Tungsten as First Wall Material in Fusion Devices: Advancements and Comparisons

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

2. Tungsten as First Wall Material in Fusion Devices • Introduction • Plasma Wall Interaction with Tungsten • Edge and Core Transport • Technological Developments • Summary

3. 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.

4. 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

5. FTU (ENEA Frascati) until 1994:poloidal limiter(steel, TZM, W) now: toroidal limiter TZM M.L. Apicella, et al., J. Nucl. Mater. 313-316

6. Alcator C-Mod (MIT) Divertor configuration witha complete set of Mo-tiles B. Lipschultz et al., Phys. Plasmas 13

7. ASDEX Upgrade (IPP Garching) Stepwise approach: remaining parts will be covered with tungsten in the 2007 campaign! R. Neu et al., Nucl. Fusion 45

8. Graphite versus Tungsten negative positive 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

9. 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!

10. Graphite versus Tungsten negative positive 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

11. Graphite versus Tungsten negative positive 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

12. Cchem (800K) (R.T) Tungsten: Erosion versus Radiation • W-erosion much lower than graphite!

13. LZ Tungsten: Erosion versus Radiation • But central W-radiation much higher!

14. Ignition Condition: Tungsten vs. Carbon

15. 10000 G. Sergienko, A. Pospieszczyk et al. 9000 8000 7000 #98038 6000 5000 4000 3000 WI(400.875nm) 2000 WI(407.436nm) OII(407.216nm) OII(407.587nm) 1000 CaII(396.847nm) OII(397.326nm) 0 OII(395.437nm) OII(409.725nm) 394 396 398 400 402 404 406 408 410 wavelength / nm 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

16. Tungsten as First Wall Material in Fusion Devices • Introduction • Plasma Wall Interaction with Tungsten • Edge and Core Transport • Technological Developments • Summary

17. Plasma Wall Interaction • Low erosion + no formation like hydro-carbons  low hydrogen retention (0.1 …1% instead of 40…100%) R. Causey, J. Nucl. Mater. 300J. Roth, M. Mayer, J. Nucl. Mater. 313-316 • W: high mass, low velocity of eroded particles ionization length << gyro radius  90% prompt redeposition C W D. Naujoks et al., Nucl. Fusion 36

18. Erosion on Target Plates/Limiter V. Philipps et al., PPCF 42

19. n r 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

20. 3+8 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

21. 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.

22. 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

23. 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.

24. 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

25. Tungsten as First Wall Material in Fusion Devices • Introduction • Plasma Wall Interaction with Tungsten • Edge and Core Transport • Technological Developments • Summary

26. 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.

27. n r 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

28. Influence of Anomalous Transport A peaked density profile without strong anomalous transportleads to strong tungsten accumulation. Central heating overcompensates neoclassical inward drift by anomalous transport! A. Kallenbach et al., Plasma Phys. Control. Fusion 47

29. 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!

30. W-concentration Erosion and transport determine concentration. AUG W-concentration strongly depending on discharge conditions!

31. Tungsten as First Wall Material in Fusion Devices • Introduction • Plasma Wall Interaction with Tungsten • Edge and Core Transport • Technological Developments Tungsten Coatings Massive Tungsten • Summary

32. 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.

33. 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!

34. 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

35. The ITER reference design test at FZJ

36. DEMO negative positive 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?

37. 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.

38. Developments for DEMO • Surfaces with reduced load: • A few mm tungsten sheets on EUROFER by PS or CVD IPP, Petten. FZJ

39. WSi0.82Cr0.45: WSi0.82: Tungsten: Oxidation rate (mg cm-2 s-1) 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

40. 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.

41. Summary • Altogether tungsten as the first wall material looks promising. • However, several open questions still remain to be solved.

42. Reserve

43. Sources for W-Erosion: ELMs Erosion on target plates:

44. Sources for W-Erosion: ICRH ASDEX Upgrade: Localized measurement on ICRH-antenna Fast (< 1ms) and localized increase  increase due to sheath rectified E-fields

45. Transport in the H-Mode Pedestal Argon seeding has to be well controlled!

46. 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)

47. Plasma Wall Interaction • Blistering: