1 / 32

Key Issues in Plasma-Wall Interactions for ITER The European Approach 

EU-PWI-Task Force. Key Issues in Plasma-Wall Interactions for ITER The European Approach  V. Philipps, J. Roth, A. Loarte  With Contributions G.F Matthews H.G.Esser G. Federici J.P.Coad U.Samm M. Mayer J.Strachan P.Wienhold P.Andrew M.Stamp A. Kirschner G. Pautasso

doctor
Download Presentation

Key Issues in Plasma-Wall Interactions for ITER The European Approach 

An Image/Link below is provided (as is) to download presentation Download Policy: Content on the Website is provided to you AS IS for your information and personal use and may not be sold / licensed / shared on other websites without getting consent from its author. Content is provided to you AS IS for your information and personal use only. Download presentation by click this link. While downloading, if for some reason you are not able to download a presentation, the publisher may have deleted the file from their server. During download, if you can't get a presentation, the file might be deleted by the publisher.

E N D

Presentation Transcript

  1. EU-PWI-Task Force Key Issues in Plasma-Wall Interactions for ITER The European Approach  V. Philipps, J. Roth, A. Loarte  With Contributions G.F Matthews H.G.Esser G. Federici J.P.Coad U.Samm M. Mayer J.Strachan P.Wienhold P.Andrew M.Stamp A. Kirschner G. Pautasso M. Rubel W. Jacob EPS 2003, Petersburg

  2. EU-PWI-Task Force Magnetic Confinement Fusion 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 Magnetic Confinement Fusion is ready to build a first machine delivering 500 MW fusion power :ITER EPS 2003, Petersburg

  3. EU-PWI-Task Force Magnetic Confinement Fusion 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,withQ=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 V. Mukhovatov, I.3.3, Wed EPS 2003, Petersburg

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

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

  6. EU-PWI-Task Force Remaining crucial issues • 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. EPS 2003, Petersburg

  7. EU-PWI-Task Force 10% long term retention Fuel retention: present database 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 Of injected fuel T. Loarer P-1.161, Mon • All present data from carbon devices indicate a long term fuel (T) retention which would be unacceptable for ITER EPS 2003, Petersburg

  8. EU-PWI-Task Force ITER has different first wall materials Simple extrapolation from present full carbon devices is not possible. Must be based on physics understanding. ITER ITER wall material Choice 700m2 Be first wall Low Z Oxygen getter 100m2 Tungsten low erosion 50 m2 Graphite CFC no melting 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. EPS 2003, Petersburg

  9. EU-PWI-Task Force B C A • Understand (better) the mechanism of fuel retention in present devices • Improve predictions for ITER • Develop Tritium control techniques Develop a full metal Tokamak scenario Develop Tritium removal techniques that are applicable for ITER Long term tritium retention EU PWI Strategies EU PWI Task Force 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 EPS 2003, Petersburg

  10. EU-PWI-Task Force Remote area D+ D+ C Implantation (saturates) Diffusion along pores Erosion area Deposition area EPS 2003, Petersburg Fuel retention: Understanding • 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 migratesglobally and locally Co-ordinated research in Tokamaks and lab experiments in PWI-TF

  11. EU-PWI-Task Force JT-60 4300 shots Ero-deposition (m) Inner Divertor tile Outer Divertor tile EPS 2003, Petersburg Erosion and Deposition in Divertor (1) JET gas box, 5750 shots P. Coad et al, PSI GIFU inner Ero-deposition (m) outer Y. Gotoh et al , PSI GIFU Inner Divertor is deposition dominated in all devices

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

  13. Erosion and Deposition in Divertor (3) EU-PWI-Task Force C: Be = 10:1 1 5 3 3 4 6 4 Carbon Carbon Beryllium G. Matthews P-3.198, Thurs • Beryllium is deposited on the plasma facing areas, no transport to shadowed regions • Carbon and deuteriumis mainly transported to shadowed areas •  Transport to remote areas is specific to carbon

  14. 3 EU-PWI-Task Force 1,2 3 1 2 3 1,0 0,8 C-deposition (nm/s ) 0,6 2 2 0,4 0,2 1 0,0 Configuration Quartz monitor (QMB) Erosion and Deposition in Divertor (4) 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 EPS 2003, Petersburg

  15. 13CH4 tracer injection in TEXTOR EU-PWI-Task Force Plasma ZC [cm] 13CH4 LCFS C0 Density -130 -140 -150 -160 -170 -180 220 260 230 240 250 RC [cm] Modelling of erosion and redeposition P. Wienhold. A. Kirschner, PSI 2000 A. Kirschner et al JET MKIIA • 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 EPS 2003, Petersburg

  16. EU-PWI-Task Force Shadowedareas CH+ D CH4 CH4 D Understanding of carbon transport 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 Substrate chemical erosion Yield 2 - 3% Trim sticking for ions, zero sticking for CxHy 8% re-erosion of re-deposited carbon species Physics of sticking and re-erosion is the key to understand carbon long range transport EPS 2003, Petersburg

  17. EU-PWI-Task Force Modelling for ITER Divertor Eroded carbon can escape towards the dome and dome pumping ducts T-removal should be considered there Tungsten Graphite Modelling 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 A. Kirschner P-3.196, Thur EPS 2003, Petersburg

  18. EU-PWI-Task Force 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 EPS 2003, Petersburg

  19. EU-PWI-Task Force ASDEX Upgrade Te ne 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 Wall Interaction (1) Neutral Pressure measurements SOL profiles • 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. ”long tails” in SOL ne & Te seen in many experiments Main chamber Plasma Interaction is main topic in TF work EPS 2003, Petersburg

  20. EU-PWI-Task Force Main chamber Plasma Wall Interaction (2) • Main chamber erosion • Absolute main chamber PW interaction • Divertor / First Wall fluxes: 12 (JET), 10 (AUG), 50 (ITER modelling) • Erosion Mechanisms Ions versus Neutrals • Erosion LocationHFS 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 EPS 2003, Petersburg

  21. EU-PWI-Task Force T-retention: Extrapolation to ITER 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 Be transport and influence of Be deposition on carbon erosion & transport are key questions for ITER ( PWI TF) EPS 2003, Petersburg

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

  23. EU-PWI-Task Force Work in plasma simulators + Dedicated lab experiments + Tokamak research • Fuel Removal • Isotope exchange on PFC side • Thermal desorption on PFC side • Oxygen venting remote areas? • Scavenger techniques ?? • and Fuel Control • Temperature tailoring • Carbon traps • Divertor geometry … Needs detailed understanding of the involved physics 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) EPS 2003, Petersburg

  24. Plasma EU-PWI-Task Force D+ 99.8 0.002 Divertor Strike zones Full metal wall: wall lifetime If the T-retention problem cannot be solved a full metal first wall concept is needed based on metals with low hydrogen retention Wall Lifetime Plasma compatibility • Steady state erosion • Low erosion materials • High local re-deposition • Lost material replaced from main chamber Transient events The main concern with metal walls is the lifetime due to melt layer erosion in transient heat loads (ELMs and Disruptions) EPS 2003, Petersburg

  25. EU-PWI-Task Force Energy on target plates Wetted area Duration of ELMs 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 Particle flux Stored energy • Material limits T < 2300o (C), < 3400o (W) • T  energy/ area/ sqrt(time)  20(40) MJ m-2 s -1/2

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

  27. EU-PWI-Task Force EPS 2003, Petersburg Lifetime: ELMS & Disruptions (3) Predictions for ITER divertor target lifetime G. Federici et al. number of ELMs for 1000 ITER pulses  4 x 105 ITER Predictions ELMs are marginally acceptable but behaviour forCFC and W not much different at moderate power densities

  28. EU-PWI-Task Force ASDEX Upgrade Poloidal Distance (m) EPS 2003, Petersburg Lifetime: ELMS &Disruptions (4) 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 Disruption Energy deposition (Thermal Quench) occurs over a large divertor area (AUG & JET) Fraction of energy deposited in Divertor: 1 (AUG), ~ 0.2 (JET) V. Riccardo, I.3.5. Wed G. Pautasso P-1.135, Mon P. Andrew P-1.54, Mon

  29. EU-PWI-Task Force 100 1 10 0.5 Disruption energy density (MJ) 1 0.1 0.1 0 5 10 15 20 25 30 Lifetime: ELMS &Disruptions (5) Fraction of energy to divertor PWI TF issues G. Federici, ITER JWS Garching, 7 Oct. 2002 2 lambda =5 mm mid • Spatial and Time Evolution of thermal Energy Flux • Dependence on Disruption Type • Energy Balance during Current Quench (Halo Currents) Current ITER specifications Disruption energy density, MJ/m Absence of melting for W SOL Broadening SOL broadening, lambda /lambda plate mid • Development of disruption mitigation techniques • Power exhaust on irregular, molten surfaces Disruptions energy deposition and mitigation are key issues for the ITER material selection

  30. 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.

  31. 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

  32. EU-PWI-Task Force Additional remarks • A graphite and a tungsten divertor should be prepared in parallel. • Possibilities to measure thefuel 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 firstwall 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.

More Related