Problem and Issues for Tokamak longpulse Operation'''

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The control of magnetically confined plasmas (in
tokamak facilities) S. Brémond Institut de
Recherches sur la Fusion par confinement
Magnétique (IRFM) CEA Centre de Cadarache
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Outline

• A short reminder of the fusion energy source
  development issue
• Basics of tokamak operation
• Control issues overview
• - Basic controls
• - Performance optimisation, advanced scenario
• - Machine protection
• Control design needs
• Conclusion

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Fusion reaction
A SHORT REMINDER OF THE FUSION ENERGY SOURCE
DEVELOPMENT ISSUE 1/3
Nb of nucleons
FUEL AND HEAT IT UP ! (to about 100 million
degrees, plasma state) EXTERNAL HEATING
required before internal heating takes over
(collisions of the He with the D-T nucleus)
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Power balance
A SHORT REMINDER OF THE FUSION ENERGY SOURCE
DEVELOPMENT ISSUE 2/3
Gravitational confinement
CONFINE IT (to avoid power losses and fuel
dilution) !
Condition for overall power gain (Lawson
criteria) n x T x tE gt
threshold
Huge mass required (sun and stars)
Fuel density
Temperature
Confinement time
Inertial confinement
Compression of a milimetric target
n 103 x standard solids tE 10-9 seconde
Magnetic confinement
Effect of magnetic field on charged particles
n 10-5 air in this room tE 1 seconde
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Magnetic confinement 1/2
A SHORT REMINDER OF THE FUSION ENERGY SOURCE
DEVELOPMENT ISSUE 3/3

• The principle

Issue losses at both ends
Issue vertical drift
B m0 Ibob / 2p R Giration radius a 1/B
The toroidal configuration
Issue (collective effects between particles)
hoop force
Helicoïdal field lines
The tokamak configuration
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A SHORT REMINDER OF THE FUSION ENERGY SOURCE
DEVELOPMENT ISSUE 3/3
Magnetic confinement 2/2
Equilibrium field external coils
SET OF MAGNETIC CONFINEMENT COILS
Two options
Wall
Scrape-off layer (SOL)
Last closed flux surface
Limiter configuration
divertor configuration
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Tokamak plasma actuators 1/2
BASICS OF TOKAMAK OPERATION 1/4
Central solenoid coil (transformer primary)

• FUEL
• Gaz valves, pellets injector, Pumps
• CONFINE
• Transformer primary coil (limited time duration)
  non inductive current drive systems -gt drive
  plasma current
• - Poloidal Field coils
• -gt set plasma position and shape
• HEAT UP
• - ohmic heating (limited capabilities)
• - Wave, Beam heating systems

Poloidal Field coils (plasma position and shape)
Fuelling system (gas valve, pellet injector not
represented here)
Pumping (cryo pumps)
ITER project view
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BASICS OF TOKAMAK OPERATION 2/4
Tokamak plasma actuators 2/2
Heating / Current Drive
NEUTRAL BEAM ION CYCLOTRON WAVE Tens of MHz
(tetrode sources) ELECTRON CYCLOTRON WAVE Around
100 GHz (gyrotrons) LOWER HYBRID WAVES 2.5 GHz,
3.7 GHz (klystrons)
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Example of Tore Supra (CEA Cadarache)
BASICS OF TOKAMAK OPERATION 3/4
PERMANENT TOROIDAL MAGNETIC FIELD
supercondutctor NbTi coils ACTIVELY COOLED
PLASMA FACING COMPONENTS water cooling, exhaust
capability 10 MW/m2 (within ITER range)
Circular cross-section Current 1.5
MA Major radius 2.4 m Minor radius 0.72
m Volume 25 m3
CONFINED PLASMA
1 Electron Cyclotron antenna (118 GHz) 0.8 MW max
, 25 MJ
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Plasma discharge schedule
BASICS OF TOKAMAK OPERATION 4/4
ITER discharge foresseen schedule
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Control issues overview
CONTROL ISSUES 1/2
second
Basic controls (MHD equilibrium, density)
Machine protection Safety
Performance optimisation Advanced scenario
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CONTROL ISSUES 2/2
Feedback control need
Target scenario
CONTROLER / SUPERVISOR Sequencing, tracking of
references and machine protection

• INTRINSIC INSTABILITIES
• PERTURBATIONS (state of the wall, internal
  profiles self-organisation, etc. )
• COMPLEXITY OF THE PHYSICS at stake (accurate
  prediction very difficult)
• ?
• Real time feedback control required

Tokamak plasma
Reliable real-time measurements needed (not
covered here)
JET diagnostics overview
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Equilibrium control basics 1/3
BASIC CONTROL 1/4
BASIC MODELING (PLASMA POSITION)
MOTION EQUATION (Magneto Hydro Dynamic
description)
RIGID DISPLACEMENT MODEL
Zero order development in
(Poloidal) beta ratio of kinetic / magnetic
pressure
Plasma self inductance
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Equilibrium control basics 2/3
BASIC CONTROL 2/4
VERTICALLY ELONGATED CROSS SECTION

- happens to allow better plasma performance
require outward curvature - but require outward
field curvature
Vertical position unstable

• Huge force Tesla MAmp 2pR
• Low mass 800m3 1020 (per m3) 10-27
  (kg/ion)
• Very fast time scale ms (MHD Alfven time)
• But we dont have to react that fast (we could
  not actually)
• because eddy current will be induced in the
  metallic structures of the vessel which will on
  certain conditions- slow down the motion to their
  L/R characteristic time where feedback controlled
  active coils may take over

Ip
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Equilibrium control basics 3/3
BASIC CONTROL 3/4
PLASMA SHAPING
ITER Poloidal cross section

• Numerical Modeling
• Direct / Inverse problem (PF coils current -gt
  magnetic surfaces topology or the inverse) 2D
  free boundary equilibrium codes (CEDRES
  developed with Univ. Nice)
• Reconstruction problem (magnetic probes at the
  vessel -gt plasma last closed surface) 2D free
  boundary reconstruction code (EQUINOX developed
  with Univ. Nice, other approach for plasma
  boundary reconstruction only developed with INRIA
  Sophia)

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Particle density control
BASIC CONTROL 4/4
ne
FUELLING SOURCES

• Recycling a badly controlled dominant fuelling
  source (depending of the history / state of the
  wall)
• Gaz puffing edge fuelling
• Pellets injection more internal fuelling

Recycling
r
Complex particle transport phenomena
Control target / issues - impurity exhaust (He
ashes) - density profile optimisation (burn
control) - radiation control (divertor)
recycling
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PERFORMANCE OPIMISATION ADVANCED SCENARIO 1/2
Advanced scenario

• internal confinement enhancement
• Control of the internal profiles (plasma current
  density, pressure, etc.)
• MHD instables modes control
• Edge localised modes, sawteeth, neoclassical
  tearing modes, resistive wall modes, etc.

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PERFORMANCE OPIMISATION ADVANCED SCENARIO 2/2
Internal profiles control (current density)
Distributed plasma current density real-time
measurement
RT equilibrium reconstruction with the EQUINOX
code (developed with Univ. Nice)
Identification of the plasma current profile (and
free boundary equilibrium) from external
measurements
Distributed plasma current density real-time
control design
New design approach based on a distributed non
linear control oriented model
Diffusivity - interior - boundary control
? magnetic flux, ?//parallell electric
resistivity, jninon inductive current density
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Machine protection issues
MACHINE PROTECTION ISSUES 1/1

• Disruption detection
• Disruption mitigation
• Plasma First Wall Component protection from
  overheating

Needs - IR imaging wrapping to PFCs
geometry - High level real time data processing
(pattern recognition, heat flux assessment,multi
diagnostic processing, etc.) collaboration under
way with INRIA Sophia
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Control design process
CONTROL DESIGN PROCESS 1/1

• Existing control mainly based on semi empirical
  design
• Model based control needs for integrated modeling

Plasma discharge flight simulator under
development
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Conclusion

• Control of tokamak plasmas an old problem, new
  issues
• Performance target requires to (ITER needs)
• - operate very close to technological limits
  (while ensuring machine protection)
• - optimise plasma performance by controlling not
  only global variables, but also local ones
  (plasma internal profiles)
• - develop an integrated management of the plasma
  discharge
• pre-qualify control algorithms (safety
  licensing)
• Main needs
• - real-time data processing algorithms
  development
• - Integrated modeling tools (model based control
  design)