Magnetic Fusion and Progress in Spherical Torus Research

1
Magnetic Fusion and Progress in Spherical Torus
Research
Roger Raman, Univ. of Washington For the NSTX
Team Thanks to Dr. Michael Bell
(PPPL) University of Rochester Laboratory for
Laser Energetics, 7 December 2007 Rochester, New
York
College WM Colorado Sch Mines Columbia
U Comp-X General Atomics INEL Johns Hopkins
U LANL LLNL Lodestar MIT Nova Photonics New York
U Old Dominion U ORNL PPPL PSI Princeton
U SNL Think Tank, Inc. UC Davis UC
Irvine UCLA UCSD U Colorado U Maryland U
Rochester U Washington U Wisconsin
Culham Sci Ctr U St. Andrews York U Chubu U Fukui
U Hiroshima U Hyogo U Kyoto U Kyushu U Kyushu
Tokai U NIFS Niigata U U Tokyo JAERI Hebrew
U Ioffe Inst RRC Kurchatov Inst TRINITI KBSI KAIST
ENEA, Frascati CEA, Cadarache IPP, Jülich IPP,
Garching ASCR, Czech Rep U Quebec
2
Outline

• Fusion basics
• Tokamak concept (the ITER device)
• Significance of Spherical Torus concept
• Divertor technology
• Plasma startup

3
Fusion Reactions of Interest for Terrestrial
Fusion Power
Plasma
Solid
4
Binding Energy, Fission, Fusion and Reaction Rates
DT
103
1
D-D
Bremsstrahlung
102
10-1
Fusion power density (AU)
Pressure of isothermal plasma (atm)
Pressure
10
10-2
10-3
1
nD nT 5 x 1019m-3 Te Ti
Mass defect mass of individual nucleons in a
nucleus mass of nucleus
10-4
0.1
102
10
1
103
Temperature (keV)
Ideal Breakeven temperature PF gt PBrem (perfect
confinement) Lawson condition Accounts for
particle losses and conversion efficiency Ignition
If alpha heating alone is enough to heat new
fuel and make up for radiation losses
5
Requirements for Magnetic Confinement D-T Fusion
Energy Development Were Understood Very Early

• Plasma conditions Lawson Criterion
• Q ? Pout/Pin gt 10 requires Ti 10  20 keV, ntE
  (6  3) ? 1020m-3s
• plasma heating, fueling, confinement, radiation
  losses
• Fusion power density 5 MWm-3 ? p 10 atm
• Need to maximize b ?p?/ Bmax2
• MHD stability and coil engineering
• Control interaction of plasma with surrounding
  material wall
• 2 MWm-2 thermal load on wall
• low impurity levels, low tritium retention
• Neutron wall loading 4 MWm-2 (needed for
  economic feasibility)
• material damage 40 dpa/yr with low radioactive
  waste
• self-sufficient tritium breeding to complete the
  fuel cycle
• High-duty cycle, essentially steady-state

J.D. Lawson, Proc. Phys. Soc. B, 70 (1957) 6
6
Two Main Fusion Concepts
Toroidal Magnetic Tokamak Stellarator
Spherical Intertial Transient Compression - eg.
Laser
Density very low (10-5 x atmospheric density),
so fuel needs to be held together For long times
(confinement time) - Pressure needs to be a few
atmospheres for sufficient power density
Density is very high, fuel heated to needed
temperatures well before density decreases
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Elements of a Fusion Power Plant
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Toroidal Magnetic Confinement Schemes - Closed
Traps

• Plasma in a simple torus doesnt have an
  equilibrium

• Gradient in B cause single particles to drift
  vertically
• Charge separation at the edges produces a
  downward E field that drives outward drift of
  plasma

• Introduce rotational transform (helical twist) to
  field lines so drifts are compensated over
  several transits
• external windings, geometrical modification
• toroidal current in the plasma itself

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Toroidal Confinement - The Tokamak Approach

• Toroidal plasma current adds a poloidal magnetic
  field to the externally applied toroidal field
  causing field lines to spiral

• Field lines form nested flux surfaces surrounding
  a magnetic axis
• Collisions cause plasma to drift outward from one
  surface to the next
• This neoclassical (Pfirsch-Schlüter) diffusion
  adds to classical diffusion
• Variation of the toroidal field from outside to
  inside traps some particles in local magnetic
  mirrors
• Trapped particles have larger orbit excursions,
  adding to diffusion
• A challenge is to drive toroidal plasma current
  continuously and efficiently
• Trapped particles plus a pressure gradient drive
  bootstrap current

10
Status of Laboratory Experiments - Lawson Diagram
Ti required for fusion has been achieved, but
needs 10x ntE Achieved ntE 1/2 required for
fusion, but needs 10xTi After 50 years,
MFE is 10 of the way. Requirements depend on
plasma profiles, impurities, synchrotron
radiation, etc
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ITER will Demonstrate the Scientific and
Technological Feasibility of Fusion Power

• ITER is a dramatic step towards self- sustained
  fusion reactions
• 500 MW(th) for gt400 s with gain gt10
• but ...
• ITER is not a self-sufficient power-producing
  plant
• New science and technology are needed for a
  demonstration power plant
• 2500 MW(th) with gain gt25, in a device with
  similar size and field
• Higher power density
• Efficient continuous operation
• Tritium self-sufficiency
• Research programs are needed to address these
  issues

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Experiments Around the World Are Investigating
and Attempting to Optimize the Magnetic
Configuration
JET, LargeTokamak EU
C-Mod,Tokamak MIT
W7-X, Large Superconducting Stellarator EU
National Spherical Torus Experiment PPPL (also
MAST EU)
EAST, SST-1, KSTAR Superconducting Tokamaks,
China, India, Korea
DIII-D, Tokamak General Atomics
LHD, Large Superconducting Stellarator JA
JT-60U, LargeTokamak JA
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Spherical Torus Extends Tokamak to Extreme
Toroidicity

• Motivated by potential for increased ? Peng
  Strickler, 1980s
• ?max ( 2?0?p?/BT2)
• BT toroidal magnetic field on axis
• ?p? average plasma pressure
• Ip plasma current
• a minor radius
• ? elongation of cross-section
• A aspect ratio ( R/a)
• q MHD safety factor (gt 2)
• Confirmed by experiments
• ??max 40
• START (UK) 1990s

Spherical Torus A 1.3, qa 12
Conventional Tokamak A 3, qa 4
Field lines
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NSTX Designed to Study High-Temperature Toroidal
Plasmas at Low Aspect-Ratio
Slim center columnwith TF, OH coils
Conducting platesfor MHD stability
Aspect ratio A 1.27 Elongation ? 2.5
(3.0) Triangularity ? 0.8 Major radius
R0 0.85m Plasma Current Ip 1.5MA Toroidal Field
BT0 0.6 (0.55) T Pulse Length 1.5s Auxiliary
heating NBI (100kV) 7 MW RF (30MHz) 6
MW Central temperature 1 3 keV
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NSTX Extends the Stability Database Significantly

• A 1.5
• k 2.3
• dav 0.6
• q95 4.0
• li 0.6
• bN 5.9mT/MA
• bT 40 (EFIT) 34 (TRANSP)

• Seeing benefits of
• Low aspect ratio
• Cross-section shaping
• Stabilization of external modes by conducting
  plates

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MAST NSTX show good confinement (?E gt 100ms)
NSTX

• Little difference in scaling for L and H modes
• In general agreement with IPB98y2 scaling
• Extends the range of R/a in scaling database
• MAST has obtained H-modes in an ohmic plasma

MAST NSTX Teams
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NSTX Approaches Normalized Performance Needed for
a Spherical Torus - Component Test Facility
(ST-CTF)

• Design optimization for a moderate Q driven
  ST-CTF
• Minimize BT required for desired wall loading ?
  Maximize ltpgt/BT2 bT
• Minimize inductive current ? Maximize fbs ?
  ?0.5?P
• Do this simultaneously ? Maximize fbsbT ? ?0.5?PbT

Goal of a driven ST-CTF DT neutron flux 1
4 MW/m2 Achievable with A 1.5, k 3, R0
1.2m, IP 8 12MA bN 5 .m.T/MA, H98y,2
1.3 bT 15  25 fBS 45  50
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NSTX Is Making Good Progress Toward the ST-CTF
While Contributing to the Physics Basis of ITER

• Ability of the ST to achieve high ? now well
  established
• Advanced mode stabilization methods and
  diagnostics are being applied to improve
  performance
• Dynamic Error Field Correction and RWM feedback
  suppression
• Unique tools available to study transport and
  turbulence
• Excellent laboratory in which to study core
  electron transport
• Investigating fast-ion instabilities
• Capability to mimic ITER situation
• Developing non-inductive startup and sustainment
  schemes
• Coaxial Helicity Injection
• Developing methods for heat flux and particle
  control
• Lithium coating of plasma-facing components,
  radiative divertors

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ST Program is an international effort
UK
USA
Russia
Brazil
Japan
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ORNL, PPPL
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Divertor Technology
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Natural divertor plasmas may offer an alternate
configuration for high performance discharges

• Formation of the Natural Divertor at low R/a
• As aspect ratio A decreases, exhaust plume expands

Inboard limited plasmas have an expanded outer
SOL Reduced, evenly spread contact on the
centre limiter. Exhibited H-mode features with
ELM-free periods
MAST Team
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Reduced Peak Heat Flux by Radiative Divertor

• Outer strike point heat flux reduced by 4-5
• No change in H-mode tE

Obtained by steady-state D2 injection into
private flux region
Small A Heat loading higher on outer divertor
plates
Soukhanovskii, et al., IAEA FEC2006, EXP4-28
LLNL
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CDX-U studies role of Lithium PFCs on plasma
operations and practical implementation issues

• Recycling Fueling
• Impurity reduction
• Performance enhancement
• Radiation lossses, core Li accumulation
• Safety issues
• Motion of liquid during PF ramps, disruptions

TiC coated shield on CS
Heat/Li shield
Tray temp. monitored
Tray has a radius of 34cm, width of 10cm, depth
of 6mm, Temp. 300C Electrical break between
two halves, Liquid Li injected into both halves
CDX-U PPPL
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A new lithium tray has been installed and
filled.Global recycling greatly reduced by clean
lithium
Improved filling technique developed by UCSD -
PISCES group Note reflections in metallic lithium
Pre-lithium fueling
Post-lithium fueling
Ip (kA)
Ip (kA)
Fueling
Fueling
Oxygen, carbon impurities virtually
eliminated Immediate 30 increase in peak plasma
current, discharge duration Loop voltage to
sustain current dropped from 2.0 ? 0.5V
Prefill only fuels entire discharge
No. of particles in CDX-U plasma
CDX-U PPPL UCSD
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In 2006, Lithium Evaporator (LITER) Experiments
Improved Particle Pumping and Energy Confinement
in H-mode
25 increase
Pre-Li
Post-Li
LITER
Pre 121270 Post 121323 t 510ms
10-15 decrease
40 increase

• TRANSP analysis
• WTOT 20 higher post-Li
• (reaches b-limit w/ same PNBI)
• HH98y 1.07 ? 1.25 post-Li
• Divertor D??emission dropped by a factor of 3-4

30 decrease
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During 2009-10, NSTX Plans to Test a Liquid
Lithium Divertor
A Candidate NSTX LLD Concept
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Basic Design Concept

• Mesh holds Li surface above container to reduce
  heat at the sides.
• Tiling of LLD sections reduces heating at the
  ends.
• Embedded heaters in deep mesh improve thermal
  control because mesh and Li are heated directly.
  Container can be cooler.

porous mesh
container ?SS
sheathed hot element
Li conforms to mesh surface but may or may not
wet the container
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Plasma Startup
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Solenoid-free plasma startup essential for the ST
concept

• Transient Coaxial Helicity Injection (CHI),
  previously demonstrated on HIT-II at U-
  Washington
• Important step in the production of a starting
  equilibrium for solenoid free operation
• Conventional tokamak uses solenoid
• ARIES-AT has no solenoid
• ST has advantages of high ß and good ?E
  
• ST reactor cannot use solenoid
• Alternate method for plasma startup is essential
  for ST concept
• Could also reduce the cost of a future tokamak
  reactor

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NSTX incorporates toroidal insulation breaks to
enable CHI operation
Transient CHI Axisymmetric reconnection leads to
formation of closed flux surfaces. Driven
(Steady State) CHI Non-axisymmetric modes needed
for closed flux generation
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Closed flux current generation by Transient CHI

• Abs. Arc

• Fast Camera (R. Maqueda, Nova Photonics L.
  Roquemore, PPPL)

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Equilibrium reconstructions consistent with
closed flux generation

• 2006 discharges operated at higher toroidal field
  and injector flux
• EFIT is done when no injector current is present
• Magnetic sensors and flux loops used in
  reconstruction

• LRDFIT (J. Menard)

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Movie of a high current discharge.

• Fast Camera
• R. Maqueda
• L. Roquemore

34

• 34

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Machine Assembly / Disassembly Schematic
Centerstack Assembly
Upper Blanket Assy Lower Blanket Assy
Upper PF coil Upper Diverter Lower Diverter Lower
PF coil
Upper Piping Electrical Joint Top Hatch
Shield Assembly
NBI Liner
Test Modules

• Disconnect upper piping
• Remove sliding electrical joint
• Remove top hatch

• Remove upper PF coil
• Remove upper diverter
• Remove lower diverter
• Remove lower PF coil

• Extract NBI liner
• Extract test modules
• Remove upper blanket assembly
• Remove lower blanket assembly

• Remove shield assembly

• Remove centerstack assembly

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NSTX is continuing to contribute to fundamental
toroidal confinement science in support of ITER
and future STs

• NSTX normalized performance approaching ST-CTF
  level
• Broad ITER and CTF-relevant boundary physics
  research program
• Developing Li technology
• Demonstrated closed-flux plasma formation in NSTX
  using CHI
• 160 kA world record for non-inductively generated
  startup current
• ST offers compact geometry high b attractive
  for CTF reactor
• ST based systems offer a cost-effective way to
  test many aspects of fusion technology for a DEMO
  (Fuelling, profile control, non-inductive startup
  and operation, mode control, fast particle
  physics)