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 Plasmatron facility VISION I Conclusion

ITER ITER like wall JET (2 options) PPCS (inspiration for DEMO) W alloy ~ 540/720 He ~ 600/990~ 540/ /167Coolant I/O T(°C) LiPbHe H2OH2OCoolant Armour material SiC f /SiCW alloy CuCrZrStructural material Model D or Self-cooled Model C or Dual-Coola. Model B or HCPB Model A or WCLL Model AB or HCLL TUNGSTEN Plasma facing materials

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

Mixed materials Stable Be-W alloys Stable Be-W intermetallics are:  ~2200°C (Be 2 W)  ~1500°C (Be 12 W)  ~1300°C (Be 22 W) melting points closer to Be than to W! Baldwin M. et al., J. Nucl. Mat. 363 (2007) What happens if Be transport into the W bulk is rapid enough that alloy formation is not limited to the near surface?

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

C W 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 (2003) Tungsten LZLZ BUT High radiative cooling rate (compared to C)

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

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

Tritium retention Kunz C. et al., J. Nucl. Mat. 367 (2007) 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 Retention diagnostics needed Minimize T inventory in the co-deposits (material choice, erosion limitation, …)

Other issues Shimomura Y., J. Nucl. Mat. 363 (2007) Degradation of in-vessel diagnostic components  Dust  Material deposition  Erosion  Neutron damage Avoidance (material choice, …) Limitation (removal methods, …) 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

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

Synergistic effects:  plasma steady-state flux  material damage by neutrons  tritium retention  mixed materials implications  … Key issues Plasma simulators:  reach high flux, low electron temperatures 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

SCKCEN 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

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

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

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 SCKCEN and JRC to transport plasmatron  Several parts of equipment is missing (were used for other projects) Reinstallation at SCKCEN for fusion applications  Refurbishment/recovery will be done in 2008  Most of technical documents were found in JRC archive Background / History New name plasmatron VISION I (Versatile Instrument for the Study of ION Interaction I)

Cold self-sustained volumetric plasma Volume: 18 litres Target diameter: ~25cm Ion energies: eV Magnetic field: 0.2T Pulse duration: steady state Flux density target: ~ ions/m 2.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. Brief plasmatron facility description Tominetti S. et al., Vuoto 26 (1997)

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

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

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 PFM requirements for ITER/DEMO R&D programme: fabrication feasibility, resilience to neutron damage, activation, … BUT performance/use depends also on: T-retention, dust production, resilience to large steady-state fluences, transient loads, surface erosion, material redeposition

Thank you for your attention Any questions?