ITER Research Needs and Possible DIII-D Contributions

Transcription

1 ITER Research Needs and Possible DIII-D Contributions Alberto Loarte Plasma Operation Directorate Acknowledgements: Members of POP and CHD directorates, together with many experts in the international fusion community The views and opinions expressed herein do not necessarily reflect those of the ITER Organization Page 1

2 Synopsis Current status of the ITER Research Plan development Update on recent developments with recommendations on plant/ system requirements Key elements of ITER s Physics R&D Needs MHD stability He H-mode Operation in the non-active Phase ELM, ELM control, Pedestal and Transport and Confinement Divertor/ PWI Integrated operation scenarios and high Z divertor Energetic particles Page 2

3 ITER Research Plan - Status Page 3

4 Research Plan: Background to Present Analysis The timeschedule for the Research Plan has developed significant uncertainties due to deferrals and potential delays in procurement of ancillary systems currently many issues unresolvable as DAs are preparing for anticipated revision of Construction schedule as requested by Council It is proposed to decouple the present analysis of the Research Plan from the Construction schedule: working hypothesis that First Plasma delayed sufficiently so that deferred/ delayed systems could become available for installation during Assembly Phase II key issue is to identify programmatic demands for system availability develop most rapid experimental programme on basis of assumption that systems can be commissioned and introduced as demanded by programme identify implications of unavailability of key systems identify risks and detail implications of investment protection, licensing, use of all-tungsten divertor Page 4

5 Assumptions for Research Plan Basic structure of operational programme developed previously retained duration of experimental campaigns and assumptions on pulsing rate as in previous analyses of ITER Research Plan Working hypothesis that all in-vessel systems installed in Assembly Phase II: includes Diagnostics, H&CD, TBM equipment Experimental programme will use an all-tungsten divertor from the start of the non-active phase In-Vessel Coils will be installed Strategy for authorization of tritium usage allows Mise en Service Partielle: allows commissioning of Tritium Plant with tritium in parallel with support to non-active tokamak operations exploits licensing strategy to make tritium available to tokamak operation as soon as possible after Mise en Service granted DD and DT phases integrated as T-Plant commissioning allows Page 5

6 Proposal on ITER Research Plan Complete Tokamak Core First Plasma Hydrogen/ Helium Phase Complete Initial Deuterium-Tritium Experiments Plasma Restart Start Torus PumpDown Assembly Phase II H/He Operation C1 H/He Operation C2 First Plasma Short Shutdown Plasma Development, H&CD Commissioning, Diagnostics, Control, DMS Commissioning Full Heating capability Pre-Nuclear Shutdown 15 MA and Disruption Characterization/ Mitigation, H&CD Commissioning, Diagnostics, Control, He H-modes, ELM Mitigation D Operation Licensing T-Plant Commissioning T-Plant Commissioning Detritiation System Qualification DT Operation Commissioning/ Hydrogen Tritium-Plant Mise en Service Procedure Tokamak Tritium Introduction T-recycle without ISS Initial low duty cycle DT operation Mise en Service Q=10 Short Pulse D Plasmas, D H- modes D H-mode Studies T-recycle with ISS Q=10 Long Pulse Hybrid Non-inductive TBM Program Scheduled Shutdowns EM-TBM TN-TBM NT/TM-TBM INT TBM Page 6

7 System Requirements for H/He C1 Tokamak Core and in-vessel components are complete H&CD in-vessel components complete and at least 60 MW of installed power available 2 NNB lines (33 MW) (also DNB) 20 MW ECRH (at least Eq Launcher transmission lines installed) 2 ICRF antennas installed, with power sources and transmission lines for one antenna ready to commission with plasma at 10 MW Additional system requirements: Error field correction coils commissioned Vertical stabilization coils commissioned In-vessel viewing system Disruption mitigation system Hot Cell (partial: facility required for Be handling) Tritium Plant (H/D commissioning ongoing see STAC-14) Desirable systems (requirement for H/He C2): ELM coils available for commissioning Multi-Purpose Deployer arm Page 7

8 Additional System Requirements for H/He C2 Additional systems which must be available in C2 campaign either to meet programmatic needs or for commissioning to meet programmatic needs of D/DT phase: ELM coil system (programmatic needs) ECRH transmission lines to Upper Launchers (possible programmatic needs, programmatic needs for D/DT phase) FulI ICRF power and all transmission lines (commission 2 nd antenna for possible programmatic needs, or programmatic needs for D/DT phase) Multi-Purpose Deployer arm (possible programmatic needs) Page 8

9 System Requirements for H/He: Diagnostics Diagnostic Magnetics for position, velocity, shape and mode structure Line averaged electron density (toroidal polarimeter/ interferometer) Runaway electron detection (hard X-rays) Impurity identification and influxes (visible and near UV spectroscopy including H α and visible bremsstrahlung), partial systems Visible TV viewing (spectroscopically filtered), partial coverage Torus pressure and gas composition (torus pressure gauges, RGA) Data for gas balance measurements Visible/IR TV viewing (full coverage) and divertor thermography Langmuir Probes and Divertor Thermocouples (divertor plasma, divertor target characterization) Divertor duct pressure and gas composition (divertor duct pressure gauges, RGAs) Impurity identification and influxes, Z eff (visible and near UV spectroscopy, including H α and visible bremsstrahlung, He, Be, Ne, Kr, Ar, W emission), complete systems Vacuum Ultra Violet spectroscopy High-Z impurity content (XRCS and Radial Soft X-ray Cameras) Halo current measurements Radiated power distribution (Bolometry) Core profiles of density, electron and ion temperatures (TS, ECE, XRCS) Edge profiles of density, electron and ion temperature (TS, ECE, Reflectometry, CXRS) Divertor pressure (Cassette pressure gauges) Dust (and erosion) monitors Stage required From first plasma From first plasma From first plasma From first plasma From first plasma From first plasma From first plasma H/He C1 H/He C1 H/He C1 H/He C1 H/He C1 H/He C1 H/He C1 H/He C1 H/He C1 H/He C2 H/He C2 H/He C2 Page 9

10 Issues/ Risks for Non-Active Phase I Protection of PFCs and in-vessel components relies on adequate diagnostic capability Commissioning of disruption detection/ mitigation: adequate diagnostic capability required for reliable disruption detection time needs to be invested in identifying operational limits and developing disruption detection and disruption avoidance scenarios significant effort required to match DMS capability to plasma parameters H-mode operation: adequate H&CD power (> 50 MW) must be commissioned for operation sufficiently early in H/He C2 to permit H-mode experiments Page 10

11 Issues/ Risks for Non-Active Phase II ELM mitigation: use of tungsten divertor implies that ELM frequency control (Waccumulation) and ELM mitigation must be developed from start of H-mode experiments Time pressure in programme is likely to mean that development of current drive capability and advanced scenarios is delayed Some damage/ failure of in-vessel components seems likely, which would require in-vessel access and delay programme: uncertainty over procurement timescale for multi-purpose deployer arm is an additional factor requirements for dust characterization/ fuel retention characterization undefined at present but may require in-vessel access Page 11

12 Issues/ Risks for Non-Active Phase III All-tungsten divertor (all-metal PFCs): necessitates adequate diagnostic coverage of plasma-wall interactions, impurity influxes and PFC damage requires cautious approach to development of higher current/ higher power scenarios to avoid early damage to PFCs early introduction of impurity seeding necessary reliable disruption detection/ mitigation must be demonstrated at each phase of programme reliable ELM control essential helium operation might be problematic due to surface modifications of W-PFCs Page 12

13 Issues/ Risks for Non-Active Phase IV Use of all-metal PFCs necessitates dedicated accompanying programme in coming years: analysis/ scaling of power decay length in limiter, divertor L-mode, H- mode, detached H-modes analysis/ scaling of disruption loads (unmitigated/ mitigated, main chamber and divertor profiles) scaling for ELM footprint (He and D plasmas, unmitigated/ mitigated ELMs) improved H-mode characterization of He plasmas (energy confinement, power threshold scaling, ELM behaviour, ELM mitigation) analysis of operational compatibility of high confinement H-modes and allmetal PFCs, and implications for ITER quantitative analysis of effect of He at ITER divertor target parameters (power, surface temperatures, He fluence) Page 13

14 System Requirements for D/DT Operation All aspects of plasma control associated with equilibrium control, heat load control, machine protection, should be commissioned Reliable disruption mitigation and ELM control should have been demonstrated All H&CD systems should be commissioned to full power and pulse durations of at least several 10s of seconds All fuelling systems installed and operational Extensive diagnostic capability should be available: list of non-active diagnostics should be fully operational diagnostic systems required for advanced control (eg current profile) should largely have been commissioned in non-active phase neutron diagnostics should be calibrated at end of pre-nuclear shutdown other systems (eg fast particle diagnostics) for burning plasma physics studies should be available for commissioning at start of D phase Tritium Plant commissioning should have progressed to allow initial low T throughput experiments to proceed Page 14

15 Key elements of ITER s Physics R&D Needs Page 15

16 MHD Stability - Disruption Loads Mature understanding exists for symmetric electro-magnetic loads Rotation of MHD modes during the CQ is an issue for ITER because of potential dynamic amplification Understanding is on an empirical level, extrapolation from existing tokamaks to ITER is not possible at present Simulation efforts (3D MHD and simplified sink & source) to improve physics picture and to assess resulting forces Uncertainties in heat loads have been accounted for by conservative assumptions, resulting in heat fluxes significantly above melting thresholds for high performance plasmas Runaway electrons can cause significant local damage of PFCs and their characterization, avoidance/mitigation has very high priority Page 16

17 Disruptions: Electromagnetic Loads What is the driver for the current asymmetries (MHD mode)? Relation between poloidal and toroidal halo current (q,m/n)? What drives rotation? Critical because of dynamic amplification! S. Gerasimov EPS 2012 Page 17

18 Disruptions: Thermal Loads Page 18

19 Runaway Electrons : Thermal Loads Page 19

20 Disruption Mitigation: Status and Issues The DMS has to simultaneously fulfill heat and electro-magnetic load mitigation and RE suppression whilst staying in a narrow operating window This and the remaining uncertainties require system with sufficient flexibility implementing SPI and MGI in ITER DMS must be available from start of first H/He campaign (slow CQ, RE) EM and Heat load mitigation are proven feasible (but asymmetries of radiation flash) RE suppression/mitigation is currently an unresolved issue: several mitigation techniques have been addressed by simulation and in experiments, most promising at present is a collisional/radiative scheme ITER will pursue a progressive start-up approach during which the disruption loads and the mitigation efficiency will be validated Page 20

21 RE mitigation: High-Z collisions Collisions with high-z impurities can lead to the dissipation of runaway energy by collisional/radiative processes Increase in pitch angle energy dissipation by synchrotron radiation Timescale in ITER given by the vertical growth time (~100 ms) and MHD stability boundaries eventually reached during vertical movement R&D needed to confirm mechanism, to optimize the process, and to extrapolate to ITER DIII-D: injection into controlled RE beam Hollmann, NF 2013 Simulation using DIII-D parameter Aleynikova, EPS 2013 Page 21

22 He H-mode Operation in non-active Phase - I ITER foresees first H-mode operation in He due to available P add and the expected threshold to access the H-mode in ITER hydrogen in ITER even at 2.65 T H-mode operation in H is unlikely both because the H-mode threshold is high and because the additional heating will be limited to 53 MW (no viable ICRH scheme) Therefore non-active operation in H-mode is based on He H-mode plasmas (7.5 MA/2.65 T) Key purpose of target of these H-modes is the preparation of DT scenarios in ITER and to demonstrate required control schemes (ELM control for W control, radiative divertor operation,.) In addition, He H-modes will give guidance for overall plasma behaviour in H-mode at ITER scale if unexpected behaviour found mitigation actions to be implemented for DT operation Page 22

23 He H-mode Operation in non-active Phase - II Effect of He on H-mode access has been the focus of previous research and the results obtained are already sufficient to plan ITER operations Characterization of He H-modes and study of scenario compatibility issues remains at a very basic level and has to be the key focus of new experiments for ITER Requirements to achieve high P ped Type I ELMy H-modes in He compared to D plasmas Characteristics of edge MHD stability and Type l ELMs in He compared to D plasmas Medium and high Z impurity behaviour in He H-modes compared to D Requirements for ELM control in He plasmas with RMPs ELM control by pellet pacing (H in He in ITER) Radiative divertor operation in He H-modes... Page 23

24 He H-mode Operation in non-active Phase - III He H-modes may present specific features that question their appropriateness as basis to develop DT scenarios in ITER Fuelling of He plasmas must be done with gas fuelling only W production and penetration through the pedestal may be more unfavourable than for DT plasmas ELM characteristics and the response to (RMP) ELM control techniques could be very different in He plasmas and in DT plasmas W divertor with He plasmas has specific PWI issues that have to be investigated ITER Fable/Dux/Coster Page 24

25 He H-mode Operation in non-active Phase - IV W/He interactions lead to surface modification and formation of nanobubbles: PISCES-B results on this first reported at Hefei ITPA Laser irradiation of W exposed to He plasma Damage more severe than for D exposure Contention is that nano-bubbles in the near surface lower the thermal conductivity More work required here new experiments at DIFFER (Magnum- PSI/Pilot-PSI under Eurofusion- ITER Collaboration) 1630 shots, t rise = 1 ms, f pulse = 10 Hz 50 MJ/m 2 /s 1/2 5.2x10 21 cm -2 Miyamoto et al., J. Nucl. Mater., 415 (2011) S657 J. Yu et al., ITPA Hefei Page 25

26 Pedestal, ELM, ELM control and Transport & Confinement Pedestal characteristics in ITER Pedestal height, width : global confinement Transport in pedestal (main ion and impurity) ELM Physics Main Plasma Energy and Particle Losses Power and Particle Fluxes to PFCs ELM control in ITER Transient Power Load controls and ELM frequency (impurity control) Compatibility with ITER scenario requirements confinement, fuelling, collisionality/density regime, radiative divertor Transport and Confinement Influence of plasma facing materials on plasma confinement L-H and H-L transitions Pellet fuelling transients Page 26

27 Pedestal Transport Residual anomalous electron transport in pedestal? Is ion and impurity transport neoclassical in pedestal? Is there a pedestal pinch for the main ion? R. Dux EPS 2013 ASDEX-Upgrade - T. Pütterich JNM 2011 If W transport neoclassical low contamination efficiency in ITER Page 27

28 ELMs : Main Plasma Energy and Particle Losses ELM simulations DW ELM Particle losses Ejection of filaments (~ 2% plasma particles) Energy losses II losses in ergodized field (large ELMs, DT) and ejection of filaments Huijsmans NF MJ 1.5 MJ Page 28

29 ELMs : Power and Particle Fluxes to PFCs ELM Power Particle Fluxes Relation between ELM wetted area and DW ELM? Physics of In/Out ELM power deposition divertor asymmetry? Sharing of ELM energy between divertor and wall? Huijsmans NF MJ 1.5 MJ Page 29

30 RMP ELM Control: ITER R&D Needs Compatibility with ITER constraints: quantification and recovery of confinement loss RMP ELM control with fuelling pellet injection impurity control in RMP ELM controlled plasmas peak power loads with RMP ELM mitigation rotating RMPs for power spreading suppression/mitigation close to LH transition LH Threshold with RMPs RMP ELM control during current ramp Suppression or 30x Mitigation: with good confinement, minimal density pumpout, at ITER q 95 and low torque how to extrapolate RMP suppression/mitigation regimes to ITER? density versus collisionality Theory/Simulation/Model RMP ELM control: criterion for ELM stabilization/ mitigation (including plasma RMP response) origin of density pumpout Page 30

31 Pellet ELM Control DIII-D demonstrated 60Hz pellet pacing: good confinement, reduced impurities and q div 1.3mm ( m/s) pellets reliably trigger ELMs minimum size: >40% of 1.3x0.9mm AUG: W: 70Hz ELM pellet pacing Lag time 10 ms independent of pellet size ELM can be triggered when p within 80% MHD limit high density pellet fuelled regime with small ELMs Also with RMP coils JET: ELM trigger more difficult in ILW (compared C) Lag time due to slower pedestal recovery No reduction of q div ITER: Non-linear MHD simulations yield minimum pellet size for ELM trigger Validation on DIII-D and JET discharges. ITER minimum pellet size sizeable fuel throughput for Q = 10 (~ s -1 ) n = 1 asymmetry ELM of divertor power flux Optimum injection position (HFS, LFS, X-point) Maximum ELM frequency does delay time limit maximum achievable ELM frequency Baylor APS-2013 Page 31

32 Vertical Kicks: JET: ILW reliable ELM trigger (up to 45 Hz) validation that ELMs driven by induced edge current Useful for impurity control in initial low current H-modes More multi-machine experiments needed QH mode: DIII-D: QH-mode with ITER relevant co-i p torque using n=3 C-coil only QH-mode sustained up to ITER relevant n e,ped /n GR = 0.8 Further QH-mode ITER scenario issues need to be addressed (pellet fuelling, high Z impurity control, radiative divertor, ) Non-linear MHD simulations in progress for extrapolation to ITER Edge H&CD: AUG: increased fraction of higher frequency ELMs with edge ECRH EAST: ELM stabilisation by LH Establish controllability of pedestal and ELM characteristics by edge ECH/ECCD/LHCD I mode: ELM Control: Alternatives C-MOD, exploratory experiments in DIII-D and AUG Page 32

33 Effect of PFCs on Confinement W PFCs do not lead to obvious operational restrictions in L-mode Compatibility with W leads to lower P ped and lower H 98 - Increasing b leads to recovery of H 98 (W p ~ P 0.5 instead W p-98 ~ P 0.3 ) ITER? Beurskens, Schweinzer EPS 2013 Page 33

34 H-Mode Exit Low or no hysteresis on edge parameters (ASDEX-U) nor edge & global parameters (JET): implications for H-mode back-transition in ITER (duration of type III H- mode phase and timescale of W plasma collapse) Important to characterize and understand timescales of energy and particle transport (including W) in H-L transition to evaluate risk of radiative collapse in ITER JET-Maggi-EPS 2013 JET + ITER modelling t ~ 5 s t ~ 2.5 s Page 34

35 Particle Transport and Fuelling Physics of main ion core plasma turbulent particle well established but being challenged by DIII-D experiments (no change n o /<n> with n*) Transport during transients shows modifications with respect to stationary phases which can affect effciency of pellet fuelling: MAST Garzotti & Valovic decrease of turbulence in grad-n < 0 region (decreased diffusion inwards) increased large and medium scale transport in grad-n > 0 region B. Baiocchi-C. Bourdelle Page 35

36 Divertor and PWI Following IC-13, it is confirmed that ITER will operate with all-metal PFCs Be first wall (~700m 2 ): - low-z limits plasma impurity contamination - low melting point - erosion/ redeposition will dominate fuel retention - melting during disruptions/ VDEs - dust production Be W divertor (~150m 2 ): - resistant to sputtering - limits fuel retention (but note Be) - melting at ELMs, disruptions, VDEs - W concentration in core must be held below ~ W Page 36

37 Heat Fluxes: limiter SOL widths In the course of dedicated limiter heat flux scaling results, a narrow feature detected on at least two devices: seen very near the LCFS similar to effects seen several times over the past 20 years ago (e.g. T-10, Tore Supra, TEXTOR) Probable cause of melting on JET IWL during first ILW campaign Page 37

38 Heat Fluxes: limiter SOL widths If the narrow feature is there ITER IWL toroidal shaping requires adjustment Decision to be made on detailed shaping of ITER First Wall on the inner side in 2014 Even if no shape change is finally made, experiments should continue to fully understand the controlling physics Assist ITER operations in the future to understand what the limitations may be scenario limitations for ramp-up/down0 Page 38

39 Heat Fluxes: divertor widths Very successful ITPA divertor width scaling effort: resulted in now well known scaling presented at IAEA 2012 l q 1/I p 1/B pol with no dependence on R or P SOL T. Eich et al, Nucl. Fusion 53 (2013) supports well Goldston s drift model Important to determine scaling for high recycling conditions Relation between edge pressure stability limits and l qii depends of edge essential that other devices contribute Page 39

40 Integrated Operation Scenarios High-Z divertor operational scenarios: is the i vs q 95 space for early diverted plasmas similar to that for limited plasmas? H-mode access/ exit scenarios with high-z divertor demonstrate ITER baseline scenario(s) with acceptable impurity control (particularly during transient phases) demonstrate high-z divertor heat flux control at high plasma performance how fast can I p be ramped up and the plasma become diverted with a high-z divertor? Over what density range? With what evolution of Z eff? Validate H&CD scenarios, particularly ICRF (including coupling) Continue investigation of candidate hybrid and steady-state scenarios for ITER Support development of integrated plasma control capability Page 40

41 Further TBM Mock-up Experiments TBM Mock-up Experiments on DIII-D H Reimerdes IAEA 2012 Ferromagnetic material in TBMs affects plasma operation at high β N n=1 EFC only able to ameliorate ~25% of the rotation decrease due to the TBMs in ITER limited effectiveness of n=1 EFC in rotating H-modes contrasts its ability to recover low locking density in L- modes remaining 75% of the rotation decrease must be caused by secondary n=1 or by n>1 components of the TBM field Further experiments needed: at β N > 2 where TBM effects are large at low rotation in ITER regime with localized error field correction near the TBM mock-up Page 41

42 z (m) ECCD NTM stabilization criteria NTM stabilization criteria used (depending on EC deposition width) for ITER Upper Launcher Design Criteria: 1. h NTM = J cd /J bs > h NTM w cd 5 cm & w cd 5 cm (w cd =Dr a, a =2 m) 3. w NTM ~ 4 cm Equivalent for Power: 1.P eta =1.2/h NTM(1MW) 2.P etaw =5/(h NTM(1MW) w cd ) & w cd < 5 cm R (m) t=520s 5/h NTM w cd j CD /j BS D. Farina Narrow w cd Toroidal Injection Angle Large w cd Page

43 Plasma Control R&D Needs R&D required for especially demanding plasma control areas: - Disruption and runaway electron control (prediction, avoidance, mitigation) - First wall and divertor heat flux control - Actuator Sharing - Exception handling - Forecasting - Fusion burn control experimental and modeling simulations - Rotation profile control - Current density profile control - Alfvén eigenmode control Define experiments to demonstrate ITER relevant control schemes Expand modelling to integrate plasma control functions into operational scenarios: - contribute to validation activities for integrated modelling of ITER scenarios Page 43

44 6. Energetic Particles Increasing priority Understand transport, redistribution and loss of fast ions 3D fields: ELM control coils, TBMs, TF ripple, ferritic inserts, AE and other modes Predictions of stability, behaviour and consequences of EP driven modes Identification of unstable modes (linear stability) not just AE Predicted (nonlinear) evolution of unstable modes and fast ion distribution Develop methods and understanding for control of EP driven modes Explore and verify possible actuators: Shape, ECH, ELM coils Develop understanding (linear stability or nonlinear saturation) to validate for ITER Development of fast ion diagnostics for ITER to validate understanding Fast ion loss detector to confirm level of losses before going to high power Diagnostic interpretation techniques and tools Influence of fast ions on equilibrium Alpha current, anisotropic pressure MHD spectroscopy Diagnostic potential of fast ion driven modes: q 0 (ω RSAE (t)), A AE (δω(t)) Page 44

45 Fast Particle loss by 3-D Fields 3-D fields by ELM control coils can produce localized loads on unexpected locations Edge ergodized fields connect field lines inside the plasma with separatrix This increases NBI ions access to loss regions Magnitude of loss sensitive to B edge structure plasma response Page 45

46 Develop methods and understanding for control of EP driven modes ECH clearly shown to impact AE and other EP driven instabilities in wide variety of machines LHD, TJ-II, AUG, KSTAR, Hl-2A, Heliotron J, and DIII-D have all observed significant modification of AE activity with ECH FIDA data (DIII-D) shows reduction in EP transport when RSAE are stabilised Analysis of experimental data underway Preliminary modelling results of DIII-D ECH/AE with TAEFL and GTC show change in AE stability with ECH deposition location consistent with observations More work needed to develop as AE control tool Page 46

47 R&D Needs for ITER in area of EP Physics Understand transport and loss of fast ions Consequences of TBMs + NTM/AE and ELM coils Continue to focus effort on developing predictive modelling capabilities for ITER scenarios Build on fast ion loss/redistribution due to localised AE data Develop and validate understanding of transitions in nonlinear behaviour (EP-4) Establish boundaries for n = 1 internal kink mode behaviour (FB/NRK/LLM) Continue exploration and development of AE control tools and understanding Impact of ECH on Alfven Eigenmode activity (EP-7) Tailoring of EP distribution with ELM coils to influence stability Page 47

48 Conclusions ITER is planning an ambitious programme of physics and technology R&D ranging across accessible burning plasma scenarios: ELMy H-mode inductive, hybrid and steady-state scenarios provide a reference basis for the tokamak design and the planning of exploitation flexibility in device design and auxiliary systems provide scope to adapt research programme in response to ongoing R&D within fusion programme The ITER Research Plan has allowed us to develop the major steps on the path towards DT fusion power production: identification of the principal opportunities and challenges R&D activities in present experimental, theory and modelling programmes will make a significant contribution to providing the physics basis and methodology for resolving the key challenges: cost effective use of the fusion programme s resources Fusion community is an integral part of the preparations for ITER operation Page 48