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I. Uytdenhouwen , J. Schuurmans , M. Decréton, V. Massaut SCK●CEN, Mol, Belgium

Installation of a Plasmatron at the Belgian Nuclear Research Centre and its use for Plasma-Wall Interaction Studies. I. Uytdenhouwen , J. Schuurmans , M. Decréton, V. Massaut SCK●CEN, Mol, Belgium G. Van Oost Ghent University, Belgium. Key issues in PWI Dedicated research facilities

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I. Uytdenhouwen , J. Schuurmans , M. Decréton, V. Massaut SCK●CEN, Mol, Belgium

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  1. Installation of a Plasmatron at theBelgian Nuclear Research Centre and its use forPlasma-Wall Interaction Studies I. Uytdenhouwen, J. Schuurmans, M. Decréton, V. MassautSCK●CEN, Mol, Belgium G. Van OostGhent University, Belgium

  2. Key issues in PWI • Dedicated research facilities • Plasmatron facility VISION I • Conclusion

  3. Model D or Self-cooled Model AB or HCLL Model A or WCLL Model B or HCPB Model C or Dual-Coola. ITER like wallJET(2 options) W alloy Structural material CuCrZr W alloy W alloy SiCf/SiC Armour material TUNGSTEN Coolant H2O He He LiPb He Coolant I/O T(°C) 140/167 ~ 540/720 ~ 600/990 ~ 540/720 ~ 540/720 ITER PPCS (inspiration for DEMO) Plasma facing materials

  4. Radiation Dilution Beryllium • Low Z, high oxygen gettering, good thermal conductivity, low solubility for hydrogen • Implantation with D, T (saturated very-near surface layers) BUT • Nuclear reactions breed T, He in bulk • High erosion yieldLow melting pointRES (radiation enhanced sublimation) • ionized Be in the plasma • deposition in divertor area • mixed materials issue (alloy formation) Loarte A. et al., NF 47 (2007) Causey R.A. et al., Fus. Eng. Des. 61 (2002)

  5. Stable Be-W alloys Mixed materials • Stable Be-W intermetallics are: • ~2200°C (Be2W) • ~1500°C (Be12W) • ~1300°C (Be22W) melting points closer to Be than to W! • What happens if Be transport into the W bulk is rapid enough that alloy formation is not limited to the near surface? Baldwin M. et al., J. Nucl. Mat. 363 (2007)

  6. Destruction by neutrons • High erosion yield Graphite • Low Z (low central radiation, radiation in boundary)Good thermomechanical properties Lack of melting BUT • T retention issue • co-depositions (surface) • depositions in gaps • n damage (bulk) Causey R.A. et al., Fus. Eng. Des. 61 (2002) Shimomura Y., J. Nucl. Mat. 363 (2007)

  7. LZ C W BUT • High radiative cooling rate (compared to C) Tungsten • Low erosion yield + no formation like hydro-carbons • low hydrogen retention (0.1 …1% instead of 40…100%) • High mass, low velocity of eroded particles • ionization length << gyro radius • 90% prompt redeposition J. Roth et al., J. Nucl. Mat. 313-316 (2003)

  8. BUT • Strongly dominated by • Low-Z intrinsic impurities • Higher sputtering by Ar, Ne • Mixed materials issue (even if only one PFM) Tungsten • Only limited concentration in plasma allowed (ppm range) • Limit W erosion (transients, sputtering, …) • High-Z impurity control by seeding with Ar, Ne • Sputtering yield by D negligible(100-1000 times smaller as for C) Federici G. et al., J. Nucl. Mat. 313 (2003)

  9. T retention due to trapping in bulkInfluenced by: • Damage sites (n irradiation) • Effective porosity (manufacturing technique, existance of cracks, …) Tungsten • Helium production • Transmutation or plasma implantation • May affect retention of T Roth J. et al., 2nd EFDA workshop, Cadarache, Sept. 2007 Ogordnikova O. et al., J. Nucl. Mat. 313 (2003)

  10. Tritium retention • Minimize T inventory in the co-deposits(material choice, erosion limitation, …) • Tritium removal shemes(min. interference with plasma operation & performance) • Heating in air or oxygen • Laser heating • flash lamps • He-O glow discharges • Predictions needed for licensing • Improvement in trapping/retention modeling • needs to be validated by experiments Kunz C. et al., J. Nucl. Mat. 367 (2007) • Retention diagnostics needed

  11. Other issues Dust • Major Safety issue for licensing (Be:toxic, C:tritium, W: active) • Reaction with water leakage and H production (co-deposits) • Reaction with air (vacuum leakage) Dust generation/characteristics must be understood, Diagnositcs needed (quantification), In-situ removal methods Degradation of in-vessel diagnostic components • Dust • Material deposition • Erosion • Neutron damage Avoidance (material choice, …) Limitation (removal methods, …) Shimomura Y., J. Nucl. Mat. 363 (2007)

  12. Key issues in PWI • Dedicated research facilities • Plasmatron facility VISION I • Conclusion

  13. Key issues • Synergistic effects: • plasma steady-state flux • material damage by neutrons • tritium retention • mixed materials implications • … • H interaction studies: • mainly by low flux, high energy ions • standard particle accelerators, ion beam devices, tokamaks BUT flux and energy influence the mechanism (retention, implantation, recycling) • Plasma simulators: • reach high flux, low electron temperatures

  14. Plasma simulators

  15. SCK•CEN Due to difficulties inherent to: • transport • characterization of tritium, beryllium and neutron activated materials It is advantageous to have • devices (Plasmatron VISION I) • in-house characterization tools (tritium lab., beryllium cells, hot cells) • knowledge and experience (BR2 matrix, fission, …) at the same location

  16. BR2 XPS TEM SEM Hot cell capabilities Hot cells • Beryllium cells • Tritium lab. • BR2 (high flux fission reactor) • Mechanical testing • Physico-chemical analysis • Microstructure characterization • Corrosion loops • Specimen preparation workshop Mechanical tests Corrosion

  17. Key issues in PWI • Dedicated research facilities • Plasmatron facility VISION I • Conclusion

  18. Background / History • ETHEL: the JRC experimental program (Ispra, Italy,1993) • European Tritium Handling Experimental Laboratory • Shut down ten years ago • Due to decommissioning of ETHEL buildings • Contract between SCK•CEN and JRC to transport plasmatron • Several parts of equipment is missing (were used for other projects) • Reinstallation at SCK•CEN for fusion applications • Refurbishment/recovery will be done in 2008 • Most of technical documents were found in JRC archive New nameplasmatron VISION I(Versatile Instrument for the Study of IONInteraction I)

  19. Brief plasmatron facility description • Cold self-sustained volumetric plasmaVolume: 18 litres Target diameter: ~25cm Ion energies: 20 - 500 eVMagnetic field: 0.2T Pulse duration: steady stateFlux density target: ~ 1020-1021 ions/m2.s • Designed for PWI studies • Installation for operation in glove box • A gas mixture with a certain D/T ratio can be created in a volume by measuring the pressure and the mass flow of D/T coming from volumes containing D and T. Both loops have a separate control system. Tominetti S. et al., Vuoto 26 (1997)

  20. UHV 1 I insulation C cathode A anode CW T target TC I TC temperature control CW PM T CW cooling water C C Gas inlet A PM permanent magnets I PM UHV1 main pumping CW UHV3 differential pumping Gas, plasma, secondary ions and neutrals analyser UHV 3 Plasma chamber Sedano L. et al., Phys. Stat. Sol. 188 (2001)

  21. Key issues in PWI • Dedicated research facilities • Plasmatron facility VISION I • Conclusion

  22. Conclusion • PFM requirements for ITER/DEMO • R&D programme: fabrication feasibility, resilience to neutron damage, activation, … • BUTperformance/use depends also on: T-retention, dust production, resilience to large steady-state fluences, transient loads, surface erosion, material redeposition • Key issues determined by synergistic effects(steady state flux, transient loads, neutron damage) Plasmatron VISION I can address several of these key issues because • tritium • beryllium • neutron irradiated materials can be studied under high flux densities, low plasma temperatures

  23. Thank you for your attention Any questions?

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