Material erosion and migration in tokamaks Centre de Recherches en Physique des Plasmas Material erosion and migration in tokamaks R. A. Pitts CRPP, Association-EURATOM Confédération Suisse, EPFL Lausanne, Switzerland with many thanks for contributions from N. Asakura1, S. Brezinsek2, C. Brosset3, J. P. Coad4, D. Coster5, E. Dufour3, G. Federici6, R. Felton4, M. E. Fenstermacher7, R. S. Granetz8, A. Herrmann5, J. Horacek, A. Kirschner2, K. Krieger5, A. Loarte9, J.Likonen10, B. Lipschultz8, A. Kukushkin6, G. F. Matthews4, M. Mayer5, R. Neu5, J. Pamela11, B. Pégourié3, V. Philipps2, J. Roth5, M. Rubel12, L. L. Snead13, P. C. Stangeby14, J.D. Strachan15, E. Tsitrone3, W. Wampler16, D. Whyte17 1JAERI, 2FZJ-Jülich, 3CEA Cadarache, 4UKAEA, 5IPP Garching, 6ITER, 7LLNL, 8PSFC-MIT, 9EFDA CSU Garching, 10VTT-TEKES, 11EFDA CSU Culham, 12Alfvén Lab. RIT, 13ORNL, 14UTIAS, 15PPPL, 16SNL,17Univ. Wisconsin,

Outline of the talk Introduction The components of migration Global migration accounting Material choices for the next step Conclusions

What is migration? Migration = Erosion Transport Deposition Re-erosion Not an operational issue in today’s tokamaks, but certainly will be in ITER and beyond ……

Migration will be important Co-deposition High erosion rates and long term migration of carbon yield high levels of Tritium retention ITER: ~50 g T per pulse 0.01-0.2 g per pulse now ITER operation suspended once 350 g T accumulated Could be fewer than ~100 pulses No proven T clean-up technology Material mixing, properties Formation of compounds and alloys through the interaction of pure materials Change of material properties Be on W forms BeW alloys already at ~800°C Surface melting point could be ~2000°C lower than for pure W

Where do erosion and migration occur? JET #62218: plasma visible light emission Limited t = 3.0 s Diverted t = 12.0 s At specific structures to protect the vacuum vessel walls or isolate the plasma-surface interaction

Poloidal cross-section Some terminology Scrape-off layer (SOL) Cool plasma on open field lines SOL width ~1 cm ( B) Length usually 10’s m (|| B) Poloidal cross-section Core plasma Divertor Plasma guided along field lines to targets remote from core plasma: low T and high n Separatrix Private flux region Inner Outer ITER will be a divertor tokamak Divertor targets

Materials in today’s tokamaks Low Z (Carbon) High Z Divertor TCV, MAST, NSTX, DIII-D, JT-60U, JET AUG (C+W) C-Mod (Mo) Limiter TEXTOR, Tore Supra FTU (Mo) The majority of today’s medium to large size tokamaks favour Carbon  extensive operational experience No melting / low core radiation / high edge radiation But T-retention problem and high erosion rates of low Z mean that high Z may be the only long term solution Living with W: see Kallenbach, I3.004, Wed.

Migration = Erosion Transport Deposition Re-erosion

Principal erosion mechanisms Sputtering Ions and neutrals Physical and chemical (for carbon) Macroscopic - transients Melt layer losses Evaporation, sublimation Not generally observed in present experiments – currently the main reason for Carbon being used in the ITER divertor Arcing, Dust (see Krasheninnikov et al, P4.019, Thurs.)

Physical and chemical sputtering Chemical (carbon) D impact Eckstein et al. Roth et al., NF 44 (2004) L21 ITER divertor flux Energy threshold  higher for higher Z substrate Much higher yields for high Z projectiles No threshold Dependent on bombarding energy, flux and surface temperature  More optimistic prediction for ITER

ELMs: an example of transient erosion Da H-mode  Edge MHD instabilities  Periodic bursts of particles and energy into the SOL. Type I ELMing H-mode is baseline ITER scenario Time (s) JET #62218 t = 19.05 s, ELM-free t = 19.06 s, Type I ELM For more on the physics of ELMs, See Huysmans, I4.002 Thurs.

ELMs can ablate Carbon on JET 1 MJ ELM  ~0.2 MJm-2 on the divertor target Peak Tsurf ~ 2500ºC ELM-free 19.73 s Radiated Power 1.0 MJ ELM 19.79 s Range of energies expected per Type I ELM in ITER ~ 0.6  3.5 MJm-2 Loarte et al, Phys. Plasmas 11 (2004) 2668

ELM ablation limits ITER divertor lifetime Acceptable lifetime before target change required: 3000 full power shots  ~1 x 106 ELMs Inter-ELM power: 5 MWm-2 Target thickness: CFC: 20 mm W: 10 mm No redeposition of ablated material No W melt layer loss Federici et al, PPCF 45 (2003) 1523 CFC ITER min. requirement W Minimum ITER ELM size Both low and high Z target materials marginal on present scalings Significant effort in the community towards ELM mitigation

Migration = Erosion Transport Deposition Re-erosion

Transport creates & moves impurities EDGE2D/NIMBUS Bypass leaks Escape via divertor plasma Ionisation Gas puff CX event Ions: Cross-field transport – high ion fluxes can extend into far SOL  recycled neutrals  direct impurity release ELMs ….. Eroded Impurity ions “leak” out of the divertor (T forces) SOL and divertor ion fluid flows – can entrain impurities Neutrals: From divertor plasma leakage, gas puffs, bypass leaks  low energy CX fluxes  wall sputtering Lower fluxes of energetic D0 from deeper in the core plasma

Experimentally, strong SOL flows D-flows: parallel Mach Number, M = v||/cs. POSITIVE towards inner target M M JET (Tore-Supra) C-Mod JT-60U Bj ● (TCV) N. Asakura, NF 44 (2004) 503 B. LaBombard, NF 44 (2004) 1047 S. K. Erents, PPCF 46 (2004) 1757 M JT-60U M M JT-60U Distance to separatrix (mm) Distance to separatrix (mm) See LaBombard, I3.007 Wed., Bonnin, P2.110, today

Using tracers to study the transport 13CH4 markers are being increasingly used to get a handle on migration  gas puff just before vent and tile retrieval – pioneered on TEXTOR 2.8g 13C, ohmic 0.2g 13C, L-mode 0.2g 13C, H-mode JET DIII-D AUG 0.0025g 13C H-mode 9.3g 13C H-mode

Top injection: C13  inner target Likonen et al, Fus. Eng. Design 66-68 (2003) 219 Wampler et al, JNM 337-339 (2005) 134 JET Start End DIII-D Simple conditions: ohmic, L-mode, no ELMs DIII-D: toroidally symmetric injection, JET: toroidally localised Data and modelling demonstrate fast flow to inner divertor Situation more complex in H-mode and other injection points For more on JET C13 expts. see Rubel, P2.004, today

Migration = Erosion Transport Deposition Re-erosion

Deposition sensitive to local conditions DIII-D Outer divertor usually hotter  favours C erosion (phys. + chem.) Inner divertor usually colder  favours C deposition (chem. only) C transport by SOL flows Similar picture on most other carbon machines Observations consistent with a contribution to the carbon source from outside the divertor Detached Groth et al., P4.015, Thurs. Whyte et al., NF 41 (2001) 1243, NF 39 (1999) 1025

Re-erosion important for C-migration Esser et al., JNM 337-339 (2005) 84 JET 57080 57082 57084 57086 0.0 0.2 0.4 0.6 0.8 1.0 1.2 Shot number L-mode Quartz Micro-Balance (QMB) C-deposition (nm/s) ERO code Strike point Reproduced by transport modelling Large increase on baseplate requires enhanced C re-erosion Chemical erosion Migration to remote areas due to magnetic and divertor geometry Kirschner et al, JNM 337-339 (2005) 17

Global migration accounting = Erosion Transport Deposition Re-erosion

A non-trivial task! Tore Supra Spectroscopic methods in plasma, post-mortem surface analysis and just plain old scraping and sweeping up  extremely rigorous balance achieved first on TEXTOR (Wienhold et al., JNM 313-316 (2003) 311) Tore Supra balance: see Dufour et al, P5.002 Friday

JET migration accounting (I) Use spectroscopic methods + modelling to compute C sources EDGE2D/NIMBUS DIVIMP/OSM Simulation of CIII emission  intrinsic sources 1 ton/year Divertor C-source = 5-10 x Wall source Strachan et al, NF 43 (2003) 922 Carbon recycles

JET migration accounting (II) Make balance for period 1999-2001 with MarkII GasBox divertor: 14 hours plasma in diverted phase (50400 s, 5748 shots) Spectroscopy + Modelling Post mortem surface analysis Deposition all at inner target Net erosion at main walls No significant divertor erosion 450g C (CIII) 215 kg/year  strong T co-deposition (1 year = 3.2 x 107 secs) ~400g C Very similar result for AUG, but overall C-balance more complex Mayer et al, JNM 337-339 (2005) 119 Likonen et al, JNM 337-339 (2005) 60, Matthews et al., EPS 2003

Tungsten migration in AUG 2002-2003 Campaign: ~1.4 hours in diverted phase (4680 s, 1205 shots) Post mortem surface analysis: Only ~12% of inboard W source deposited in divertor ~ few % to upper divertor and other main chamber surfaces W-coated: (40% of total area) 1.3x1018s-1 W erosion not balanced by non-local deposition – most is promptly redeposited  simpler than C picture Larger Larmor radius helps at higher mass 0.5x1017s-1 ~1.5 kg/year W+ W0 1.1x1017s-1 Krieger et al, JNM 337-339 (2005) 10

Material choices for the next step An ITER-like first wall at JET

Current materials choice for ITER Be for the first wall Low T-retention Low Z Good oxygen getter W C for the targets Low Z Does not melt 350 MJ stored energy W for the baffles High threshold for CX neutral sputtering CFC Fallback option Be wall, all-W divertor Castellations for stress relief  co-deposition in gaps? Driven by the need for operational flexibility

An ITER-like wall in JET Option 1 Option 2 Be Option 1 or 2 to be chosen in 2006: Objectives Demonstrate low T-retention Study melt layer loss (walls and divertor)  ELMs and disruptions Study effect of Be on W erosion Be and W migration Demonstrate operation without C radiation Refine control/mitigation techniques  ELMs and disruptions Demonstrate routine / safe operation of fully integrated ITER compatible scenarios at 3-5MA  Power upgrade to 40-45 MW  Experiments from 2009 onwards

 Full wall materials tests in current machines Conclusions Erosion and migration: Complex materials and physics Not an operational issue now But will be in ITER and beyond Optimisation of core plasma performance and wall lifetime cannot be decoupled Refine predictive capability Still significant uncertainties …….  Full wall materials tests in current machines

Reserve slides

ELMs might also erode the main walls A. Herrmann, AUG Main chamber thermography on AUG Type I ELMs: ~25% of stored energy drop deposited on non- divertor components ELM ion energies measured at JET walls agree with recent theory Suggests: Eion > 1 keV on ITER  erosion problem, even for high Z wall Herrmann et al, P1.006 Mon.

Can SOL ion flows transport material? Yes, but picture is complex – theory and experiment not yet reconciled ErxB, pxB EqxB Poloidal Pfirsch-Schlüter Bj Ballooning Divertor sink Parallel Simplified – shown in the poloidal plane only

Carbon balance: TEXTOR, Tore Supra Carbon Sources (g/h) von Seggern et al, Mayer et al., Phys. Scripta T111 (2004) Wienhold et al, von. Seggern et al., JNM 313-316 (2003) Brosset et al., JNM 337-339 (2005) 311, E. Tsitrone et al., IAEA 2004 TEXTOR TS Toroidal limiters: 22 7 Carbon Sinks (g/h) Toroidal limiters: 10 1 “Obstacles”: 6 0.5 Low sticking – also AUG Bumper: 1 ? Neutralisers: 1 1-2 Very good balance considering the scope for error TEXTOR deposition extrapolates to ~220 kg/year of plasma Tore-Supra balance still preliminary Pump ducts: 0.02 ? Pumped out: 1-2 0.2-2 Total: 19-20 2.7-5.5 Dufour et al, P5.002 Friday

Similar observations at JET Net inner divertor deposition and little net erosion in outer divertor implies net wall source Macroscopic flakes in regions not generally visible to plasma  migration to remote areas  high levels of T-retention Flakes Coad et al., JNM 313-316 (2003) 419

JET migration accounting (II) Make balance for period 1999-2001 with MarkII GasBox divertor: 16 hours plasma Spectroscopy + Modelling Post mortem surface analysis Deposition all at inner divertor Surface layers are Be rich  C chemically eroded and migrates, Be stays put 20g Be (BeII) 450g C (CIII) 215 kg/year  strong T co-deposition ~400g C 22g Be Very similar result for AUG, but overall C-balance more complex Mayer et al, JNM 337-339 (2005) 119 Likonen et al, JNM 337-339 (2005) 60, Matthews et al., EPS 2003