Riscaldamento e current drive del
plasma del reattore ITER
Torino 17 Maggio 2007
Arturo Tanga
ITER IO
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Sommario
• Introduzione: La sfida
scientifica/tecnologica dell progetto ITER
• I sistemi di riscaldamento e current drive
• Risk reduction e ricerca e sviluppo
• Iniettori di neutri
• Sistemi a radio frequenza
• Conclusioni
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What is ITER Today?
• ITER (“the way” in Latin) is the essential next step in
the development of fusion.
• Objective - to demonstrate the scientific and
technological feasibility of fusion power.
• The world’s biggest fusion
energy research project.
• An international
collaboration.
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The Mission
•
Up to steady state fusion power production.
•
Plasma makes 10x more power than needed to run it.
•
Optimise plasma behaviour.
•
Have dimensions comparable to a power station.
•
Produce about 500 MW of fusion power.
•
Demonstrate or develop all the new technologies required for fusion
power stations, except materials endurance.
•
Obtain license for construction and operation.
•
Operate for about 20 years.
•
Cost about €5bn to construct (over 9 years) and €5bn to operate
(about 20 years) and decommission.
Cadarache Site
The ITER building
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Construction Sharing
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Immagine del sito dopo
la costruzione
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Integrated Project Schedule
LICENSE TO
CONSTRUCT
ITER IO
2005
2006
2007
Bid
2008
TOKAMAK
ASSEMBLY STARTS
2009
2010
2011
2012
FIRST
PLASMA
2013
2014
2015
2016
EXCAVATE
Contract
TOKAMAK BUILDING
OTHER BUILDINGS
Construction License Process
First sector Complete VV
TOKAMAK ASSEMBLY
Install PFC
cryostat
MAGNET
Bid
Bid
Contract
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Install CS
COMMISSIONING
Vendor’s Design
Contract
VESSEL
Complete
blanket/divertor
PFC
TFC CS
fabrication start
First sector
Last TFC
Last CS
Last sector
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Contributo italiano all’ ITER
• L’ITER e’ vicino, solo 4 ore d’ auto.
• L’Italia ha una forte tradizione in Fusione per
ITER.
• Polito e’ socio fondatore nel consorzio per la
ciclotronica ionica (Bosia, Maggiora).
• Lo studio e stabilizzazione dei modi si basa sugli
studi fatti qui (Porcelli et gruppo)
• Lo sviluppo degli ioni neutri sara stabilito a
Padova.
• Diagnostiche saranno costruite a Milano e
Frascati.
• ITER RF design fatto da Bosia
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Main Features of the ITER Design
Central
Solenoid
Nb3Sn, 6
modules
Outer
Intercoil
Structure
Toroidal Field Coil
Nb3Sn, 18, wedged
Poloidal Field Coil
Nb-Ti, 6
Machine Gravity Supports
(recently remodelled)
Blanket Module
421 modules
Vacuum Vessel
9 sectors
Cryostat
24 m high x 28 m dia
Port Plug (IC
Heating)
6 heating
3 test
blankets
2 limiters/RH
rem.
diagnostics
Torus Cryopump
8, rearranged
Divertor
54 cassettes
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1022
Tipple Product - Diagramme, Status 1996
ITER
ignition
How do we size a
Reactor class Fusion
machine (1)
JT-60U
JET
1021
TFTR
break-even
JT-60U
DIII-D
H-mode
reversed shear
LHD, W7-X
1020
10
Ti (keV)
• Ti > 10 keV
=> ne x τE ~ 6.0 1020 m-3s
=> ne ~ 1.0 1020 m-3
=> τE ~ 6.0 s
H-mode Confinement Scaling Relation:
100
-> (to optimise D-T reaction rate)
-> (close to Density Limit: nL ~ I/a2π)
-> (defines more or less machine size)
ELMy
0.41 0.19 1.97 0.58 0.78
τE,th
= 0.0562× I 0.93B0.15P−0.69ne,19
M R ε κ
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Typical Example of an Inductive Tokamak Pulse (ITER)
Burn
Burn
Termination
Current
Rampdown
Current
Rampup
PFB
Pfus
-200
Heating
PF
Magnetization
Plasma Initiation
PF Reset
and Recool
End Pulse
Begin Pulse
0
tSOF tSOB
Time(s)
IP0
Ip
(~400s)
600
700 900
1400
ISOD
φSOP
φOH
φSOB
ΓDT
DT
Refuel
φEOB
neB
ne
5%
fHe
Paux EC(2MW)
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Current and “q” profile for an
inductive Tokamak pulse
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ITER - Mission
Expected long Pulse and Steady State
Performance in ITER
Technical
• Aim at steady state
with Q>5
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Heating & current drive systems on ITER
Design scenarios
•
Plasma operation has been designed with variable combination of
heating and current drive systems: 2 (3) NB H&CD injectors 33 – 50
MW, 20~40MW ECH, 20~40MW ICH, 0~40MW LH, 3 MW ECH for
start up, 3.5 MW DNB.
•
Baseline: The start-up configuration requires 33 MW NB,20MW
ICH, 20MW ECH,0MW LH, 3 MW ECH for start up, 3.5 MW DNB.
•
The machine configuration is consistent with the possibility of
implementation of various operating scenarios.
The H&CD systems cost app. 400Me
•
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2
20
1
20
1
0
0
73
4
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Neutral beam injection: principles
Accelerator
Ion
Neutraliser
source
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Residual ion
Dump (RID)
Plasma
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The injector
The Injector can be separated in
beam components (Ion Source, Accelerator, Neutralizer, Residual
Ion Dump and Calorimeter)
other components (cryo-pump, vessels, fast shutter, duct, magnetic
shielding, and residual magnetic field compensating coils)
9m
15m
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5m
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ITER NBI requirements
Neutral beam injection is required since the beginning of ITER
operation
The NBI system consists of 2 (+1) beams for Auxiliary Heating
and Current Drive
Beam parameters:
P=16.5MW
I=40A
V=1MV ( to heat the core plasma)
t pulse=3600s
1MeV neutrals implies negative ions for efficient neutralisation (60%)
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NBI injectors in ITER
2+1 NBI
DNB
Plan view
tangential injection
Vertical cross section view
On/off axis injection by tilting the beam axis vertically
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Ion source design
Drivers
RF driven
Variable
capacitors
(C2)
Arc driven
Front view
Fixed
capacitors
(C1)
8 RF drivers
72 filaments
Source
case
(b)
1m
Support
structure
Electrodeposited
copper
Bias bus
bar
Integrated
design
Magn
shield
SINGAP
extractor
PG bus
bar
Both sources mechanically compatible with reference and alternative accelerator
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Neutral beam Test Facility
Experiment
Power supply
Maintenance
Cooling towers
Auxiliary systems
At present work is in progress to adapt the generic site to Padova
site, which has been proposed by EU as the Test Facility site
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RF systems
IC H&CD: 1 antenna in un port equatoriale, 20 MW,
frquenza varaibile da 35 a 65 MHz
EC H&CD 1 antenna equatoriale 20 MW @170 GHz,
e 4 port superioriPer il controllo dei modi
LH CD 20 MW per il current drive (upgrade) @5.4 GHz
Ulteriori upgrade di IC e EC
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ICH Requirements
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Example of 2 antenna configurations
Side view
Front
Artist view
Back
• Matching units in interspace or port cell
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ICH&CD system
• Conceptual design in progress with important
contributions from EU, IN and US on the ICH system
components.
• Antenna concept guidelines chosen in May-June, and
will allow to specify the generators and transmission
lines.
• Procurement packages for Transmission lines (US) and
generators (IN) will be defined for October 2007.
• Detailed design phase for antenna port plug and port cell
matching units will start this summer for 3 years as an
EU collaboration CYCLE, in parallel with corresponding
R&D
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ITER ECH&CD System
170 GHz
gyrotrons
Start-up
gyrotrons
1. 170 GHz, 20
MW
2. 127.5 GHz, ≥2
MW, pulsed
[ Function]
Heating for Q>10 and
L-H transition,
Current drive for Steady
State operation,
Stabilisation of MHD
instabilities,
Plasma start-up assist.
Corrugated
waveguide
Upper
launchers
Equatorial
launcher
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Lavoro teorico
• Modelling of macroscopic magnetic
islands in tokamaks
• F. Porcelli, et al
a Istituto Nazionale Fisica della Materia
and Dipartimento di Energetica,
Politecnico di Torino,
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R&D Status of 170 GHz gyrotron
Specifications :RF output power≥
≥1MW, Pulse duration=400s~3600s (CW),
Efficiency≥
≥50% with a depressed collector.
Gyrotron (RF)
Gyrotron (JA)
Coaxial Gyrotron (EU)
0.6MW/250s,
0.6MW/1hr/46%, 0.82MW
CW tube was constructed.
0.5MW/300s,
/600s/56%, 1MW /800s /55% ). High power test in 2007.
0.95MW/70s.
Issues: Different Gyrotrons, magnets, power supplies & thermal requirements, spares
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Launchers with front steering mirrors
~6m
[Upper Launcher]
Miter bend
[Equatorial
Launcher]
[Steering range]
Upper launcher: Vertically 53º~69º
with 18º inclination for USM
and 39º~61º with the inclination of
20º for LSM.
Equatorial launcher: Horizontally
20˚~45˚
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Transmission lines
Window
Front
shield
Driver
unit
24 RF beams
20 MW
~3
Steering mirror
m
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High Power Transmission Line
at 170 GHz (63.5mm)
JAEA
Objectives
• High power rf experiment for the launcher components
• Demonstration of MW-level & CW transmission at 170GHz
In-line switch
Teflon attenuator
SS load
Pre-load
SS load
End of line
Switch
GV
Transmission line (~40m, waveguides, 6bends, 1switch, 2 loads)
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Lower Hybrid
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Why LHCD on ITER?
"ITER Plant Design Specification: Physics performances. ITER Device should:
-achieve extended burn in inductively driven plasmas with the ratio of fusion power to auxiliary
heating power of at least 10 for a range of operating scenarios and with a duration sufficient to
achieve stationary conditions on the timescales characteristic of plasma processes.
-aim at demonstrating steady-state operation using non-inductive current drive with the ratio of
fusion power to input power for current drive of at least 5." Hybrid scenario
•
•
LHCD is the most efficient non inductive method : efficiency @ 5GHz in the
range of 0.2~0.3x1020A/Wm² at (Te,ne) expected for ITER, far above NBI,
ICRH, ECRH.
Off-axis CD needed for long shots at high βN in ITER, and other systems are
inefficient
☺ Higher power density
☺ Less power absorbed
by α
q factor
5GHz :
Higher RF losses
LH
Radius
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LHCD System
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LHCD present concerns
•
Confirmation of scenarios :
–
–
•
0 power for start up (procurement package uncredited)
20 or 40 MW in later configurations
Confirm choice of frequency :
–
–
5 GHz correspond to low alpha absorption, but high technical difficulties
3.7 GHz well known, but larger absorption on alphas;
•
R&D program on source (if 5 GHz) and RF window and
transmission line technologies.
•
Provision of space for generators (assembly hall),
power supplies; cooling requirements (in start up
configuration)
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Conclusioni
• La costruzione di ITER e’cominciata.
• Essa durera circa 10 anni
• Una parte importante sono i sistemi di
riscaldamento e di current drive.
• Un grosso sforzo di ricerca e necessario per
raggiungere questo obbiettivo.
• Allo stesso tempo dovremmo provvedere gli
strumenti per fare gli esperimenti con queste
apparecchiature, ed I loro effetti sul plasma, per
raggiungere le prestazioni necessarie.
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