1SOFT 2006 Warsaw

M. KaufmannSupported by H. Bolt, R. Dux, A. Kallenbach and R. Neu

Tungsten as First Wall Material in Fusion Devices

2SOFT 2006 Warsaw

Tungsten as First Wall Material in Fusion Devices

1. Introduction2. Plasma Wall Interaction with Tungsten3. Edge and Core Transport4. Technological Developments5. Summary

3SOFT 2006 Warsaw

Introduction: PLT with tungsten limiter (1975)

Consequence of accumulation and central radiation!Since then most tokamaks and stellarators have used graphite as first wall material.

V. Arunasalam et al., Proc. 8th Conf. EPS, Prague 1977

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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

5SOFT 2006 Warsaw

FTU (ENEA Frascati)

until 1994:poloidal 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 Tungstenpositive 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

9SOFT 2006 Warsaw

Graphite versus Tungstentungsten: 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!

10SOFT 2006 Warsaw

Graphite versus Tungstenpositive 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

11SOFT 2006 Warsaw

Graphite versus Tungstenpositive 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!

(R.T)

Cchem (800K)

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Tungsten: Erosion versus Radiation

But central W-radiation much higher!

LZ

0

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Ignition Condition: Tungsten vs. Carbon

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Gain Experience: Diagnostic

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

394 396 398 400 402 404 406 408 410

0

100020003000400050006000700080009000

10000

OII

(407

.587

nm

)

OII

(407

.216

nm

)

CaI

I(39

6.84

7nm

)

OII

(409

.725

nm

)

OII

(395

.437

nm

)

OII

(397

.326

nm

)

WI(

407.

436n

m)

#98038

WI(

400.

875n

m)

wavelength / nm

G. Sergienko,A. Pospieszczyk et al.

W-lines at high temperature to determine core concentration (AUG)

graphite: extensive experience tungsten: limited experience

A. Thoma et al., Plasma Phys. Control. Fusion 39

A. Pospieszczyk et al., to be published

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Tungsten as First Wall Material in Fusion Devices

1. Introduction2. Plasma Wall Interaction with Tungsten3. Edge and Core Transport4. Technological Developments5. Summary

17SOFT 2006 Warsaw

Plasma Wall Interaction

Low erosion + no formation like hydro-carbons low hydrogen retention (0.1 …1% instead of 40…100%)

W: high mass, low velocity of eroded particles ionization length << gyro radius 90% prompt redeposition

W C

D. Naujoks et al., Nucl. Fusion 36

R. Causey, J. Nucl. Mater. 300J. Roth, M. Mayer, J. Nucl. Mater. 313-316

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Erosion on Target Plates/Limiter

V. Philipps et al., PPCF 42

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Typical ITER reference H-mode pressure profile forms steep edge pedestal:

Sources for W-Erosion: ELMs

ELMs produce main chamber erosion and target plate erosion.In both cases sputtering by low Z-components dominant.

Pedestal breaks down during ELMs!

n

r

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.

3+8

Quantitative agreement with calculations Extrapolation to ITER: no problem! R.Dux, to be published

R. Dux et al., accepted for publ. in J. Nucl. Mater

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Sources for W-Erosion: ICRH

Localized boronization by ECRH helps to identify zone of Mo-erosion.

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

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Sources for W-Erosion: ICRH

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

Conclusions:

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Sources for W-Erosion: ICRH

Can one reduce the sheath potential?

Lots of open questions!

Faraday screen parallel to field lines: small effect

Is tungsten ITER/reactor compatible?ICRH reactor compatible?

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!

controlled argon seeding

Control by thermo currents through divertor plates:

A. Kallenbach et al., J. Nucl. Mater. 337-339

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Tungsten as First Wall Material in Fusion Devices

1. Introduction2. Plasma Wall Interaction with Tungsten3. Edge and Core Transport4. Technological Developments5. Summary

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Neoclassical Transport

Neoclassical transport by Coulomb collisions including drift motion leads to two fluxes.

Strong peaking of tungsten concentration in case of peaked density profiles ( small) is expected.nL

)/1)5.0...25.0(/1( TnD LLDZv

diffusion:

inward drift:

2/1 ZD

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Transport in the H-Mode Pedestal

Steep density profile strong inward drift!

n

r

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!

In summary, one expects with a high probability no peaked W concentration profiles in a burning device!

ZvZD D /1/1 2

C. Angioni, A.G. Peeters, Phys. Rev. Let. 96

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W-concentration

W-concentration strongly depending on discharge conditions!

Erosion and transport determine concentration.

AUG

31SOFT 2006 Warsaw

Tungsten as First Wall Material in Fusion Devices

1. Introduction2. Plasma Wall Interaction with Tungsten3. Edge and Core Transport4. 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.

33SOFT 2006 Warsaw

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!

34SOFT 2006 Warsaw

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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DEMOpositive 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?

37SOFT 2006 Warsaw

Developments for DEMODuctile 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

SEIF Study, EFDA-S-RF-1, April 2001

Loss of coolant and intense air ingress:formation of radioactive WO3-compounds with high evaporation rate which can leave hot vessel.

Tungsten:

WSi0.82Cr0.

45:

Oxidation rate (mg cm-2 s-1)

WSi0.82:

6 0 0 ° C

8 0 0 ° C

1 0 0 0 ° C

6 0 0 ° C

8 0 0 ° C

1 0 0 0 ° C

6 0 0 ° C

8 0 0 ° C

1 0 0 0 ° C

1 0

- 7

1 0

- 6

1 0

- 5

1 0

- 4

1 0

- 3

1 0

- 2

1 0

- 1

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 (< 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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