Key Issues in Plasma-Wall Interactions for ITER The European Approach V. Philipps, J. Roth, A. Loarte With Contributions G.F MatthewsH.G.EsserG. Federici J.P.CoadU.Samm M. Mayer J.StrachanP.WienholdP.Andrew M.StampA. KirschnerG. Pautasso M. RubelW. Jacob Key Issues in Plasma-Wall Interactions for ITER The European Approach V. Philipps, J. Roth, A. Loarte With Contributions G.F MatthewsH.G.EsserG. Federici J.P.CoadU.Samm M. Mayer J.StrachanP.WienholdP.Andrew M.StampA. KirschnerG. Pautasso M. RubelW. Jacob EU-PWI-Task Force EPS 2003, Petersburg

Magnetic Confinement Fusion is ready to build a first machine delivering 500 MW fusion power : ITER Controlled Thermonuclear Fusion has the potential to open a new primary energy source to mankind The fuel (deuterium and lithium) is cheap and worldwide accessible – This is also a contribution to a peaceful world EU-PWI-Task Force EPS 2003, Petersburg Magnetic Confinement Fusion

EU-PWI-Task Force EPS 2003, Petersburg ITER – a worldwide undertaking Europe, Russia, Japan, Canada, USA, China, S.-Korea goals 500 MW fusion power, Q=10 with burn time 7 min Quasi-steady-state plasma operation,with Q=5 and 'Hybrid' scenarios with pulse length up to 30 min Integration of physics and technology (Tritium, breeder blanket, super conductors, heating) Four ITER sites offered Four ITER candidate sites Magnetic Confinement Fusion V. Mukhovatov, I.3.3, Wed

JET has achieved simultaneously the essential dimensionless ITER parameters Confinement Pressure Density Purity Radiation Shaping Pulse duration From JET to ITER EU-PWI-Task Force EPS 2003, Petersburg JET ITER Scaling: Plasma performance JET ITER

EU-PWI-Task Force Challenge to Technology and Plasma Wall Interaction ELMs and disruptions Lifetime and T-retention From JET to ITER

EU-PWI-Task Force EPS 2003, Petersburg Control of MHD modes (NTM) high plasma pressure  -particle heating Current drive efficiencies Control of steady state heat load Control of transient heat loads (ELMs and Disruptions) Lifetime of plasma facing components Long term Tritium inventory limit (350g) ITER is an experiment to analyse these questions. Remaining crucial issues

All present data from carbon devices indicate a long term fuel (T) retention which would be unacceptable for ITER JET T experience Long term fuel (T) retention JET (T)10% TFTR(T)13% TEXTOR (D)8% Similar observations in Tore- Supra and various devices Equivalent ITER T limit (350g) would be reached in less than 50 shots 10% long term retention EU-PWI-Task Force EPS 2003, Petersburg Fuel retention: present database Of injected fuel T. Loarer P-1.161, Mon

EU-PWI-Task Force EPS 2003, Petersburg ITER 700m 2 Be first wall Low Z Oxygen getter 100m 2 Tungsten low erosion 50 m 2 Graphite CFC no melting ITER wall material Choice A European Task Force on Plasma Wall Interaction has been formed to focus the EU- PWI research on the critical questions of Tritium retention and Wall Lifetime. ITER has different first wall materials Simple extrapolation from present full carbon devices is not possible. Must be based on physics understanding.

Long term tritium retention EU PWI Task Force Strategies Understand (better) the mechanism of fuel retention in present devices Improve predictions for ITER Develop Tritium control techniques A Develop Tritium removal techniques that are applicable for ITER B Develop a full metal Tokamak scenario C EU-PWI-Task Force EPS 2003, Petersburg EU PWI Strategies EU-PWI TF structure Coordinated experiments and data analysis in JET (Task Force E & FT) and EU Fusion associations Accompanying Technology Programme Contact persons in each association

EU-PWI-Task Force EPS 2003, Petersburg Diffusion along pores Implantation (saturates) D+D+ D+D+ C Erosion area Deposition area Remote area Tritium is retained by co-deposition with carbon, on the plasma facing sides or on remote areas Understanding of T-codeposition is understanding of where and how carbon is eroded and how carbon migrates globally and locally Fuel retention: Understanding Co-ordinated research in Tokamaks and lab experiments in PWI-TF

EU-PWI-Task Force EPS 2003, Petersburg outer Erosion and Deposition in Divertor (1) P. Coad et al, PSI GIFU JET gas box, JET gas box, 5750 shots Ero-deposition (  m) JT shots inner Inner Divertor tile Ero-deposition (  m) Outer Divertor tile Y. Gotoh et al, PSI GIFU Inner Divertor is deposition dominated in all devices

The outer divertor can be erosion or deposition dominated Depending on ?  In/out asymmetry of Divertor Conditions  Differences in SOL || Flows  Influence of temperature  Divertor Geometry Adressed in PWI-TF EU-PWI-Task Force EPS 2003, Petersburg ASDEX Upgrade, PSI 2002, V. Rohde Inner Divertor Outer V.Rohde P-1.154, Mon Erosion and Deposition in Divertor (2)

Beryllium is deposited on the plasma facing areas, no transport to shadowed regions Carbon and deuterium is mainly transported to shadowed areas  Transport to remote areas is specific to carbon EU-PWI-Task Force 3 4 C: Be = 10:1 Carbon Beryllium G. Matthews P-3.198, Thurs Carbon Erosion and Deposition in Divertor (3)

EU-PWI-Task Force EPS 2003, Petersburg Local geometry determines the C-deposition on the louver entrance QMB and sticking monitors ( M. Mayer O-2.6A, Tues) show that the carbon deposition is mainly line of sight of the place of origin Quartz monitor (QMB) ,0 0,2 0,4 0,6 0,8 1,0 1,2 Configuration C-deposition (nm/s ) Erosion and Deposition in Divertor (4)

EU-PWI-Task Force EPS 2003, Petersburg 13 CH 4 tracer injection in TEXTOR LCFS 13 CH 4 Plasma P. Wienhold. A. Kirschner, PSI 2000 Modelling of erosion and redeposition A. Kirschner et al RC [cm] ZC [cm] C 0 Density With standard assumptions (2% erosion yield, „TRIM“ sticking of redeposited species): - modelled C-fluxes to the louvres much too low (JET) and - locally redeposited carbon (TEXTOR) much too low Good matching of Be transport JET MKIIA

EU-PWI-Task Force EPS 2003, Petersburg Assumptions: carbon atoms eroded in a first step can be re- eroded with higher yields after re-deposition  Enhanced movement of carbon along surfaces to shadowed areas Assumptions: carbon atoms eroded in a first step can be re- eroded with higher yields after re-deposition  Enhanced movement of carbon along surfaces to shadowed areas Trim sticking for ions,zero sticking for C x H y 8% re-erosion of re-deposited carbon species Understanding of carbon transport Physics of sticking and re-erosion is the key to understand carbon long range transport Substrate chemical erosion Yield 2 - 3% D CH 4 D CH + Shadowed areas

Graphite Tungsten Standard assumptions Carbon deposition:  5% of C-erosion flux  0.7 gT retention / ITER shot Enhanced re-erosion Carbon deposition:  14% of C-erosion flux  2 gT retention / ITER shot Eroded carbon can escape towards the dome and dome pumping ducts T-removal should be considered there EU-PWI-Task Force EPS 2003, Petersburg Modelling for ITER Divertor A. Kirschner P-3.196, Thur Modelling

EU-PWI-Task Force EPS 2003, Petersburg Erosion redeposition in divertor: summary Inner divertor deposition dominated always No unique behaviour of outer divertor Long range transport is specific of carbon Main chamber erosion dominated area in general (with local or global material redistribution) The material deposited in the divertor is mainly from main chamber erosion (mostly C at present, Be in ITER) JET: material balance, divertor Be deposition AUG: material balance and tungsten divertor experience DIII: spectroscopic analysis Main chamber ion PWI is significant and underestimated in the past

”long tails” in SOL n e & T e seen in many experiments ASDEX Upgrade TeneTene larger divertor closure  moderate decrease of neutral pressure in main chamber minimum main chamber pressure set by main chamber ion plasma wall contact main chamber contact determined largely by ELMs? J. Neuhauser et al. SOL profiles Neutral Pressure measurements EU-PWI-Task Force EPS 2003, Petersburg Main chamber Plasma Wall Interaction (1) W.Fundamenski O-4.3C, Fri A.Herrmann P-1.155, Mon B. Lipschultz P-3.197, Thurs A.Kallenbach P-1.159, Mon Main chamber Plasma Interaction is main topic in TF work

EU-PWI-Task Force EPS 2003, Petersburg Main chamber erosion Absolute main chamber PW interaction Divertor / First Wall fluxes: 12 (JET), 10 (AUG), 50 (ITER modelling) Erosion MechanismsIons versus Neutrals Erosion Location HFS versus LFS Material migration Measurements and understanding of SOL Flows Modelling based on ExB and Bxgrad B drifts underestimates measured flows Important to predict ITER outer divertor behaviour Main chamber Plasma Wall Interaction (2)

EU-PWI-Task Force EPS 2003, Petersburg The ITER Be-first wall will reduce the Carbon deposition and associated T-retention 1. No C-flux into the divertor, but a similar Be-flux [present modelling: 6 g Be/shot, better quantification needed] 2. Be is not transported to remote areas 3. Be-layers on the plasma facing sides of the divertor contain less T and are easier to access for cleaning 4. Chemical sputtering of the underlying C-substrate in the inner is reduced/suppressed T-retention: Extrapolation to ITER Be transport and influence of Be deposition on carbon erosion & transport are key questions for ITER (  PWI TF)

EU-PWI-Task Force EPS 2003, Petersburg Chemical erosion completely suppressed by adding 0.1 % Be to the plasma Questions to address Thermal stability of Be layer during transient heat pulses Be erosion-deposition in the outer divertor PISCES R. Doerner P-2.162, Tues Beryllium experiments in PISCES

EU-PWI-Task Force EPS 2003, Petersburg Fuel removal and control Control of fuel retention and fuel removal will be essential in any wall material scenario and needs more attention in present research (major topic of PWI-TF work) Fuel Removal Isotope exchangeon PFC side Thermal desorptionon PFC side Oxygen ventingremote areas? Scavenger techniques?? and Fuel Control Temperature tailoring Carbon traps Divertor geometry … Work in plasma simulators + Dedicated lab experiments + Tokamak research Needs detailed understanding of the involved physics

EU-PWI-Task Force EPS 2003, Petersburg Full metal wall: wall lifetime Steady state erosion Low erosion materials High local re-deposition Lost material replaced from main chamber D+D Plasma Divertor Strike zones If the T-retention problem cannot be solved a full metal first wall concept is needed based on metals with low hydrogen retention Plasma compatibility Wall Lifetime Transient events The main concern with metal walls is the lifetime due to melt layer erosion in transient heat loads (ELMs and Disruptions)

Particle flux Stored energy EU-PWI-Task Force Lifetime ELMs & Disruptions (1) In ELMs or disruptions part (<10%) or all of the plasma- stored-energy is lost on a short-time scale to the walls Material limits T < 2300 o (C), < 3400 o (W)   T  energy/ area/ sqrt(time) Wetted area Duration of ELMs Energy on target plates  20 (40) MJ m -2 s -1/2

Divertor wetted area during ELM similar to between ELMs some ELM energy can reach the main chamber T. Eich, A. Herrmann, this meeting EU-PWI-Task Force Duration of Divertor ELM Energy Pulse well correlated with II B Ion Transport   220  s for ITER Elm / in between Elm Lifetime ELMs & Disruptions (2)

EU-PWI-Task Force EPS 2003, Petersburg number of ELMs for 1000 ITER pulses  4 x 10 5 Predictions for ITER divertor target lifetime ITER Predictions ELMs are marginally acceptable but behaviour for CFC and W not much different at moderate power densities Lifetime: ELMS & Disruptions (3) G. Federici et al.

Disruption Energy deposition (Thermal Quench) occurs over a large divertor area (AUG & JET) Fraction of energy deposited in Divertor: 1 (AUG), ~ 0.2 (JET) ASDEX Upgrade Poloidal Distance (m) V. Riccardo, I.3.5. Wed G. Pautasso P-1.135, Mon P. Andrew P-1.54, Mon EU-PWI-Task Force EPS 2003, Petersburg Disruptions: Present ITER specifications All disruption energy is lost at the strike zones with narrow deposition (broadening of 3) half of the melt layer is lost per event Lifetime: ELMS &Disruptions (4)

EU-PWI-Task Force Disruption energy density, MJ/m 2 SOL broadening, lambda plate /lambda mid Current ITER specifications G. Federici, ITER JWS Garching, 7 Oct Absence of melting for W lambda mid =5 mm Disruption energy density (MJ) Fraction of energy to divertor Spatial and Time Evolution of thermal Energy Flux Dependence on Disruption Type Energy Balance during Current Quench (Halo Currents) PWI TF issues Disruptions energy deposition and mitigation are key issues for the ITER material selection SOL Broadening Lifetime: ELMS &Disruptions (5) Development of disruption mitigation techniques Power exhaust on irregular, molten surfaces

EU-PWI-Task Force Conclusions (1) Predictions for long term tritium retention for ITER are critical, but are from full carbon machines. An integrated approach and understanding of Where and how impurities are produced in the main chamber How they are transported towards the divertor How the material is transported inside the divertor Influence of Be deposition on carbon erosion and transport is necessary to predict the T retention under the ITER wall material conditions.

EU-PWI-Task Force Conclusions (2) In parallel co-ordinated work is necessary on In situ control of T retention Fuel (T) removal which can be employed under ITER conditions Disruption power deposition characteristics Disruption mitigation More Tokamak experience is needed for metal wall conditions High Z first wall, tokamak behaviour under non- carbon wall conditions Be first wall with carbon and tungsten in the divertor R.Neu P-1.123, Mon

EU-PWI-Task Force Additional remarks A graphite and a tungsten divertor should be prepared in parallel. Possibilities to measure the fuel retention in the non activated phase of ITER are needed. The possibility to change ITER in a later state from a low to a high Z first wall should be evaluated. ITER needs flexibility to adopt Tritium control and removal techniques that have to be developed in parallel and tested in present devices.