IEA Technology Collaboration Programme on Tokamak Programmes (CTP TCP)

Annual Briefing 2022


1.     Preface

The objective of the CTP TCP is to advance the physics and technologies related to toroidal plasmas. This is achieved by strengthening cooperation among tokamak programmes, enhancing the effectiveness and productivity of the research and development (R&D) effort related to the tokamak fusion concept, contributing to and extending the scientific and technology database of toroidal confinement concepts, and providing a scientific and technological basis for the successful development of fusion power.


Travel difficulties continued due to the covid-19 pandemic but the situation improved in 2022 and some activities with onsite participation were possible, in addition to the ones conducted using remote participation tools. These covered the participation in the International Tokamak Physics Activities topical group meetings, workshops and joint experiments.


2.     Chair’s report 2022



The Executive Committee of the CTP TCP unanimously resolved through a written procedure in a decision ending on 31st August 2022 to invite the Government of Switzerland and United Kingdom to join the CTP TCP as a Contracting Party. The respective invitations were sent on 30 September 2022. The Government of Switzerland has accepted the invitation.


It was confirmed that the interest of Thailand to join the CTP TCP expressed a few years back have not been pursued recently by our colleagues in Thailand and that no contacts have taken place in the last year.


The Executive Committee expressed once more that Russia would be welcomed as a member to the CTP TCP but added that given the present political situation it is very unlikely that any developments would take place in the near future and agreed no immediate actions on this matter.



·       12th Executive Committee Meeting at the ITER Organisation Headquarters, St. Paul-lez-Durance, France, Thursday 8th December 2022

·       25th meeting of the International Tokamak Physics Activity Coordinating Committee (ITPA CC)

and 13th ITPA meeting for Joint Experiments at the ITER Organisation Headquarters, St. Paul-lez-Durance, France, 6th-7th December 2022

·       The KSTAR conference, which usually takes place in January or February did not take place.


CTP TCP Implementing Agreement Modernisation


The CTP TCP Executive Committee unanimously approved the revised CTP TCP Implementing Agreement prepared by the IEA secretariat and revised by the parties. The new CTP TCP Implementing Agreement is now in force, as amended on 8 December 2022.


Election of the New Chair of the TCP-CTP


The CTP TCP Executive Committee unanimously approved the nomination of Shunsuke Ide (JP) as Chair of the CTP TCP Executive Committee for the period from 1st March 2023 to 28th February 2026.


Status report and milestones achieved



The ITER project has continued making steady progress towards first plasma with 77% of the work completed. 6 out of 8 toroidal field coils have been delivered by Japan and 8 out of 10 toroidal field coils have been delivered by the EU. The poloidal field coil manufacturing progressed with PF6 (CN), PF5 (EU) completed and in the Tokamak pit, PF2 (EU) completed and in storage, PF4 (EU) last stages of manufacture, PF3 (EU) last double pancake completed and PF1 (Russia) factory acceptance test completed. More than 85% of the site civil work is now completed and the base of the Cryostat procured by India was successfully inserted into the Tokamak Pit. Technical issues have been found on dimensional non-conformity of vacuum vessel sectors and cracking on the thermal shield cooling pipes; and solutions to these issues have been identified and a plan for implementation has been developed. The support from the ITPA continues to be very important to finalise the design of components and for the optimisation of the ITER Research Plan. Regarding the Disruption Mitigation System (DMS) using Shatter Pellet Injection, experiments on AUG studied the impact of fragment size, velocity, shatter plume geometry and scaling in all-metal devices; experiments on KSTAR developed new shatter units with 12.5˚ angle that allowed high velocity fragment injection; DIII-D exploited the Thomson Scattering measurement capabilities; and JET used two identical barrels to develop runway mitigation schemes at high current.


Joint European Torus (JET), Axisymmetric Divertor Experiment (AUG) Upgrade, Tokamak à Configuration Variable (TCV), Mega-Amp Spherical Tori (MAST) Upgrade, and Tungsten (W) Environment in Steady-state Tokamak (WEST)


Benign plasma termination was achieved on TCV using real time control of Neon injection, Deuterium massive gas injection and plasma compression. Fully detached L-H transition was achieved on AUG using active control before the heating ramp and ELM control using X-point radiation. He-induced material modifications at Tungsten surfaces and associated impact on power handling, retention, and erosion have been studied in tokamaks and linear devices. Tungsten fuzz growth, net erosion and re-deposition occurred at the outer target plate in AUG but no Tungsten fuzz growth was detected on JET.




Coil energisation was interrupted by an incident on the joints of the EF1 coil, due to insufficient voltage holding capability at the joint. Repair of the joint and reinforcement of potentially weak points on the coil circuits, is almost completed. The integrated commissioning will be resumed after the completion of the work and final confirmation of the voltage holding capability. Based on the trilateral agreement between the ITER organisation (IO), F4E and QST, information sharing with the IO on the assembly and the integrated commissioning is effectively on going. The information on the coil incident is also shared with the IO as a lesson learned for ITER.



Experimental Advanced Superconducting Tokamak (EAST)

Recent EAST upgrades allowed higher Heating and Current Drive power, capable of further exploration of scenarios with a full metal wall, allowing 1056 seconds long-pulse plasma operation achieved with robust plasma control and 300 seconds high-performance H-mode. The high beta poloidal regime was extended and long pulse H-mode operation under ITER-like conditions was achieved.


HL-2A tokamak

High-beta scenarios based on double transport barriers have been realised in HL-2A and simulations suggest that Edge Localised Modes (ELM) could be controlled through turbulence regulation by impurity seeding.


HL-2M tokamak

A new real time feedback control system developed in HL-2M allowed operation with the plasma current above 1 MA and divertor configuration.



Korean Superconducting Tokamak Advanced Research (KSTAR)

In KSTAR, stationary high internal inductance mode discharges were developed at lower magnetic field for long pulse operation. A new internal transport barrier mode in the ion temperature was achieved, well correlated with the fast ion fraction suggesting that the turbulence is suppressed by fast ions. Real-time control of Resonant Magnetic Perturbations leads to long-pulse Edge Localised Mode (ELM) suppression with minimum plasma performance degradation. Self-organisation in the electron temperature profile was measured by Electron Cyclotron Emission imaging diagnostics.  It was found that turbulence spreading around magnetic islands leads to rapid heat transport and reconnection inside the island and the density pump-out is in good agreement with model-based predictions.


United States

Doublet III D-shaped Tokamak (DIII-D)

DIII-D experiments focussed on ITER preparation, core-edge integration, helicon commissioning, tungsten experiments, non-Edge Localised Modes scenario studies and the negative triangularity campaign. It was shown that off-axis Neutral Beam Injection broadens the current profile and decreases the fast-ion transport. It was concluded that the L-H power threshold decreases with increasing non-resonant magnetic fields and Zeff at low collisionality; making possible to reduce the L-H threshold in ITER Hydrogen and Helium plasmas by more than 50%.


National Spherical Torus Experiment Upgrade (NSTX-U)

In NSTX-U, the manufacturing of a new inner Toroidal Fields and Ohmic Heating bundle is in progress and the recovery project is 70% complete. An independent project review of the new proposed NSTX-U recovery cost and schedule baseline was completed in November 2022.




ADITYA-UG experiments in 2022 focussed on plasma parameter enhancement in circular plasmas, disruption mitigation studies with inductively driven pellet injection, shaped plasma experiments using upper and lower divertor coils, real-time density feedback control, deuterium plasma operation, 42 GHZ Electron Resonant Heating  two pulses operation, drift tearing mode rotation studies through electrode biasing, gas-puff induced cold pulse propagation, Neon and Argon impurity seeding and transport studies.


Steady State Superconducting Tokamak (SST-1)

SST-1 was mostly in shutdown and maintenance in 2022.


Future plans


A unified scientific programme spanning over all European devices was developed to take advantage of the large operational space available. The focus will be on a possible JET DTE3 campaign, which duration and scope are subject to the achievements of key performance indicators set in the radiative scenarios and shattered pellet injection preparation experiments. Joint mission-driven actions will continue in EAST, HL-2A and HL-2M tokamaks, which operate as user facilities focussing on wide collaborations.


Future KSTAR upgrades in support of ITER and K-DEMO include Helicon Current drive for long-pulse steady-state operation, plasma auxiliary central heating up to 20 MW of Neutral Beam Injection and Electron Cyclotron Heating and Tungsten divertor for DEMO relevant high-power handling of up to 10 MW/m2. In DIII-D, a High-field-side Lower Hybrid antenna installation is scheduled following operation in 2023.


The Indian Fusion Program developed a 25 year roadmap that foresees two new machines leading to Indian DEMO: a Fusion Neutron Source (FNS) based on a compact spherical tokamak with Copper coils, with a magnetic field of 2T, major radius of 1.5 m, and minor radius of 0.9 m, with full DT operation capabilities with pulses up to 10 seconds; and SST-2 a steady state conventional aspect ratio tokamak with major radius of 4 m, minor radius of 1.4 m, and magnetic field larger than 5 T.