.9 T, iotavac ≈ 0.5

Almost quiescent high- phase, MHD-activity in early medium- phase

In general, not limited by any detected MHD-activity.

IP = 0, but there can be local currents

Peak ~8%

Similar plasmas with B = 0.9 – 1.1 T, either NBI-alone, or combined NBI + OXB ECH.

Much higher than predicted linear stability limit ~ 2%

54022

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For compact, quasi-symmetric, sustainable high-beta configurations:1. Can beta ~6.4% be achieved and sustained at good confinement? What is the

maximum useful beta? 2. Can low alpha loss be achieved? Can alpha loss due to MHD instabilities be mitigated

by operation at high density?3. Develop a workable divertor design with moderate size and power peaking, that

controls impurities and enables ash pumping.4. Demonstrate regimes of minimal power excursions onto the first wall (e.g. due to

disruptions and ELMs). 5. Under what conditions can acceptable plasma purity and low ash accumulation be

achieved?6. Is the energy confinement at least 2 times ISS95 scaling? How does it extrapolate to

larger size? 7. Characterize other operational limits (density, controllable core radiation fraction)8. How does the density and pressure profile shape depend on configuration and plasma

parameters?9. Can the coil designs be simplified? Can physics requirements be relaxed, by

a. Reduction of external transformb. Elimination of stability from optimizationc. Reducing flux-surface quality requirementsd. Increased helical ripple

10. What plasma control elements and diagnostics are required?

NCSX Research Plan is focused on addressing these questions and developing their scientific basis.

ARIES-CS Physics R&D Needs