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

2. 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 reporting period

3. 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

4. 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

5. 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

6. Personnel assignments
Large number of bi-lateral and multi-lateral research collaborations arranged over the 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

7. 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
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 period
Publication list included as part of the RfE dossier

8. 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)

9. 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)

10. 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)

11. 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)

12. 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)

13. 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)

14. 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)

15. 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)

16. 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

17. 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

18. 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 (
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