PWI questions of ITER review working groups WG1 and WG8 : Materials Introduction EU PWI TF V. Philipps, EU PWI TF meeting, Oct 2007, Madrid V. Philipps,

Slides:



Advertisements
Similar presentations
J. Roth, EU PWI TF, SEWG Fuel Retention, Cadarache, June 15, 09 Tritium inventory: Joint international scaling for ITER WP09-PWI-01-01/IPP/PS Status by.
Advertisements

G. Arnoux (1/19) SEWG on transient heat loads Ljubljana, 02/10/2009 Heat load measurements on JET first wall during disruptions G. Arnoux, M. Lehnen, A.
EU-PWI Taskforce EU PWI TF Meeting Nov. 4 – 6, 2009, Warsaw Summary of the PSI facility review meeting presented by R. Neu based on the Summary of the.
Slide Oct 2005, EFDA PWI meeting, CEA CadaracheI.S. Landman, FZ-Karlsruhe FZK Investigations on Wall Surfaces and Tokamak Plasma 1 Forschungszentrum.
Report IPP Garching EU Task Force PWI Meeting, Cadarache Oct Max-Planck-Institut für Plasmaphysik compiled by Arne Kallenbach (IPP - EU-PWI.
Alberto Loarte EU Plasma-Wall Interaction Task Force Meeting – CIEMAT – 10 – EFDA Plasma Edge Technology Programme Monitoring of 2006 Activities.
Report on SEWG mixed materials EU PWI TF meeting Madrid 2007 V. Philipps on behalf of SEWG members Mixed material formation is a among the critical ITER.
R. Doerner, Oct. 18, 2005 EU PWI TF meeting, France Beryllium and carbon mixed-material studies R. P. Doerner, M. J. Baldwin, J. Hanna and D. Nishijima.
E. Tsitrone, EU PWI TF meeting, oct 2007, Madrid1 E. Tsitrone, J. Roth, A. Loarte, J. Paméla The EU-PWI Task Force: Task agreements for 2008 EU PWI.
A new look at the specification of ITER plasma wall interaction and tritium retention J. Roth a, J. Davis c, R. Doerner d, A. Haasz c, A. Kallenbach a,
1 3rd meeting Working Group 5. 2 Agenda WG5 2nd Meeting Agenda: Discuss existing proposals Indicate missing proposals Allocate persons to develop proposals.
ITER & dust Dust = major issue in term of ITER safety In vessel dust quantities: extrapolation from current experiments In vessel dust diagnostics & removal.
Max-Planck-Institut für Plasmaphysik EURATOM Assoziation Interaction of nitrogen plasmas with tungsten Klaus Schmid, A. Manhard, Ch. Linsmeier, A. Wiltner,
Th Loarer - SEWG on Fuel retention – JET, July Th Loarer with special thanks to S Brezinsek, J Bucalossi, I Coffey, G Esser, S Gruenhagen.
WP10-PWI (02)/TEKES/BS(PS) Characterization of retention mechanisms in AUG Monitoring meeting of the EFDA PWI SEWG on Gas Balance and Fuel Retention,
R. Doerner, EU PWI Task Force Meeting, Frascati, Italy, Oct , 2008 US-EU Collaboration on Mixed Materials for ITER: Past, Present and Future R. P.
SEWG Meeting HIGH-Z, Ljubljana, October 2009 I. Tungsten distribution on limiters after WF 6 injection in TEXTOR II. SEM and EDX of Melted Tungsten Rods.
Kazuyoshi Sugiyama, SEWG meeting, Culham, July Outline: 1.Introduction 2.Experimental procedure 3.Result 4.Summary Kazuyoshi Sugiyama First.
ERO modelling of local 13 C deposition at the outer divertor of JET M. Airila, L. Aho-Mantila, S. Brezinsek, P. Coad, A. Kirschner, J. Likonen, D. Matveev,
Max-Planck-Institut für Plasmaphysik EURATOM Assoziation K. Schmid SWEG Deuterium retention in graphite samples exposed to beryllium-seeded.
SEWG Gas Balance 2007 © Matej Mayer First results on deuterium depth profiling in W tiles M. Mayer 1, V.Kh. Alimov, V. Rohde 1, J. Roth 1, A. Herrmann.
1TPL Dismantling Project Review 08/12/06 Bernard Pégourié TORE SUPRA Association EURATOM-CEA D program: Specific experiments before dismantling Purpose:
Joint SEWGs-TFE meeting S. Brezinsek22/07/2008 TF E Impact of N 2 on carbon chemistry in JET S. Brezinsek, Y. Corre and TFE.
EU PWI Task Force V. Philipps, SEWG mixed materials, JET ITER-like Wall Project : Material choice, issues to investigate and role of new SEWG ITER-like.
1. Qualifying carbon as PFC Erosion (see report S. Brezinsek ) along plasma wetted areas, effect of substrate Local C migration to gaps Fuel retention.
V.Philipps, SEWG Gas balance and fuel removal, JET, , Association EURATOM – FZJ Effect of disruptions on fuel release from JET walls V. Philipps,
TFE Th Loarer – SEWG – 12 September Euratom Th Loarer V Philipps 2, J Bucalossi 1, D Brennan 3, J Brzozowski 4, N Balshaw 3, R Clarke 3, G Esser.
D retention and release behaviour of Be/C/W mixed materials
Max-Planck-Institut für Plasmaphysik EURATOM Assoziation K. Schmid SEWG meeting on mixed materials Parameter studies for the Be-W interaction Klaus Schmid.
Slide Nov 2006, EFDA PWI meeting, LjubljanaI.S. Landman, FZ-Karlsruhe Modelling on Wall Surfaces and Tokamak Plasma Consequences of ITER Transient.
1E. Tsitrone PWI TF meeting, 27-29/10/2008, Frascati Euratom International context : ITPA (International Tokamak Physics Activity) Changes in the ITPA.
R. Doerner, EU SEWG meeting, JET. July 9-10, 2007 Recent Results from PISCES Program R. P. Doerner, M. J. Baldwin, G. De Temmerman, J. Hanna, D. Nishijima,
J. Roth Bilateral Agreements EU-DOE: e.g.: PISCES operation (following talk by Russ Doerner) Bilateral Agreement EU-Russia: collaboration on Material damage.
Member of the Helmholtz Association Carbon based materials: fuel retention and erosion under ITER-like mixed species plasma conditions Arkadi Kreter et.
R. Doerner, EU SEWG meeting, JET. July 9-10, 2007 Co-deposition/Co-implantation R. Doerner, M. Baldwin, G. De Temmerman, D. Nishijima UCSD K. Schmid, Ch.
Institute for Plasma Physics Rijnhuizen D retention in W and mixed systems in Pilot-PSI G. De Temmerman a, K. Bystrov a, L. Marot b, M. Mayer c, J.J. Zielinski.
6 th EU PWI TF Meeting Madrid, Oct Tritium Inventory in ITER: Laboratory data and extrapolation from tokamaks Th Loarer, J Roth, S Brezinsek, A.
9 th ITPA Meeting on Divertor/SOL physics, May 7-10, 2007, Garching Improvements to the B2 wall model: coatings and layers X. Bonnin, M. Warrier, D. Coster.
Fyzika tokamaků1: Úvod, opakování1 Tokamak Physics Jan Mlynář 9. Plasma edge and plasma-wall interaction Limiters, divertors, basic SOL model, blobs, modes.
© Olga Ogorodnikova, 2008, Salamanka, Spain Current status of assessment of Tritium inventory in all-W device O.V. Ogorodnikova and E. d’Agata.
First Wall Heat Loads Mike Ulrickson November 15, 2014.
Spectroscopy of hydrocarbon in low temperature plasmas : Results from JT-60U T. Nakano J apan A tomic E nergy A gency, Ibaraki, Japan. 6-9/11/2006 ITPA.
Y. Ueda, M. Fukumoto, H. Kashiwagi, Y. Ohtsuka (Osaka University)
R. Doerner, May 9, 2005 PFC Program Review, PPPL PISCES ITER-simulation experiments on Mixed-Materials (Be, C, W) R. P. Doerner, M. J. Baldwin and D. Nishijima.
R. Doerner, IAEA CRP on H in Materials, Vienna, Sept. 26, 2006 Mixed-material studies in PISCES-B R. P. Doerner, M. J. Baldwin, J. Hanna and D. Nishijima.
Recent JET Experiments and Science Issues Jim Strachan PPPL Students seminar Feb. 14, 2005 JET is presently world’s largest tokamak, being ½ linear dimension.
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.
Divertor/SOL contribution IEA/ITPA meeting Naka Nov. 23, 2003 Status and proposals of IEA-LT/ITPA collaboration Multi-machine Experiments Presented by.
ASIPP Development of a new liquid lithium limiter with a re-filling system in HT-7 G. Z. Zuo, J. S. Hu, Z.S, J. G. Li,HT-7 team July 19-20, 2011 Institute.
How do we deal with the power/energy fluxes we have derived for ELMs, disruptions or others C. Kessel, PPPL ARIES Project Meeting, Jan 23-24, 2012, UCSD.
Edge Localized Modes propagation and fluctuations in the JET SOL region presented by Bruno Gonçalves EURATOM/IST, Portugal.
Selection of Plasma Facing Components for ITER and ITER Operational plan V. Chuyanov (Thanks to Y. Shimomura ) ITPA Garching May
W coating of CFC tiles for the JET new wall - Task Agreement: JW6-TA-EP2-ILC-05 Manufacturing and testing of W-coated CFC tiles for installation in JET.
1 US PFC Meeting, UCLA, August 3-6, 2010 PFC Activities in Alcator C-Mod G.M. Wright, H. Barnard, B. Lipschultz, D.G. Whyte, S. Wukitch Plasma Science.
ITPA - Meeting, Toronto; Session 3 - High Z studies 3 - High-Z studies (Chair - A. Herrmann) 16:25 (0:10) A. Herrmann - Introduction 16:35.
1 Max-Planck-Institut für Plasmaphysik 10th ITPA meeting on SOL/Divertor Physics, 8/1/08, Avila ELM resolved measurements of W sputtering MPI für Plasmaphysik.
Introduction of 9th ITPA Meeting, Divertor & SOL and PEDESTAL Jiansheng Hu
Background Long term tritium retention is one of the most critical issues for ITER during the tritium phase. It is mandatory to evaluate the long term.
EFDA EUROPEAN FUSION DEVELOPMENT AGREEMENT Task Force S1 J.Ongena 19th IAEA Fusion Energy Conference, Lyon Towards the realization on JET of an.
ITPA May 2007 © Matej Mayer Carbon Erosion and Transport in ASDEX Upgrade M. Mayer 1, V. Rohde 1, J.L. Chen 1, X. Gong 1, J. Likonen 3, S. Lindig 1, G.
B WEYSSOW 2009 Coordinated research activities under European Fusion Development Agreement (addressing fuelling) Boris Weyssow EFDA-CSU Garching ITPA 2009.
ERO code development A. Kirschner M. Airila, D. Borodin, S. Droste, C. Niehoff  The ERO code  ERO code management  Modelling of CH 4 puffing in ASDEX.
Overview of recent work on carbon erosion, migration and long-term fuel retention in the EU-fusion programme and conclusions for ITER V. Philipps a Institute.
Page 1 Alberto Loarte- NSTX Research Forum st - 3 rd December 2009  ELM control by RMP is foreseen in ITER to suppress or reduce size of ELM energy.
J. Roth: ITPA SOL/DIV, Avila, Jan Prediction of ITER T retention levels with W PFCs J. Roth, and the SEWG Fuel retention of the EU Task Force on.
Achievements of the first year of plasma operation with the
Surface Analysis of Graphite Limiter and W-coating Testing on HT-7
ITER consequences of JET 13C migration experiments Jim Strachan, PPPL Jan. 7, 2008 Modeled JET 13C migration for last 2 years- EPS 07 and NF paper in prep.
Paper TH/P Summary Slide (Simulation of Beryllium Erosion and Surface Damage under ITER-like Transient Plasma Heat Loads) It is expected that the.
Presentation transcript:

PWI questions of ITER review working groups WG1 and WG8 : Materials Introduction EU PWI TF V. Philipps, EU PWI TF meeting, Oct 2007, Madrid V. Philipps, B.N. Bazylev, I.S. Landman, L. Colas R. Neu, K. Schmid, R. Dux, R. Doerner, J. Roth, Ch. Linsmeier, V. Bobkov, A. Loarte, A. Kallenbach, A. Kirschner

1.Document the considerations excluding the use of Be also for high heat-flux components in the divertor power handling : Be armour must be thin to allow margins for transients erosion: Ero with no Be from the main chamber (worst case) maximum gross Be erosion 50nm/s with redeposition : 9nm/sec (maximum). with Be/D 0.1% 1nm/s. Calculations indicate Be/D < 0.1% in the outer divertor net Be erosion EU PWI TF V. Philipps, EU PWI TF meeting, Oct 2007, Madrid

2. Clarify the need for additional protection limiters in the vessel to protect the latter during start-up, plasma control excursions and transient events. Support these statements by consistent start-up scenarios The definition of expected wall loads in steady state and in transients done in another group steady state wall loads edges must be protected against parallel power flux (between panels, portplugs, upper dump plates needs a double rough like shaping ( uncertainty of X point position) transients: effective control of ELMS, disruptions, VDes must be on board EU PWI TF V. Philipps, EU PWI TF meeting, Oct 2007, Madrid

3. Melting of the Be wall during transients (including the radiation flash from mitigated disruptions Data on Be melt layer behaviour are from Memos code calculations, not much validated, validation better for W ( RF cooperation) Melting by ion impact protection by vapour shielding Radiation impact no protection, larger evaporation depths. Melt hills and craters of several microns per ELM by Lorentz force,adds up during multiple ELM melting Effect of melt surface roughness on power loading and melting not included and not well known. ELM size Rayleigh Taylor (RT) instability, can be somewhat mitigated by macrobrush design with brush < 4cm ELM size Kevin Helmholtz instability, does not depend on the brush size, large droplets and melt layer splashing. Increasing ELM size EU PWI TF V. Philipps, EU PWI TF meeting, Oct 2007, Madrid

4. Assess design measures to suppress impurity production associated with ICRF on ITER 5. W erosion influx in start-up, steady state and during ELMs W influxSteady stateELM induced Fast ionsICRH induced Main chamber W/s W/s1.2 x10 16 /s W/s Divertor But still uncertainties on The real impurity composition of the ITER plasma (both intrinsic and seeded) The plasma transport in the SOL in particular the outer region, The estimation of the ELM induced impurity erosion The effect of ICRH induced impurity production, evaluated in a separate EFDA task (estimation of small contribution of W erosion by fast ions is more solid). EU PWI TF V. Philipps, EU PWI TF meeting, Oct 2007, Madrid

Conclusions: W plasma impurity contamination can be kept below the critical values under scenario I steady state plasma conditions, for which no accumulation is expected. But : W sources estimated under reference scenario of a partially detached inner and outer divertor small W influx from the lower divertor high flux areas. Loss of detachment increases significantly the gross W influx and the plasma contamination depends on the divertor impurity retention No statement was done on advanced scenarios which may lead to impurity accumulation Start up in a full W ITER difficult or impossible according to calculations. But largely by avalanche effects due to W self sputtering. AUG demonstrate start up with a full W wall with the inner wall as start up area. More information is needed in the area to make firm conclusions but a full W ITER would imply a severe risk to obtain successful routine plasma start up. TEXTOR had severe start up problems with a full set of W coated limiters EU PWI TF V. Philipps, EU PWI TF meting, Oct 2007, Madrid

6. Evaluate the influence of mixed-material formation on plasma compatibility and tritium retention 1. Be-W alloy formation: Present view: inner ITER divertor and the dome is in Be deposition mode Effective Be 12 W formation between K (at lower temperatures the Be diffusion in W is to low and a pure Be layer forms on top of the W, at higher temperatures Be evaporates leading to a thin Be allow layer) the temperature of the Be layer deposition areas on the upper inner W target is not high enough for Be-W alloy formation Temperature excursions by ELMS& disruptions or X point Marfe formation can lead to formation of some Be-W alloying Outer divertor: no (thick) Be layer is expected to form. 2. C- chemical erosion: Complete suppression of C erosion on inner target No full suppression of C chemical erosion on outer target EU PWI TF V. Philipps, EU PWI TF meeting, Oct 2007, Madrid

3. T-retention Retention in Be by implantation saturates,estimated amount of retention from T implantation in the whole Be first wall about 7 gT. Retention by codeposition with Be/C Larger scatter in database Temperature the most important variable. Mixed Be/C and W/C codeposited layers seems not to retain significantly more D compared with the pure materials Layer structure seems to affects the retention: layers deposited at higher ion energies tendency to retain more D than those with lower ion energies. At present a value of D/Be = 8% on the plasma facing side of the inner divertor and a lower value of 1% on the dome area of ITER is recommended. EU PWI TF V. Philipps, EU PWI TF meeting, Oct 2007, Madrid ITER Be walls: gT /shot Inner divertor alone

ITER C- outer divertor: erosion deposition modelling (ERO-code), carbon target, no background impurity flux Carbon target Gross erosion: 7.7·10 22 C/s, 98% redeposition C- deposition on dome 2% C/sec Beryllium Target Gross erosion: 3.2·10 22 Be/s 90% redeposition 10% deposition on dome Be/s x 0.4 x 400= D,T/shot 0.5g T- retention in C layers on dome alone Net erosion Deposition