columbia u spherical tokamak plasma...

ST2004-9/29-10/1/04 ST Science & Fusion Energy

Martin PengNSTX Program Director

Oak Ridge National Laboratory@ Princeton Plasma Physics Laboratory

Joint Spherical Torus Workshop andUS-Japan Exchange Meetings (STW2004)

29th September – 1st October, 2004Kyoto University

Yoshida-Honmachi, Kyoto, Japan

Spherical Tokamak Plasma Science & Fusion Energy Development

Supported by

Columbia UComp-X

General AtomicsINEL

Johns Hopkins ULANLLLNL

LodestarMIT

Nova PhotonicsNYU

ORNLPPPL

PSISNL

UC DavisUC Irvine

UCLAUCSD

U MarylandU Rochester

U WashingtonU Wisconsin

Culham Sci CtrHiroshima U

HISTKyushu Tokai U

Niigata UTsukuba U

U TokyoJAERI

Ioffe InstTRINITI

KBSIKAIST

ENEA, FrascatiCEA, Cadarache

IPP, JülichIPP, Garching

U Quebec


Spherical Tokamak (ST) Offers Rich Plasma Science Opportunities and High Fusion Energy Potential

• What is ST and why?• Scientific opportunities of ST

• How does shape (κ, δ, A …) determine pressure? • How does turbulence enhance transport?• How do plasma particles and waves interact?• How do hot plasmas interact with walls?• How to supply magnetic flux without solenoid?

• Contributions to burning plasmas and ITER• Cost-effective steps to fusion energy• Collaboration


Tokamak Theory in Early 1980’s Showed Maximum Stable βT Increased with Lowered Aspect Ratio (A)

• A. Sykes et al. (1983); F. Troyon et al. (1984) on maximum stable toroidal beta βT:

βTmax = C Ip / a ⟨B⟩ ≈ 5 C κ / A qj; ⟨B⟩ ≈ BT at standard AC ≈ constant (~ 3 %m·T/MA) ⇒ βN

⟨B⟩ = volume average B ⇒ BT

κ = b/a = elongationA = R0/a = aspect ratioqI ≈ average safety factor Ip = toroidal plasma currentBT ≈ applied toroidal field at R0

• Peng & Strickler (1986): What would happen to tokamak as A → 1?

− How would βN, κ, qj, change as functions of A?

Z

a ab

0 RR0

PlasmaCross

Section


• Naturally increased κ ~ 2; ITF < Ip, IPF < Ip ⇒ higher Ip; lower device cost• Increased Ip/aBT ~ 7 MA/m·T ⇒ βTmax ~ 20%, if βN ~ 3• Increased Ip qedge /aBT ~ 20 MA/m·T ⇒ improved confinement?

ST Tokamak

ITF / Ip(~aBT / Ip)

ΣIPF / Ip

κ

Natural Elongation, κ Small Coil Currents/Ip (qedge~2.5)

R R A

Z

0

A = 2.5κ ≈ 1.4

A = 1.5κ ≈ 2.0

ST Plasma Elongates Naturally, Needs Less TF & PF Coil Currents, Increases Ip/aBT ⇒ Higher βTmax


Very Low Aspect Ratio (A) Introduces New Opportunities to Broaden Toroidal Plasma Science

ST Plasmas ExtendsST Plasmas ExtendsToroidal ParametersToroidal Parameters

A = R/a can be ≥ 1.1

START – UKAEA Fusion

How does shape determine pressure?• Strong plasma shaping & self fields

(vertical elongation ≤ 3, Bp/Bt ~ 1)• Very high βT (~ 40%), βN & fBootstrapHow does turbulence enhance transport?• Small plasma size relative to gyro-radius

(a/ρi~30–50)• Large plasma flow (MA = Vrotation/VA ≤ 0.3)• Large flow shearing rate (γExB ≤ 106/s)How do plasma particles and waves interact?• Supra-Alfvénic fast ions (Vfast/VA ~ 4–5) • High dielectric constant (ε = ωpe

2/ωce2 ~ 50)

How do plasmas interact with walls?• Large mirror ratio in edge B field (fT → 1)• Strong field line expansionHow to supply mag flux without solenoid?• Small magnetic flux content (~ liR0Ip)


ST Research Is Growing Worldwide

HIST (J)

Globus-M

CDX-U

START TS-3,4TST-MHIST, LATE

Pegasus

ETE

MAST

NSTXHIT-II Proto-Sphera SUNLIST

Rotamak-ST

HIT-I

TST-2 (J) TS-3 (J)TS-4 (J)

Pegasus (US)

CDX-U (US)

NSTX (US)

HIT-II (US)

MAST (UK)

ETE (B)

SUNIST (PRC)

Proof of Principle (~MA)Concept Exploration (~0.3 MA)

Globus-M (RF)


Pegasus Explores ST Regimes As Aspect Ratio → 1


NSTX Exceeded Standard Scaling & Reached Higher Ip/aBT, Indicating Better Field and Size Utilization

• Verified very high beta prediction ⇒ new physics:βT = 2µ0⟨p⟩ / BT0

2 ≤ 38%βN = βT / (Ip/aBT0) ≤ 6.4⟨β⟩ = 2µ0⟨p⟩ / ⟨B2⟩ ≤ 20%

• Obtained nearly sustained plasmas with neutral beam and bootstrap current alone– Basis for neutral beam

sustained ST CTF at Q~2 – Relevant to ITER hybrid mode

optimization

• To produce and study full non-inductive sustained plasmas– Relevant to DEMO

Columbia U, LANL, PPPL

CTF β requirement well within stabilityLimits, without using active control

CTF

DEMOβN = 6


Detailed Measurements of Plasma Profiles Allows Physics Analysis and Interpretations

Plasma Flow ShearingRate up to ~106 /s


Strong Plasma Flow (MA=Vφ/VAlfvén~0.3) Has Large Effects on Equilibrium and Stability

• Internal MHD modes stops growing

• Pressure axis shifts out by ~10% of outer minor radius

• Density axis shifts by ~20%

0.4 0.6 0.8 1.0 1.2 1.4R(m)

0.0

0.2

0.4

0.6

0.8

1.0

1.2

1.4 107540 330ms

MA

Te(keV)

ne (m-3)4x1019

0.5 2.01.0 1.5

-2

-1

1

2

0

R(m)

Z(m

)

pressure surfacesmagnetic surfaces

Equilibrium Reconstruction with Flow

Columbia U, GA, PPPL, U Rochester


High-Resolution CHERS, SXR, and In-Vessel BR and BP Sensors Reveal Strong Mode-Rotation Interaction

CHarge-Exchange Recom-bination Spectroscopy

(CHERS) shows vφ collapsepreceding β collapse

SXR shows rotating 1/1mode during vφ decay

1/1 Island

In-vessel sensors measurerotating mode as vφ decays

before mode locking

Aliased n=1 rotating mode

RWM, NTM, 1/1 modes, and rotation physics of high interest to ITER

Sabbagh, Bell, Menard, Stutman


Active Control Will Enable Study of Wall Mode Interactions with Error Fields & Rotation at High βT

Columbia U, GA, PPPL

Res

istiv

e W

all M

ode

Gro

wth

Rat

e (s

-1)

Internal Sensors

Conducting Plates

Vacuum Vessel


Global and Thermal τE’s Compare Favorably with Higher A Database

• TRANSP analysis for thermal confinement

τ E<t

herm

al>

(ms)

τE<ITER-H98p(y,2)> (ms)

12080400

120

80

40

00.00 0.01 0.02 0.03 0.04 0.05

0.00

0.05

0.10

τ E<m

ag> [

s]

τE<ITER-97L> [s]

H-modeL-mode

x2.5

x1.5

• Compare with ITER scaling for total confinement, including fast ions

L-modes have higher non-thermal component and comparable τE! Why?Bell, Kaye, PPPL


Ion Internal Transport Barrier in Beam-Heated H-Mode Contrasts Improved Electron Confinement in L-Mode

Magnetic Axis

Kinetic Profile Local Error Sampling

Axis Edge

Regionsrequiringimproved

dataresolution

TransportBarrierregionwhere

χi ~ χiNC

andχe >> χi

Columbia U, Culham, ORNL, PPPL

But L-mode plasmas show improved electron confinement! Why?

L-mode H-mode

H-modeL-mode

T e(k

eV)

n e(1

013 /c

m3 )

iITB?

iITB?


Transport Analysis of NSTX Plasmas Using TRANSP Confirms This Contrast

• χe >> χi ~ χNCLASS in most H-mode• χe ~ χi in L-mode• Diagnostic Resolution improvements continue

L-mode


Analysis Shows Stability to Modes at Ion Gyro-Scale & Strong Instability at Electron Gyro-Scale (H-Mode)

Em

issi

vity

(mW

/cm

3 )

Ne 8,9+

MHD event

Ne pufft (ms)

R (cm)• Driven by T and n

gradients• kθρi < 1 (ion gyro-

scale) stable or suppressed by Vφshear

• kθρi >> 1 (electron gyro-scale) strongly unstable

Micro-instability

calculations

• Dimp ~ DneoclassicalImpurity Diffusivity

• χion ~ χneoclassical

• χelec >> χion

Thermal Conductivity

In ion confinement zone

Core Transport Physics

Cadarache, JHU, PPPL, U. Maryland

iITB?


A Broad Spectrum of Energetic Particle Driven Modes is Seen on NSTX

10

100

1000

0.1 0.2 0.3TIME (s)

FRE

QU

EN

CY

(kH

z)

“Fish-Bones”

TAE

CAE/GAE

108170

Do these Alfvén Eigenmodes (AEs) and fish-bones (f.b.s) Interact to expel energetic particles?

PPPL


TAE’s, “Fish-Bones,” and CAE/GAE’s Can Interact to Expel Energetic Particles

Synchronous sudden activities of• Edge Dα rises; D-D neutron drops• Fish-bone modes rises• TAE mode crashes• Separately, asynchronous drops of f.b.

and CAE modes• So far observed for βT ≤ 10% and Ip ≤

700 kA ⇒ high-β effects?• NPA measured depletion for 50-80 kV at

higher βT – MHD (m/n=4/2) induced?• Nonlinear effects relevant to lower β

burning plasmas (ITER)

H-alpha (a.u.)

Plasma Current (MW)

Pnbi (MW)

0.8

0.4

0.0

1.0

0.01.0

0.0

shot 108530

4

00.2 0.30.0 0.1 0.4

TIME (s)0.20 0.25 0.30

Neutrons (1014/s)

~rm

s(B

)

103

102

101

103

102

101

100

102

101

100

10-1CAE 500 - 1500 kHz

TAE 70 - 150 kHz

f.b. 10- 40 kHz

(Ip = 0.65 MA, Pb = 3.6 MW, βT = 10%)

PPPL


NSTX RF Research Explores High Dielectric (ε ~ 100) Effects for Efficient Heating & Current Drive

M. Ono (1995): High Harmonic Fast Wave (HHFW) decay (absorption) rate:

k⊥im ~ ne / B3 ~ ε / B,ε = ωpe

2 / ωce2 ~ 102

Laqua et al (1997): Conversion of oblique O-mode to slow X-mode to Electron Bernstein Wave (EBW):

Electromagnetic

Electrostatic

Over-densePlasma

Launcher/Receiver

&


HHFW: Heat Electrons and Trigger H-Modes, Relevant to Slowly Rotating ITER Plasmas

• Antenna operated in 6x(0-π) phasing for slow wave: kT ≈ 14m-1

0.5

0.0

0123

0

50

0.00.10.20.3

0.0 0.1 0.2 0.3 0.4 0.5Time [s]

PHHFW [MW]

WMHD, We [MJ]

ne(0), <ne> [1020m-3]

Ip [MA]

Hα [arb]

0.5 1.0 1.5Radius [m]

0.0

0.1

0.2

0.3

0.4

0

1

ne [1020m-3]

Te [keV]

0.243s

0.293s

0.343s

0.393s

0.243s

0.293s

0.343s

0.393s

LeBlanc


Frequency = 14 GHz

Electron Bernstein Wave: Oblique "O-X-B" Launch Is Resilient to Changes in Edge Density Gradient

• Optimum n// = 0.55; toroidal angle ~ 34o

from normal to B

• > 75% coupling for O-X-B antenna with ± 5 degree beam spread

EBW Coupling (%)80604020

OPTIPOL/GLOSI

Efficient conversion between ECW and EBW predicted.


EBW Emission Shows Near-Circular Polarization and Trad/Te ~ 70%, Consistent with Modeling

0.4

1.6

1.2

0.8

0

1.0

2.0

ThomsonScatteringTe (keV)

TotalEBW

Trad (keV)

(IncludingWindow/Lens

Loss)

Ratio ofRadiometer

Signals

Time (s)

FieldPitch(Deg.)

10

40

Magnetic Field Pitch35-40 Degrees

Frequency = 16.5 GHz

OPTIPOL/GLOSI

Pol

oida

l Ang

le (d

eg.)

Toroidal Angle (deg.)

-90 900

0

-90

90

> 90% EBW Coupling

B

10%

10%

Approx. Antenna Acceptance AngleAverage Coupling ~ 70%


Bt = 3.75 kGb = 42%

28 GHz

Modeling Predicts that 28 GHz EBW Can Drive Efficient Off-Axis Current at Plasma β ~ 40%

• EBW ray tracing, deposition and CD efficiency being studied withGENRAY & CQL3D for frequencies between 14 to 28 GHz

GENRAY/CQL3D [Bob Harvey, CompX]

Frequency = 28 GHzEBW Power = 3 MWTotal Driven Current = 135 kA

30

15

0 0.2 0.80.60.4

1.5

Z(m)

-1.5R(m)

1.40.4

Antenna

r/a

Cur

rent

Den

sity

(A/c

m2)


Strong Diffusion Near Trapped-Passing BoundaryEnables Efficient Ohkawa Current Drive

CompX GENRAY/CQL3D

NSTX, βT = 42%


ST Plasma Edge Possesses Large Mirror Ratio &Geometric Expansion of Scrape-Off Layer (SOL)

0

10

20

30

1.55 1.60 1.65 1.70R (m) in SOL

A B

Are

a Ex

pans

ion

Rat

io Larger GeometricExpansion

0

0.50

0.75

1.00

0 2 4 6 8

Distance Along SOL Field Line (m)

inboardoutboard

|B| (

T)

Larger Mirror Ratio (MR)→ More Instabilities→ Larger ⊥ Loss→ Thicker SOL

Scrape-Off Layer Geometryof Inboard Limited ST Plasma

0.25 0.5 0.75 1 1.25 1.5

-1

-0.5

0.5

1

1.5

R(m)

Z(m)

0

APlasmaFlux toLimiter

Divertor

Limiter

SOL FluxSurface

PlasmaEdge

B:PlasmaFlux ToDivertor


Increased SOL Mirror Ratio (MR) ⇒ Increased Footprint & Decreased Peak of Divertor Heat Flux

High & Low δ Divertor Bolometer Measurements

Rdiv (m) 0.36 0.75

SOL MR ~ 3 ~ 1.5

∆div (m) ~ 0.3 ~ 0.12

Factor of ~2 in Rdiv and MR⇓

Factor of ~3 in ∆div

Why?

SOL MR ≈ 3 SOL MR ≈ 1.5


Plasma Edge Studies Reveal Turbulence and“Blobs” Important to Divertor Flux Scaling Studies

He manifold

Side-viewingreentrant window

H-mode

L-mode

Broadly Based Study:• Gas Puff Imaging

views along field lines (PPPL, LANL)

• Very fast camera, 105/s (PSI)

• Reflectometers and edge (UCLA, ORNL)

• Reciprocating probe(UCSD)

• Divertor fast camera(Hiroshima U)

• IR Cameras (ORNL), Filterscope (PPPL)

• Modeling (PPPL, UCSD, LLNL, Lodestar)

105710


Tray after ~40 discharges.

Liquid lithium tray limiter in CDX-U

CDX-U Is Testing Innovative Lithium Plasma Facing Component Effects, to Control Recycling

• First successful test of toroidal liquid lithium tray limiter• Dramatic reduction in plasma edge fuel recycling, lowering impurity

influx and loop voltage• NSTX tests of lithium pellets and lithium wall coating in 2004

PPPL, UCSD, ORNL, SNL


CHI + OH

OH only

HIT-II Experiment

Culham, KAIST, Kyushu-Tokai U, PPPL, U Tokyo, U Washington

New absorber insulator installed

Capacitor bank to be installed

Three Outer Poloidal Field Startup Scenarios, e.g.:

Outboard Field Null

Flux contours 20 kA

- 20 kA

2.8 kA

Coaxial Helicity Injection Tests

Solenoid Free Start-Up via Coaxial Helicity Injection & Outer Poloidal Field Coil Are Being Tested


JT-60U Tests on Solenoid-Free Start-Up via RF and NBI Offers Additional Exciting Opportunities

Startup RF + NBI

RF Ramp-up

U. Tokyo, JT-60U – JAERI

• JT-60U: from 200 kA to 700 kA with LHW + NBI (2002)

• PLT: 100 kA with LHW (1980s)

• CDX-U, TST-2: up to 4 kA with ECH

• MAST: 1-MW ECH

• NSTX: to develop and test up to 4-MW EBW in 5 years

• Utilize outer PF coil induction with simple ramp


High βT

High τE

Co-NBI plasmas:• Improved vertical

control ⇒ higher κ

• βp t 1 and IBS/Ip d 0.5

• βN ~ 6 and βT > 20%

• Reduced VL

• Help developing ITER hybrid scenario

• Driven steady state ST plasmas (CTF).

• Need to reduce ELM size

Nearly Sustained Plasmas with Broader Values of κ, li, Ip, and βT Can Contribute to ITER Hybrid Scenario


NSTX Made Large Progress in Producing and Studying the Science of Attractive Sustained Plasmas

FY01-03

• EFIT02• Peak βT• All shapes

NSTX Potential

FY04

Frac

tion

of S

elf-D

riven

Cur

rent

f B

S~

0.5

×ε½

β P


Long-Pulse H-Mode Plasmas Made Large Progress in Physics Basis for Next-Term ST Science Facilities

Well positioned to address the science of sustained high-performance plasmas.

ARIES-AT (4MW/m2)

CTF-ST}

τpulse/τE Normalized pulse length

Nor

mal

ized

bet

a x

norm

aliz

ed c

onfin

emen

t NSTX OperationARIES-ST (4WM/m2)

4

2

1MW/m2

NSST


Research Topics to Achieve Long-Pulse, High Performance Plasmas Are Identified

• Enhanced shapingimproves ballooning stability

• Mode, rotation, and error field control ensures high beta

• NBI and bootstrap sustain most of current

• HHFW heating may contribute to bootstrap

• EBW provides off-axis current & stabilizes tearing modes

• Particle and wall control maintains proper density

CompX, MIT, PPPL, ORNL, UCSD


Answering the Plasma Science Questions Also Enable Cost-Effective Steps toward Fusion Energy

Smaller unit size for sustained fusion burn

⇒How does turbulence enhance transport?

Efficient fusion α particle, neutral beam, & RF heating

⇒How does plasma particles and waves interact?

Simplified smaller design, reduced operating cost

⇒How to supply magnetic flux without solenoid?

Lowered magnetic field and device costs

⇒How does shape determine pressure?

Survivable plasma facing components

⇒How do hot plasmas interact with wall?

Optimize Fusion DEMO & Development Steps

⇒Plasma Science Questions in Extended ST Parameter Space


Future ST Steps Are Estimated to Require Moderate Sizes to Make Key Advances toward DEMO

6030~15~0.01~0.01Duty factor (%)

~2.1

~12

~4

~300~50

~5

~2.6

~10

0.3

1

5

Single-turn; No-solen.Single-turn; No-solen.Multi-turn; SolenoidMulti-turn; SolenoidTFC; Solenoid

~4~1−−WL (MW/m2)

~3100~77~10−Pfusion (MW)

Steady stateSteady state~501Pulse (s)

~1.8~1.10.6BT (T)

~25~9~51.5Ip (MA)

~3.2, ~0.5~3, ~0.5~2.7, ~0.72.5, 0.8κ, δ

~2~0.8~0.90.65a (m)

~3~1.2~1.50.85R (m)

Practicality of Fusion Electricity

Energy Development, Component Testing

Performance Extension

Proof of PrincipleMission

DEMOCTFNSSTNSTXDevice


ST Research Has Broad and Growing Opportunities for Collaborations

• Exploratory ST’s in Japan– TST-2: ECW-EBW initiation– TS-3,4: FRC-like β~1 ST plasmas– HIST: helicity injection physics– LATE: solenoid-free physics

• Active participation in ITPA (ITER)– A and β effects on confinement, ITB,

ELM’s, pedestal, SOL, RWM, and NTM; scenarios, window coating, etc.

• ST Database with MAST, U.K.– NBI H-mode, transport, τE

– EBW H&CD (1 MW, 60 GHz), FY03– Divertor heat flux studies, FY03-04– NTM, ELM characterization

• DIII-D & C-Mod collaboration– Joint experiments: RWM, Fast ion

MHD, pedestal, confinement, edge turbulence, X-ray crystal spectrometer

• MST: electromagnetic turbulence, EBW

C-Mod (U.S.)DIII-D (U.S.)

MST (U.S.)

MAST (U.K.)


Spherical Tokamak (ST) Offers Rich Plasma Science Opportunities and High Fusion Energy Potential

• Early MHD theory suggested ST could permit high β, confirmed recently by experiments

• Recent research identified new opportunities for addressing key plasma science issues using ST• Results have been very encouraging in many scientific

topical areas

• ST research contributes to burning plasma physics optimization for ITER

• ST enables cost-effective steps toward practical fusion energy

• ST research is highly collaborative worldwide


Backup VUs


Minimizing Tokamak Aspect Ratio Maximizes Field Line Length in Good Curvature ⇒ High β Stability

Tokamak Compact Toroid (CT)

Spherical Tokamak (ST)

Bad CurvatureGood Curvature

Magnetic Field LineMagnetic Surface

Small-R close to Tokamak & large-R close to CT.


ST Is Closest to Tokamak; Operates with High Safety Factor and More Comparable Self & Applied Fields

Field Reversed Configuration

Reversed Field PinchDipole

Spheromak

Impr

ove

Plas

ma

Stab

ility

at H

igh

β→

0A

vera

ge S

afet

y fa

ctor

, qav

g5

ExampleExampleFusion ConfigurationsFusion Configurations

Spherical Tokamak(Spherical Torus)

(NSTX, MAST, TST-2,TS-3,4, HIST, LATE, etc.)

Tokamak &Advanced Tokamak

(DIII-D, C-Mod,K-Star, JT60-U, etc.)

Stellarators(QPS)In Design

(NCSX)Being Built

0.5 (Applied Field)/(Applied + Plasma-Produced Field) 1(Self-Organized) Increase Controllability → (Externally Controlled)

(LDX, etc.)

(SSPX, etc.)

(MST, etc.)

(LDX, etc.)

LHD


In MAST, However, Counter NBI Reduces Electron Energy Loss

High flow shear scenario on MAST (Co- & Counter-NBI)

• Counter-NBI produces stronger ωSE ~ 106 s-1 and strong local reduction in χe at broader radius

• Pressure gradient contribution to Er reinforces that due to Vφ with ctr-NBI

• Strong ExB flow shear and weak magnetic shear s ~ 0 produced by NBI heating during current ramp

• With co-NBI ion thermal transport reduced to N.C. level χi ~ χi

NC

with weaker reduction in χe

• Strong ExB flow shear ωSE > γm

ITG and s ~ 0 at minimum of χi,e

0.0 0.2 0.4 0.6 0.8 1.00

3

4

5

2

1

0.0

1.5

1.0

0.5

ρ

[keV

]

ne

Ti

Te

.2 s

0.0 0.2 0.4 0.6 0.8 1.00

3

4

5

2

1

0.0

1.5

1.0

0.5

ρ

[10

19 m

-3]

#8302, 0.2 s

0.0 0.2 0.4 0.6 0.8 1.0

ρ

[m

2 s

-1]

8575, 0.2 s

1.0

100

10

0.1

χi

χeNC

0.0 0.2 0.4 0.6 0.8 1.0

ρ

#8302, 0.2 s

1.0

100

10

0.1

#8

χφ

χiN

#830

Co-NBI Counter-NBI


Detailed Diagnosis and Gyrokinetic Analysis of β ~ 1 Turbulence Has Broad Scientific Importance

Can k¦ρι ≥ 1 turbulence at β ~ 1 be understood?

Region to be tested

Armitage (U. Colorado)

• Astrophysics turbulence dynamics: cascading of MHD turbulence to ion scales is of fundamental importance at β > 1

• Fusion’s gyrokinetic formalism apply to astrophysical turbulence, covering shocks, solar wind, accretion disks

• Laboratory ST plasmas provide validation of formulism

Gyrokinetic turbulence simulation in accretion disk of supermassive black hole at galactic center, assuming damping of turbulence by plasma ions vs. electrons


Single-turn demountable center leg for toroidal field coil required to achieve small size and simplified design.Fast remote replacement of all fusion nuclear test components (blanket, FW, PFC) & center post required to permit high duty factor & neutron fluence.Large blanket test areas ∝ (R+a)κa.

Adequate tritium breeding ratio & small fusion power from low Arequired for long term fuel sufficiency.High heat fluxes on PFC.Initial core components could use DEMO-relevant technologies (such as from ITER and long-pulse tokamaks).12-MA power supply – Single-turn TF.

Features Required by High DutyFactor & Neutron Fluence

Optimized Device Configuration Features of ST Can Fulfill the CTF Mission Effectively

columbia u spherical tokamak plasma...

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