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Tungsten as First Wall Material in Fusion Devices
M. Kaufmann Supported by H. Bolt, R. Dux, A.
Kallenbach and R. Neu
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Tungsten as First Wall Material in Fusion Devices

1. Introduction
2. Plasma Wall Interaction with Tungsten
3. Edge and Core Transport
4. Technological Developments
5. Summary

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Introduction PLT with tungsten limiter (1975)
V. Arunasalam et al., Proc. 8th Conf. EPS,
Prague 1977
Consequence of accumulation and central
radiation! Since then most tokamaks and
stellarators have used graphite as first wall
material.
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Tokamaks with High-Z-surfaces

• Limiter tokamaks
• FTU (ENEA)
• Textor (FZJ)
• Divertor tokamaks
• Alcator C-Mod (MIT)
• ASDEX Upgrade (IPP)
• future JET
• ITER

M.L. Apicella et al., Nucl. Fusion 37
A. Pospieszczyk et al., J. Nucl. Mater. 290-293
B. Lipschultz et al., Nucl. Fusion 41
R. Neu et al., Plasma Phys. Control. Fusion 38
J. Pamela, this conference
G. Janeschitz,J. Nucl. Mat. 290-293
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FTU (ENEA Frascati)
until 1994poloidal limiter(steel, TZM,
W) now toroidal limiter TZM
M.L. Apicella, et al., J. Nucl. Mater. 313-316
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Alcator C-Mod (MIT)
Divertor configuration witha complete set of
Mo-tiles
B. Lipschultz et al., Phys. Plasmas 13
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ASDEX Upgrade (IPP Garching)
Stepwise approach remaining parts will be
covered with tungsten in the 2007 campaign!
R. Neu et al., Nucl. Fusion 45
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Graphite versus Tungsten
positive
negative
graphite low central radiation
high erosion radiation in
boundary tritium co-deposition
forgives overload
destruction by neutrons
tungsten low erosion
high central radiation low
tritium co-deposition accumulation in
centre resistant to neutrons
critical with overload
radioactive, however, short decay
time
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Graphite versus Tungsten
tungsten low erosion
high central radiation low
tritium co-deposition accumulation in
centre resistant to neutrons
critical with overload

Test in linear machines of limited relevance!
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Graphite versus Tungsten
positive
negative
graphite low central radiation
high erosion radiation in
boundary tritium co-deposition
forgives overload
destruction by neutrons
tungsten low erosion
high central radiation no
tritium co-deposition accumulation in
centre resistant to neutrons
critical with overload

JET/ITER-generation
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Graphite versus Tungsten
positive
negative
graphite low central radiation
high erosion radiation in
boundary tritium co-deposition
forgives overload
destruction by neutrons
tungsten low erosion
high central radiation no
tritium co-deposition accumulation in
centre resistant to neutrons
critical with overload

DEMO-generation
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Tungsten Erosion versus Radiation

• W-erosion much lower than graphite!

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Tungsten Erosion versus Radiation

• But central W-radiation much higher!

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Ignition Condition Tungsten vs. Carbon
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Gain Experience Diagnostic
graphite extensive experience tungsten
limited experience

• W-lines at low temperature to determine influx
  (Textor)

W-lines at high temperature to determine core
concentration (AUG)
A. Pospieszczyk et al., to be published
A. Thoma et al., Plasma Phys. Control. Fusion 39
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Tungsten as First Wall Material in Fusion Devices

1. Introduction
2. Plasma Wall Interaction with Tungsten
3. Edge and Core Transport
4. Technological Developments
5. Summary

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Plasma Wall Interaction

• Low erosion no formation like hydro-carbons
  ? low hydrogen retention (0.1
  1 instead of 40100)

R. Causey, J. Nucl. Mater. 300J. Roth, M. Mayer,
J. Nucl. Mater. 313-316
W high mass, low velocity of eroded particles
? ionization length ltlt gyro radius ? 90
prompt redeposition
C
W
D. Naujoks et al., Nucl. Fusion 36
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Erosion on Target Plates/Limiter
V. Philipps et al., PPCF 42
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Sources for W-Erosion ELMs
Typical ITER reference H-mode
pressure profile forms steep edge
pedestal
Pedestal breaks down during ELMs!
ELMs produce main chamber erosion and target
plate erosion. In both cases sputtering by low
Z-components dominant.
A. Herrmann et al., accepted for publ. in J.
Nucl. Mater
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Sources for W-Erosion NBI

• Fast particles losses from neutral beam injection
  can be identified as a tungsten source on
  limiters.
• Increase during ELMs.

R. Dux et al., accepted for publ. in J. Nucl.
Mater
Quantitative agreement with calculations ?
Extrapolation to ITER no problem! R.Dux, to be
published
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Sources for W-Erosion ICRH
Alcator C-Mod
In ICRH heated plasmas without boronization
high radiation by molybdenum. Strongly reduced
by boronization. However, effect lasts only for
10s total pulse duration.
B. Lipschultz et al., Phys. Plasmas 13
Localized boronization by ECRH helps to identify
zone of Mo-erosion.
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Sources for W-Erosion ICRH
Conclusions

• small zone on top of divertor responsible
  for Mo-erosion.
• field lines map back to antenna.
• sheath potential 100-400eV

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Sources for W-Erosion ICRH
Can one reduce the sheath potential?
Faraday screen parallel to field lines small
effect
Is tungsten ITER/reactor compatible? ICRH reactor
compatible?
Lots of open questions!
Vl.V. Bobkov et al., accepted for publ. in J.
Nucl.
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Replacement of Carbon as Radiator

• Carbon radiates in the plasma boundary.
• It reduces therefore the load to the target
  plates considerably.
• It is highly self-regulating!
• Replacement by a noble gas such as Argon or Neon
  seems necessary Robust feed back method is
  needed!

Control by thermo currents through divertor
plates
controlled argon seeding
A. Kallenbach et al., J. Nucl. Mater. 337-339
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Tungsten as First Wall Material in Fusion Devices

1. Introduction
2. Plasma Wall Interaction with Tungsten
3. Edge and Core Transport
4. Technological Developments
5. Summary

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Neoclassical Transport
Neoclassical transport by Coulomb collisions
including drift motion leads to two fluxes.
diffusion
inward drift
Strong peaking of tungsten concentration in case
of peaked density profiles ( small) is
expected.
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Transport in the H-Mode Pedestal

• Steep density profile ? strong inward drift!

ELMs wash tungsten out!
High ELM frequency is required anyhow to reduce
load to target plates!
P. Lang et al., Nucl. Fusion 45
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Influence of Anomalous Transport
A peaked density profile without strong anomalous
transport leads to strong tungsten accumulation.
Central heating overcompensates neoclassical
inward drift by anomalous transport!
A. Kallenbach et al., Plasma Phys. Control.
Fusion 47
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Influence of Anomalous Transport
Anomalous transport induced by central heating
can easily overcompensate neoclassical inward
drift
Recent theoretical work no turbulent transport
mechanisms for strong high Z-ions inward drift!
C. Angioni, A.G. Peeters, Phys. Rev. Let. 96
In summary, one expects with a high probability
no peaked W concentration profiles in a burning
device!
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W-concentration
Erosion and transport determine concentration.
AUG
W-concentration strongly depending on discharge
conditions!
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Tungsten as First Wall Material in Fusion Devices

1. Introduction
2. Plasma Wall Interaction with Tungsten
3. Edge and Core Transport
4. Technological Developments Tungsten Coatings
   Massive Tungsten
5. Summary

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W-Coatings on Graphite

• In present day devices with low particle
  fluencies W-coating on graphite is used
  - because of lower eddy and halo currents.
  - because of lower weight.

Different techniques are available, e.g.-
physical vapor deposition (PVD)- chemical vapor
deposition (CVD)- plasma spray (PS)
H. Maier et al., accepted for publ. in J. Nucl.
Mater.
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W-Coatings on Graphite JET

• In JET the ITER like wall project is under
  preparation.
• The first wall will be partly covered with
  tungsten.

green Bered W-Coatingblue massive W
(probably) highly loaded areas 200µ sheath by
PSothers PVD
Highly loaded areas can be later replaced by
uncoated graphite!
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Massive W-Structures JET

• High particle fluencies (ITER, DEMO) massive
  W-structures are necessary.
• They are castellated - because of eddy
  currents (JET)- because of different thermal
  expansion (ITER, DEMO).

FZJ
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The ITER reference design
test at FZJ
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DEMO
positive
negative
graphite low central radiation
high erosion radiation in
boundary tritium co-deposition
forgives overload
destruction by neutrons
tungsten low erosion
high central radiation no
tritium co-deposition accumulation in
centre resistant to neutrons
critical with overload

DEMO-generation
Is ITER DEMO-relevant? Can the first wall be
exchanged?
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Developments for DEMO
Ductile to brittle transitiontemperature (DBTT)
high.Problem e.g. in W-steel-connections
He-cooled divertor (FZK)
Nuclear loads increase DBTT. Development of
W-alloys can reduce that problem.
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Developments for DEMO

• Surfaces with reduced load
• A few mm tungsten sheets on EUROFER by PS or CVD

IPP, Petten. FZJ
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DEMO Safety Issues
Loss of coolant and intense
air ingress formation of radioactive
WO3-compounds with high evaporation rate which
can leave hot vessel.
SEIF Study, EFDA-S-RF-1, April 2001
F. Koch, H. Bolt, subm. to Physica Scripta
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Summary

• In a fusion reactor, low-Z as a first wall
  material (graphite, Be) will have to be replaced
  by tungsten.
• So far, plasma experiments have demonstrated that
  in most scenarios the tungsten erosion of the
  surfaces and its concentration in the central
  plasma can be kept sufficiently low.
• In certain scenarios with high edge temperatures
  this may, however, not be the case.
• In addition, the high erosion in the
  neighbourhood of an ICRH antenna needs particular
  attention.
• As an intermediate solution, the coating of
  graphite with tungsten is an available
  technology.
• Technological solutions for the highly loaded
  divertor targets in a fusion reactor are under
  development.
• The relatively high ductile to brittle transition
  temperature, however, poses specific problems.

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Summary

• Altogether tungsten as the first wall material
  looks promising.

However, several open questions still remain to
be solved.
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Reserve
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Sources for W-Erosion ELMs
Erosion on target plates
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Sources for W-Erosion ICRH
ASDEX Upgrade Localized measurement on
ICRH-antenna
Fast (lt 1ms) and localized increase ?
increase due to sheath rectified
E-fields
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Transport in the H-Mode Pedestal
Argon seeding has to be well controlled!
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• Tungsten has 200 times larger conductivity than
  graphite,therefore eddy and halo currents
  larger.
• Tungsten has 8.5 times larger mass density than
  graphite.
• In case of low particle fluencies often
  W-coating on graphite are
  used.
• Different techniques are available, e.g.-
  physical vapor deposition (PVD)- chemical vapor
  deposition (CVD)- plasma spray (PS)

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Plasma Wall Interaction

• Blistering

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