Seminar in ASIPP, 11 Dec Page 1 Status of ITER Project and Issues of Plasma-Wall Interaction Michiya Shimada With contribution from Richard Pitts and David Campbell ITER Organization Seminar in ASIPP, Hefei 11 Dec. 2009 Seminar in ASIPP, 11 Dec Page 2 Status of ITER ITERs objectives ITER design goals Main parameters of ITER ITER construction site ITER schedule Design Review PWI issues Choice of Plasma-Facing Materials Wall conditioning Contents Seminar in ASIPP, 11 Dec Page 3 The overall programmatic objective: to demonstrate the scientific and technological feasibility of fusion energy for peaceful purposes The principal goal: to design, construct and operate a tokamak experiment at a scale which satisfies this objective ITER is designed to confine a DT plasma in which - particle heating dominates all other forms of plasma heating: a burning plasma experiment ITERs objectives Seminar in ASIPP, 11 Dec Page 4 Physics: ITER is designed to produce a plasma dominated by -particle heating produce a significant fusion power amplification factor (Q 10) in long-pulse operation aim to achieve steady-state operation of a tokamak (Q = 5) retain the possibility of exploring controlled ignition (Q 30) Technology: demonstrate integrated operation of technologies for a fusion power plant test components required for a fusion power plant test concepts for a tritium breeding module ITER Design Goals Seminar in ASIPP, 11 Dec Page 5 The main parameters of ITER are chosen to fulfill ITERs goals Seminar in ASIPP, 11 Dec Page 6 ITER Construction Site Seminar in ASIPP, 11 Dec Page 7 Updated Schedule (IO Proposal) Seminar in ASIPP, 11 Dec Page 8 During the Design Review that was conducted during the period , the recommendations were made in the following physics area: Expansion and revision of the heat loads specifications associated with unmitigated disruptions, VDEs and ELMs have confirmed their serious consequences; implementation of their mitigation measures have been recommended Improvement of the plasma shaping and position control capability The divertor target material TF ripple Design changes and/or R&D programmes have been implemented in response to each of these recommendations. In some cases further analysis and experimental work is required, either to complete design specifications (e.g. in-vessel coils) or to provide an improved physics basis for the operation of ITER (e.g. the use of a full tungsten divertor). Design Review Seminar in ASIPP, 11 Dec Page 9 Heat load specifications Heat load specifications of PFCs have been revised to reflect recent experimental results [Loarte, IAEA 08] New specifications cover the steady-state heat loads as well as transient heat loads e.g. disruptions, VDEs and ELMs These specifications confirm very serious consequences of ELMs, disruptions and VDEs on PFCs, indicating the need of mitigating or avoiding these phenomena These specifications have large uncertainty, requiring continued experiments in the existing tokamaks Seminar in ASIPP, 11 Dec Page 10 ELM induced erosion Lifetime of PFCs Results from Russian plasma simulators: Recommended threshold for damage 0.5 MJm -2 adopted by ITER Efficient mitigation methods needed energy density / MJm negligible erosion erosion at PFC corners CFC energy density / MJm negligible erosion melting of tile edges W Erosion limit for CFC reached due to PAN fibre erosion Erosion limit for W reached due to melting of tile edges Increasing PAN fibre erosion Increasing melting and droplet ejection Crack formation was observed at energy densities 0.7 MJ/m 2. Repetitive sub-threshold ELM investigations ongoing in JUDITH2 Unmitigated ELM:~10 MJ/m 2 Seminar in ASIPP, 11 Dec Page 11 Results with magnetic control look promising: studies underway to design control coil system for ITER ELM pacemaking using pellet injection also effective: quantitative basis for application in ITER being studied DIII-D Magnetic ControlAUG Pellet Pacemaking ELM Control/ Mitigation Seminar in ASIPP, 11 Dec Page 12 In-Vessel Coils A set of resonant magnetic perturbation (RMP) coils under design: consists of 9 toroidal x 3 poloidal array on (outboard) internal vessel wall vertical stabilization coils consist of upper/ lower loops forming saddle coil RMP Coils VS Coils Seminar in ASIPP, 11 Dec Page 13 Disruptions occur in tokamak plasmas when unstable p(r), j(r) are developed MHD unstable modes grow plasma confinement is destroyed (thermal quench) plasma current vanishes (current quench) Typical timescales Thermal quench < 1ms deposition of plasma thermal energy on PFCs Current quench > 10 ms deposition of plasma magnetic energy by radiation on PFCs & runaway electrons Typical values for ITER current quench W poloidal ~ 1 GJ c.q. ~ ms q rad ~ 35 70 MWm -2 A wall ~ 700 m 2 q rad c.q. 1/2 ~ 710 MJm -2 s -1/2 (no Be melting) JET Disruptions Seminar in ASIPP, 11 Dec Page 14 When a loss of vertical position control takes place: plasma impacts wall with full plasma energy high localized heating mitigation required Control issues Detection of loss of vertical position control Fast stop of plasma by massive gas injection, killer pellets, etc. Issues of effectiveness, reliability of mitigation method, as well as additional consequences (runaway electrons) need to be addressed in experiment ITER simulation Halo current layer Vertical Displacement Events - VDEs Seminar in ASIPP, 11 Dec Page 15 The development of high pressure impurity gas injection looks very promising for disruption/ VDE mitigation: efficient radiative redistribution of the plasma energy - reduced heat loads reduction of plasma energy and current before VDE can occur substantial reduction in halo currents (~50%) and toroidal asymmetries DIII-D Disruption/ VDE Mitigation Seminar in ASIPP, 11 Dec Page 16 PF and CS capabilities have been improved for better flexibility of operation Seminar in ASIPP, 11 Dec Page 17 Optimization of the distribution of ferromagnetic material in the vacuum vessel shell has been made so as to minimize the level of TF ripple and its possible impact on the quality of H-mode plasma confinement Seminar in ASIPP, 11 Dec Page 18 New WBS structure (2009) for PWI breaks down into 6 key areas: (underline: partially covered in this talk) T-retention and inventory control Tungsten R&D Heat fluxes to PFCs Dust Wall conditioning Erosion and migration Consistent with priority R&D Topic Areas established in 2008 ITER PWI Research Plan and now the focus of ITPA DIVSOL TG Overall PWI priorities Seminar in ASIPP, 11 Dec Page 19 ITER materials choices Be for the first wall Low T-retention Low Z Good oxygen getter Driven by the need for operational flexibility For H and part of D phase: C for the targets Low Z Does not melt Excellent radiator W for the dome/baffles High Y phys threshold For D and DT phases: Be wall, all-W divertor* To avoid problem of T-retention W CFC Beryllium Surface areas: Be: 700 m 2, W: 100 m 2 CFC: 50m 2 Expedited R&D should be pursued for the use of tungsten Seminar in ASIPP, 11 Dec Page 20 A consequence of full tungsten divertor Since the acceptable level of tungsten impurity in the plasma is ~10 -5 while a few % level of concentration is acceptable for light impurities such as beryllium and carbon, optimization of operating scenarios is important to avoid localized melting and contamination of the core plasma with tungsten ions. Therefore the mitigation of transient heat loads must be demonstrated during the non-active (H/He) phase. Seminar in ASIPP, 11 Dec Page 21 Ongoing collaborations Emphasis has been on answering urgent design questions, notably for the first wall design. Several PWI experiments performed as a direct result of requests from the IO, others through ITPA Channel: Start-up heat loads: DIII-D (APS 2009 Poster D. Rudakov et al.) Tore Supra (experiments underway) Secondary divertor heat loads DIII-D (APS 2009 Poster J. Watkins et al.) TCV (experiments underway) Toroidal uniformity of divertor gas injection C-Mod (first part of experimental programme complete, rest before end 2009) Experimental tests of ITER erosion-migration modeling strategy EAST migration limiter pre-tests underway in view of dedicated experiment Tore Supra proposal made for a possible experiment Seminar in ASIPP, 11 Dec Page 22 Wall conditioning * Remarkable contribution from EAST and HT-7 Goals reduction of impurity, tritium retention, dust and particle recycling Schemes baking (divertor: 350 C, FW: 240 C, VV: 200 C, no Bt) glow discharge cleaning (6 electrodes, no Bt) RF (IC and EC; extensive review by C. Schueller) separatrix sweeping disruptive discharge vacuum cleaning (during vent) If HF GDC is efficient and feasible for ITER, its impact on ITER operation would be tremendous Seminar in ASIPP, 11 Dec Page 23 Be deposit layer can desorb most of tritium with baking of 350 C Seminar in ASIPP, 11 Dec Page 24 Possible issue: tritium removal from the dust and flakes deposited under the divertor cassettes 350 C baking 200 C baking Seminar in ASIPP, 11 Dec Page 25 IC wall conditioning in HT-7 (Hu, 2007) Gas removal rates (x10 21 atoms/h) HDCO He D2D O2O He/O (4:1)4.87 O-ICD had a factor of 4-6 higher H removal rate than He-ICD. O-ICD shows ~ 20 times higher deposit removal rate than He-ICD and D 2 ICD and the efficiency of deposit removal of O-ICD is comparable to O-GDC; C removal rates in O-ICD may correspond to a T removal rate of 0.12 gT/h in ITER; C removal rates in D-ICD may correspond to a T removal rate of gT/h Seminar in ASIPP, 11 Dec Page 26 Issues with wall conditioning with oxygen Consequences to plasma operation in-vessel components Tritium system (corrosion by DTO ) Seminar in ASIPP, 11 Dec Page 27 Comment about HF GDC Design Review of ITER GDC system will be carried out toward the end of 2010 If HF GDC is shown to be efficient, we should start the design from the beginning of 2010 The information on the efficiency compared with other wall conditioning schemes e.g. ICWC and DC GDC will be essential Contribution in this area will be appreciated!!!