Request for Extension of the Implementing Agreement for Co-operation on Tokamak Programmes Richard A. Pitts ITER Organization

Acknowledgements I wish to thank D. Borba, CTP Secretary, for his effort and invaluable assistance in preparing the Request for Extension Dossier And ITER Organization Science Division colleagues for assistance in surveying the publication output in joint activities over the 2012-2017 reporting period

Objective and scope of the CTP-TCP (1) Advance fusion science and technology by strengthening cooperation amongst tokamak programmes. Promote Exch related to the experimental programmes of the partner tokamak facilities Design and planning of experiments to contribute to the database for next generation tokamak devices  support of activities identified by the International Tokamak Physics Activity (ITPA) operating under the auspices of the ITER Organization since 2009

Objective and scope of the CTP-TCP (2) Experimental, theoretical and technical studies in Plasma equilibrium and stability Energy and particle transport Plasma heating and current drive Plasma-wall interaction and divertor physics Pedestal physics, including Edge Localized Mode (ELM) control Energetic particle-driven instabilities, transport and confinement Integrated scenario development Plasma fuelling Plasma diagnostics

CTP Contracting Parties (CP) Currently 7 CPs 3 IEA member countries: Japan, South Korea, USA 1 partner country India 3 International Organizations: EURATOM, ITER Organization, ITER China Domestic Agency

Personnel assignments Large number of bi-lateral and multi-lateral research collaborations arranged over the 2012-2017 reporting period amongst the 7 CPs 2013 2014 2015 2016 EU – KO: 6 Exch, 68 ppd EU – US: 1 Exch, 21 ppd IO – EU: 10 Exch, 32 ppd JA – EU: 7 Exch, 2 ppy, 38 ppd JA – KO: 5 Exch, 45 ppd JA – US: 3 Exch, 18 ppd KO – EU: 4 Exch, 68 ppd KO – JA: 1 Exchange, 5 ppd KO – US: 5 Exch, 284 ppd US – EU: 12 Exch, 215 ppd US – KO: 6 Exch, 1 ppy 34 ppd EU – US: 5 Exch, 61 ppd IO – EU: 14 Exch, 38 ppd IO – KO: 2 Exch, 9 ppd IO – US: 4 Exch, 35 ppd IO – JA: 1 Exch 3 ppd JA – EU: 6 Exch, 2 ppy, 23 ppd JA – US: 2 Exch, 8 ppd KO – EU: 2 Exch, 7 ppd US – EU: 10 Exch, 1 ppy, 73 ppd US – KO: 11 Exch, 174 ppd US – EU: 5 Exch, 55 ppd US – KO: 4 Exch, 40 ppd US – CN: 22 Exch, 365 ppd JP – US: 4 Exch, 35 ppd JP – EU: 2 Exch 375 ppd EU – US: 6 Exch, 90 ppd IO – EU: 6 Exch, 19 ppd KO – EU: 3 Exch, 55 ppd EU – CN: 9 Exch, 109 ppd EU – JP: 1 Exch, 10 ppd EU – KO: 3 Exch, 34 ppd EU – US: 14 Exch, 234 ppd JP – KO: 1 Exch 10 ppd JP – US: 2 Exch, 15 ppd JP – EU: 7 Exch, 405 ppd KO – EU: 1 Exch, 30 ppd US – EU: 7 Exch, 35 ppd US – KO: 7 Exch, 66 ppd US – CN: 21 Exch, 432 ppd 60 personnel exchanges 828 ppd, 3 ppy 57 personnel exchanges 431 ppd, 3 ppy 52 personnel exchanges 1034 ppd 73 personnel exchanges 1380 ppd

Meetings and publications: 2012-2017 5 ExCo meetings at the ITER Headquarters 5 Joint Experiment Workshops (JEX) with the ITPA 61 meetings of the ITPA Topical Groups across all nations of the ITER Members 3 workshops on Theory and Simulation of Disruptions (see http://tsdw.pppl.gov/index.html) Joint activities generated ~90 publications in peer reviewed journals (e.g. Nucl. Fusion, Plasma Phys. Control. Fusion, Phys. Plasmas, Phys. Rev. Lett., Fus. Sci. Tech., Fus. Eng. Des.) over the 2012-2017 period Publication list included as part of the RfE dossier

Key physics achievements Selected results from joint activities given in the next few slides To provide a flavour of the very extensive work performed in the key topical areas (scenario development, ELM control, tungsten transport, plasma boundary, energetic particles, diagnostics)

Integrated Operation Scenarios Review the parameter space of the main ITER baseline (QDT = 10) operation scenario* Review captures more than 15 years of operational experience from several tokamaks Expected ITER operational points fall within the operation space achieved Operational space differs for metal-wall operation (as in ITER) Large variation in normalized energy confinement Normalized energy confinement Normalized plasma density *A.C.C. Sips et al, IAEA FEC (Kyoto, 2016)

Edge Localized Mode (ELM) control using 3D fields Based on original work on DIII-D, ITER will be equipped with 27 internal coils to create 3D magnetic fields for ELM control Thanks to collaborative work under the CTP, this control scheme now demonstrated on many tokamaks (DIII-D, AUG, KSTAR, EAST) and for ITER-like edge plasmas (DIII-D, AUG) ITER AUG R. Nazikian et al, IAEA FEC (Kyoto, 2016)

Divertor power load scaling Improvements in infra-red camera technology and collaborative effort provides new extrapolation to ITER* lq,measured (mm) lq,regression (mm) lq H-mode inter-ELM Heat flux on target Only significant scaling is lq  1/Ip Predicts lq ~ 6 mm on the target for ITER in baseline QDT = 10 at 15 MA About 5x lower than previously thought and unfavourable Has stimulated a great deal of follow-up activity (ongoing) *T. Eich et al, Nucl. Fusion 53 (2013) 093031

Energetic particles Strong interaction seen (AUG, DIII-D, KSTAR) between 3D fields for ELM control and fast ion losses* Careful analysis of experiments helps to validate state-of-the-art modelling allowing predictions for ITER to be made for the simultaneous achievement of ELM control whilst minimizing fast ion losses *M. Garcia-Munoz et al., Nucl. Fusion 53 (2013) 123008

Control of tungsten (W) transport and accumulation* Burning plasmas must be very pure  too much W due to divertor erosion will reduce fusion burn and even terminate discharges due to radiative instabilities Experimentally validated physics model for W transport from the edge to the core has been developed for ITER W accumulation control modelled and demonstrated with ITER-like plasmas and heating schemes (AUG, Alcator C-Mod, JET, …) AUG W density peaking ITER Central pe/ptot *C. Angioni et al, A. Loarte et al., IAEA FEC (Kyoto, 2016)

Coordinated research on cleaning of diagnostic mirrors in ITER* Coating of diagnostic first mirrors is a major issue faced by ITER  in-situ cleaning is critical for sufficient lifetime Mirror cleaning test in EAST Radio Frequency plasma mirror cleaning Test of mirror materials RF discharge provides efficient cleaning Single crystal molybdenum and rhodium are prime mirror materials First ever successful mirror cleaning test in a tokamak (EAST, China) *e.g. L. Moser et al., Nucl. Fusion 55 (2015) 063020

Significant facility upgrades in many CTP partners (here just 5 examples) New tokamak HL-2M (China) under construction. High beta, high bootstrap current plasma and tests of advanced divertor configurations Vacuum Vessel welding to 340 completed end Nov. 2016 JT-60SA (Japan): all new JET-scale superconducting tokamak on schedule for operation in 2019 First plasma December 2016 EAST in-vessel, 2016 ADITYA tokamak (India) being upgraded from limiter to divertor device Tore Supra tokamak (EU) converted to WEST for tests of ITER divertor technology in an integrated tokamak environment New ITER-like tungsten divertor in EAST (China)

Progress at the ITER Organization* Rework in 2015 of baseline schedule (‘Staged Approach’) Dec. 2025 Jun. 2026 Jun. 2028 Dec. 2028 Jun. 2030 Sep. 2031 Jun. 2032 Mar. 2034 Mar. 2035 Dec. 2035 FP PFPO-1 PFPO-2 FPO Int. Com I (12 M) Engineering Operation (SC Magnets) (6 Months) Assembly II (24 Months) Integrated Comm. II (6 Months) Pre-Fusion Power Operation-I (18 Months) Assembly III (15 Months) Integrated Comm. III (9 Months) Pre-Fusion Power Operation-II (21 Months) Assembly IV (12 Months) Integrated Comm. IV (9 Months) DT Ops. Ptotal=73 MW Ptotal=73 MW 100 PECRH=20 MW (all launchers) Possibly PICRH = 10 MW PECRH=20 MW PICRF=20 MW PNB=33 MW PECRH=20 MW PICRF=20 MW PNB=33 MW Heating Power (MW) 50 PECRH=8 MW (upper launcher) 4 operational phases with staged introduction of heating power and measurement capability Currently reworking the Research Plan to accommodate the new schedule  folding in newest R&D obtained in the Fusion Community and with the active participation of experts within CTP partners *See talk by L. Coblentz for more on construction status

Disruption mitigation: good example of a multi-lateral CTP collaboration Mitigation of disruptions a critical issue for ITER Framework agreement between 3 CTP partners (IO, EU, US) to design, construct, install and operate ITER relevant hardware on JET Test the concept of Shattered Pellet Injection (SPI) for mitigation of runaway electrons during disruption current quench Test a 3-barrel SPI system (ITER prototype) on JET  largest existing tokamak  more energetic and larger runaway target plasmas in the same wall material combination as ITER

Strategy for extended CTP Mostly business as usual ExCo will continue to meet annually at ITER HQ Organized jointly with ITPA JEX  maximize synergy between the two activities Workshops on Theory and Simulation of Disruptions will continue Next edition, PPPL July 17-19 2017 (http://tsdw.pppl.gov/) Membership and outreach Russia invited to join as CP in 2015, but no response to CTP request Will invite Russia again in 2017, via a different contact person Will invite Australia as CP in 2017 following their interest in collaboration with ITER and via ITPA