Introduction of The 2nd International Symposium on Lithium Applications for Fusion Devices April , 2011 Princeton, New Jersey, USA Jiansheng Hu

参加人员 61 人， – USA(39 ， 24 from PPPL) – Spain(2) – Italy(6) – China(1) – Russia(4) – Japan(2) – Others(7)

Sessions Lithium in Magnetic Confinement Experiments Overview Talks – NSTX Plasma Operation with a Liquid Lithium Divertor – New progresses of lithium coating or plasma facing material in ASIPP(EAST, HT-7) – Recent Results from the Lithium Tokamak experiment (LTX) – Li collection experiments on T-11M and T-10 in framework of Li closed loop concept – Plasma behavior in presence of a liquid lithium limiter(FTU) – Recycling and Sputtering Studies in Hydrogen and Helium Plasmas under Lithiated Walls in TJ-II – Lithization on RFX-mod reversed field pinch experiment Lithium in Magnetic Confinement Topical Experiments Special Liquid Lithium Technology Session Lithium Laboratory Test Stands Lithium Theory/modeling/comments Innovative Lithium Applications

NSTX Plasma Operation with a Liquid Lithium Divertor NSTX Solid Lithium Surfaces Must Be Renewed Between Discharges Liquid Lithium Divertor (LLD) Installed in NSTX with Porous Molybdenum Face to Hold Lithium Solid Li Coating Reduces D-recycling, Reduces H-mode Power Threshold,Broadens Te Profiles, Decreases Electron Thermal Diffusivity,Suppresses ELMs, Improves Confinement

LLD operation successfully demonstrated with strike point on lithium-filled surface (required for NSTX-U to have full power, long pulse Li PFC conditions)" – Long term thermal response dominated by thermal mass of the copper" – No macroscopic evidence of surface damage by lithium or heating" LLD in its effect on plasma performance did not clearly differ from evaporative lithium coatings" – LLD static lithium surface exhibited a degradation due to D and impurity – LLD temperatures exceeded melting point of lithium but lithium compounds from impurities remained undissolved on surface For the first time, sufficient lithium was deposited, sufficiently fast, to allow at least 30 discharges on the LLD, and then 150 discharges on the lithiated graphite without LITER needed between discharges (can be extended for longer pulses) Issues of lithium vacuum chemistry need investigation for both static liquid lithium analysis, and the design of flowing lithium system for NSTX-U.

Lithium in NSTX proved to be an exceptionally powerful tool for H-mode plasma performance: Global confinement improved through electron confinement improvement by ~ 20 – 30% with strong lithium pumping. Contributed to the highest confinement H-mode with H98y2 < 1.7. H-mode power threshold significantly reduced by ~ 20 – 30%. Completely stabilized ELMs. Very little core lithium contamination (< 1%) found. Improved HHFW and EBW (RFs ) performance by controlling edge density. Contributed to the non-inductive CHI start-up success by controlling impurities. Improved plasma shot reliability: shots / week increased ~ 40% over pre-lithium by controlling impurities.

NSTX upgrade NSTXNSTX Upgrade Fusion Nuclear Science Facility Aspect Ratio = R 0 / a  1.3  1.5 Plasma Current (MA)124  10 Toroidal Field (T) P/R, P/S (MW/m,m 2 )10, 0.2*20, 0.4*30  60, 0.6  1.2 Molybdenum tiles on inboard divertor -Replace 48 second row tiles with 1” moly tiles -Includes three tiles with embedded diagnostics -Lithium coating with LITER ~2 x outer LLD rate -Plasma heating can liquify lithium surface

Recent results from LTX and near-term plans LTX has a full, 5 m2 heated, conformal wall(400 C now, 600 C planed) Surface coated by evaporation.

Li collection experiments on T-11M and T-10 in framework of Li closed loop concept Li closed loop concept +CPS T-10 T-11M

Plasma behaviour in presence of a liquid lithium limiter （ FTU ） 1.Better plasma performance with Lithium than boronization 2.Radiation losses are very low less than 30% 3.With lithium limiter much more gas has to be injected to get the same electron density with respect to boronized and fully metallic discharges > 10 times 4.Operations near or beyond the Greenwald limit are easily performed 5.For q>5 the Greenwald limit has been exceed by more than a factor 1.5 at I p =.5 MA B t =6T (n e = m -3 ) and n G >1.3 at I p =.7MA B t =7.2T Bt=6T (n e = m -3 ) 6.T e in the SOL is 50% higher while increase in n e is much smaller in lithium discharges 7.Operations are generally more easy to perform and the behavior of the machine is more reliable. 8.Discharge recovery after a disruption is prompt Liquid lithium surface Heater Li source S.S. box with a cylindrical support Mo heater accumulator Ceramic break Thermocouples 100 mm34 mm

RECYCLING AND SPUTTERING STUDIES IN HYDROGEN AND HELIUM PLASMAS UNDER LITHIATED WALLS IN TJ-II More than three years of operation of TJ-II under Li walls : ~10000 shots. Main results: - Low recycling wall for H (10%) and He (82%) at RT - Highly improved density control - Routine operation under 2 NBI (800 kW) heated plasmas -Transition to H mode - Improved Confinement/E content - Development of peaked profiles - Decreased Li sputtering yield

Lithization on RFX-mod reversed field pinch experiment Lithization has been tested on RFX-mod by Li pellet – Thin Li wall coating provides clear effect on recycling and edge transport – Lithium conditioning disappear very rapidly – Loosing of conditioning ability is not related to ‘true’ Lithium removal Liquid Lithium Limiter has been also tested – Not an easily technique on RFPs with a carbon wall – Easily used as evaporator but with high toroidal asymmetry on deposition Post-lithization effects – Lithium is present in graphite when no effects are observed on plasma – After vessel-venting impurity and density control deny good operation – Good performance are recover by medium current shots and 30’-60’ of He-GDC in between Future experiments – New Lithization tests are planed this year on RFX-mod with LLL used first as evaporator and then as limiter – To improve toroidal deposition symmetry multi-evaporator or centrifugal pellet injector are also foreseen

Lithium in Magnetic Confinement Topical Experiments V. A. Soukhanovskii: Recycling, Pumping and Divertor Plasma-Material Interactions with evaporated lithium coatings in NSTX M. A. Jaworski: Modification of the Electron Energy Distribution Function during Lithium Experiments on the National Spherical Torus Experiment J. Kallman: Determination of Effective Sheath Heat Transmission Coefficient in NSTX Discharges with Applied Lithium Coatings A.G. McLean: Liquid Lithium Divertor surface temperature dynamics and edge plasma modification under plasma-induced heating and lithium pre-heating R. Nygren: Thermal Modeling of the Surface Temperatures on the Liquid Lithium Divertor in NSTX F. Scotti: Surface reflectivity and carbon source studies with the Liquid Lithium Divertor in NSTX R. Maingi: Effect of Lithium Coatings on Edge Plasma Profiles, Transport, and ELM Stability in NSTX V. Surla: Characterization of transient particle loads during lithium experiments on the National Spherical Torus Experiment D. Frigione: High Density and Pellet Injection Experiments with Lithium Coated Wall on FTU Tokamak A. V. Vertkov: Status and prospect for the development of Liquid Lithium Limiters for Stellarotor TJ-II E. Granstedt : Effect of Lithium Wall Conditioning and Impurities in LTX D.P. Lundberg : Fueling of LTX Plasmas with Lithium Plasma Facing Components C.H. Skinner: Plasma facing surface composition during Li evaporation on NSTX and LTX

Special Liquid Lithium Technology Session M. Abdou: Summary of current R&D efforts for liquid metal based blankets and ITER TBM M. Kondo: Improvement of compatibility of liquid metals Li and Pb-17LI Y. Hirooka: Cluster/Aerosol Formation and Hydrogen Co-deposition by Colliding Ablation Plasma Plumes of Lithium and Lead M. Kondo: Hydrogen transports at interface between gas bubbling and liquid breeders F. Groeschel : The IFMIF Target Facility engineering design and the validation of key issues within the IFMIF-EVEDA Project G.Miccichè: Status of the activities for the development of the remote handling techniques for the maintenance of IFMIF target assembly system D. Bernardi: IFMIF Lithium Target I. Lyublinski: Module of Lithium Divertor for KTM Tokamak M. Narula: Fast flowing liquid lithium divertor concept for NSTX

Module of Lithium Divertor for KTM Tokamak Flowing liquid limiter+CPS Lots of section in radical direction

Lithium Laboratory Test Stands J.P. Allain: Lithium-based surfaces controlling Fusion plasma behavior at the plasmamaterial interface C.N. Taylor: Deciphering energetic deuterium ion interactions with lithiated ATJ graphite T.Abrams: Investigation of LLD Test Sample Performance Under High Heat Loads V.Yu. Sergeev: Lithium technologies for edge plasma control A.B. Martín: Electrical characteristics of lithium surfaces exposed to a plasma B. Rais: Lithium particle detector for fusion applications. S. Jung: Laboratory Investigation of an Effect of Lithium on ICRF Antenna in DEVeX N.R.Murray: Capillary Wicking of Lithium on Laser--‐Textured Surfaces

Lithium Theory/modeling/comments P. S. Krstic : Dynamics of deuterium retention and sputtering of Li-C-O surfaces J.N. Brooks : Modeling of plasma/lithium-surface interactions in NSTX: status and key issues” M Romanelli: Turbulent Transport in Lithium Doped Fusion Plasmas C.S. Chang: Kinetic understanding of Neoclassical Lithium Transport R.D. Smirnov: Modeling of lithium dust injection and wall conditioning effects on edge plasmas with DUSTT/UEDGE code M. Ono: Recent progress of NSTX lithium research and opportunities for magnetic fusion research

Innovative Lithium Applications: I. Tazhibayeva: Study of Processes of Hydrogen Isotope Interaction with Lithium CPS D. Ruzic: Lithium / Molybdenum Infused Trenches (LiMIT): A heat removal concept for the NSTX inner divertor L. E. Zakharov: Design guidance for the flowing lithium systems in tokamaks D. K. Mansfield: Pacing Small ELMs at High Frequency using Spherical Lithium Granules and a Dropper / Impeller Injection Technology D. Andruczyk: Electrostatic Lithium Injector (ELI) R. Goldston: Draft Mission and Specifications for an Integrated PMI-PFC Test Stand Y.M. Goh: Concept Development and Engineering Considerations of a Steady-State Lithium-Coated Divertor S.W. Brown: 6Li – An Enabling Material for Fusion Abraham Sternlieb: Making turn toward fusion development

Discussion “Is lithium PFC viable in magnetic fusion reactors such as ITER?”  Handling Divertor High Heat Flux  Removal of D, T, Impurities  Divertor Steady-State Heat Removal  Liquid Lithium Flow in Magnetic Field  Long Term Lithium Corrosion  Flowing Liquid Lithium Safety  Compatibility with Hot Wall

Next meeting Rome, Italy