Introduction of

The 2nd International Symposium onLithium Applications for Fusion Devices

April 27 - 29, 2011Princeton, New Jersey, USA

Jiansheng Hu2011.05.03

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.

• 2010 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

NSTX NSTX Upgrade Fusion Nuclear Science Facility

Aspect Ratio = R0 / a 1.3 1.5 1.5

Plasma Current (MA) 1 2 4 10Toroidal Field (T) 0.5 1 2-3P/R, P/S (MW/m,m2) 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 Ip=.5 MA Bt=6T (ne=3.2 1020 m-3) and nG>1.3 at Ip=.7MA Bt=7.2T Bt=6T (ne=4 1020 m-3)

6. Te in the SOL is 50% higher while increase in ne 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 surfaceHeater

Li sourceS.S. box with a cylindrical support

Mo heater accumulator

Ceramic break

Thermocouples

100 mm 34 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 10-20 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 NSTXR. 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 ExperimentD. 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 LTXD.P. Lundberg : Fueling of LTX Plasmas with Lithium Plasma Facing ComponentsC.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-17LIY. Hirooka: Cluster/Aerosol Formation and Hydrogen Co-deposition by

Colliding Ablation Plasma Plumes of Lithium and LeadM. Kondo: Hydrogen transports at interface between gas bubbling and liquid

breedersF. Groeschel : The IFMIF Target Facility engineering design and the validation

of key issues within the IFMIF-EVEDA ProjectG.Miccichè: Status of the activities for the development of the remote

handling techniques for the maintenance of IFMIF target assembly systemD. Bernardi: IFMIF Lithium TargetI. Lyublinski: Module of Lithium Divertor for KTM TokamakM. 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 StandsJ.P. Allain: Lithium-based surfaces controlling Fusion plasma behavior

at the plasmamaterial interfaceC.N. Taylor: Deciphering energetic deuterium ion interactions with

lithiated ATJ graphiteT.Abrams: Investigation of LLD Test Sample Performance Under High

Heat LoadsV.Yu. Sergeev: Lithium technologies for edge plasma controlA.B. Martín: Electrical characteristics of lithium surfaces exposed to a

plasmaB. 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 CPSD. Ruzic: Lithium / Molybdenum Infused Trenches (LiMIT): A heat removal concept forthe NSTX inner divertorL. E. Zakharov: Design guidance for the flowing lithium systems in tokamaksD. K. Mansfield: Pacing Small ELMs at High Frequency using Spherical Lithium Granulesand a Dropper / Impeller Injection TechnologyD. Andruczyk: Electrostatic Lithium Injector (ELI)R. Goldston: Draft Mission and Specifications for an Integrated PMI-PFC Test StandY.M. Goh: Concept Development and Engineering Considerations of a Steady-StateLithium-Coated DivertorS.W. Brown: 6Li – An Enabling Material for FusionAbraham 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