IEA Technology Collaboration Programme on Tokamak Programmes (CTP TCP)

Annual Briefing 2023


1.     Preface and Introduction

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. The main accomplishments are integrated in the International Tokamak Physics Activities, related topical group meetings, workshops and joint experiments. The CTP TCP contributes to the development of ITER physics R&D via joint experiments, development of modelling codes, assessment of key subjects and the development of real-time control schemes. The focus is on minimising the risk and maximising the capability of ITER focussing on high performance with the full Tungsten wall, disruption mitigation, Edge Localised Modes mitigation, detached divertor control and burn control. On the short term ITER aims to create and international team integrating the knowledge of tokamak operation ahead of ITER commissioning. In 2023 important steps were taken towards the development of the tokamak concept for a fusion power plant with the inauguration and first plasma of the JT-60SA Super Conducting Tokamak in Japan and the successful third and final Deuterium-Tritium campaign on the Joint European Torus (JET). JT-60SA is now the largest operating tokamak in the world and the Joint European Torus (JET) will cease operations at the end of 2023 following 40 years of successful experiments that laid the ground for the physics and technology basis for ITER.


2.     Chair’s report 2023



Switzerland joined the CTP TCP officially as a Contracting Party on 18th October 2022 and the nominated representatives from Switzerland are Ambrogio Fasoli and Paolo Ricci. The IEA CTP TCP Executive Committee approved by the written procedure ending on 31st August 2022 to invite the Government of United Kingdom to join the CTP TCP as a Contracting Party.  Exchanges between IEA and the UK Government are ongoing and the UK have informed the IEA in January 2024 that clearance have been received to proceed with the membership process in order to join the CTP, ESEFP, FM, NTFR, PWI and ST TCPs. The IEA are still waiting to receive the formal paperwork, but the process should be completed soon.



·       14th Executive Committee Meeting at the ITER Organisation Headquarters, St. Paul-lez-Durance, France, Thursday 20th December 2023

·       26th meeting of the International Tokamak Physics Activity Coordinating Committee (ITPA CC)

and 14th ITPA meeting for Joint Experiments at the ITER Organisation Headquarters, St. Paul-lez-Durance, France, 18th-20th December 2023


Participation of the private sector


At the 26th ITPA Coordinating Committee (ITPA CC) and the 14th CTP TCP Executive Committee meeting, the US communicated the interest of private companies to have the option to be involved in the ongoing CTP activities, such as the International Tokamak Physics Activities (ITPA). The modernization of the IEA CTP Implementation Agreement put in place the "limited sponsor" framework, which is a streamlined procedure for involvement by a non-contracting party in one TCP workstream for a period of less than three years. The US suggested to the ITPA Coordinating Committee that a working group be formed to establish an explicit procedure to facilitate private sector involvement in the ITPA. The CTP TCP Executive Committee agreed to follow up the outcome of the Working Group and explore the possibilities within the CTP TCP for utilising the “limited sponsor” framework. The CTP TCP Executive Committee chair and secretary agreed to follow up on these questions with IEA in the near term including at an the upcoming FPCC meeting to be held in Greifswald, Germany.


CICLOP: A Joint Task of the Co-operation on Tokamak Programmes (CTP) and Stellarators and Heliotrons (SH) Technology Collaboration Programmes (TCPs)


The IEA TCP CTP Executive Committee agreed to establish initially the CICLOP Joint Task including the parties that are members of both TCPs (CTP and SH), i.e. initially without the full participation of Korea with observer status and proposed to discuss in the meanwhile the options available to fully involve Korea as soon as possible.


Status report and milestones achieved



The ITER project has continued making steady progress towards first plasma with most plant support systems now in the commissioning phase. The construction of the Toroidal Field coils have now been completed and the construction of the poloidal field coils is now nearly completed. The second module of the Central Solenoid has been installed on the assembly platform and the third module has safely arrived from San Diego. The two power supply systems include the steady-state electrical network, commissioned in January 2019; and the pulsed-power electrical system (sometimes called “reactive power compensation”), for which the equipment is largely installed. ITER’s cooling water systems will be capable of removing more than 1 GW of heat and this equipment installation is complete, and the system is now in pre-commissioning phase. The Cryogenics Plant equipment installation is complete and has entered pre-commissioning phase with the largest portion of the cryogenic pipes now installed. The project for the cold test of the Toroidal Field coils has been launched in 2023. Challenges related to first of the kind components are being addressed including geometric non-conformities found in the Vacuum Vessel sector field joints and thermal shield defects identified due to chloride stress corrosion. A new ITER project baseline is being developed with a stepwise safety demonstration aiming at achieving the earliest start of the ITER nuclear phase minimizing technical risks and early demonstration of tokamak engineering systems with realistic and reliable assembly, commissioning, and operation. 


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)


Joint experiments on JET with high beta long pulse scenarios observed for the first-time flux-pumping following on the previous results obtained at the ASDEX Upgrade tokamak. Joint activities for comparison between equilibrium reconstruction with Japanese and European codes shows a similar behaviour of the plasma shape and centroid evolution on JT-60SA first plasma experiments. Experiments were carried out in WEST on plasma start-up with high atomic number material limiters versus low atomic number material limiters. The limiter phase on Boron-Nitride tiles shows a lower radiated power fraction of ~40% compared with Tungsten tiles of ~80%. Experiments on WEST showed deposits build up on the divertor during a high fluence exposure of 5×1026 D/m2. The deposit build up is mainly on the high field side of the inner strike line on the divertor. These deposits build up leads to events where impurities enter the plasma, hampering long pulse operation and an increased probability of disruptions.




The Integrated Commissioning had been suspended since the incident on an equilibrium control coil (EF-1) in March 2021 and resumed in May 2023. Repair and Risk Mitigation actions following the incident included: repair of the joints and reinforcement of similar structures with pre- and post- confirmation with mock-ups and local/global Paschen tests; modification of the coil power supply to avoid unnecessary voltage application to the coils, e.g. change of the grounding position and reduction of voltage ripples; improvements in monitoring of the cryostat vacuum to stop energisation of the coils when vacuum conditions have deteriorated; and limits on the operation voltage and current in the poloidal field coils.


Following these repairs, the first tokamak plasma of 130 kA was achieved in JT-60SA on 23rd November 2023 and in the following days a diverted plasma of 1.2 MA was achieved. Due to the coils voltage and current limitations, the available toroidal electric field is low, and lower than what is expected in ITER. Therefore, plasma breakdown and ramp-up were challenging. However, preparatory studies using a novel plasma equilibrium simulator, MECS, enabled achieving the first plasma quickly even with the operational limitations. 



Experimental Advanced Superconducting Tokamak (EAST)

Repeatable high-performance plasmas achieved in EAST with duration of 403 seconds with the Tungsten Divertor, fully non-inductive plasma with dominant electron heating with no injected torque, good confinement properties, grassy Edge Localised Modes with frequency above 2.5kHz and stationary divertor temperature. EAST has extended the operation regime closer to high confinement high beta steady state scenarios. EAST experiments demonstrated high confinement, high density, and high bootstrap current towards steady-state operational scenarios in ITER and CFETR and further developed inductive scenarios at q95 ~ 3 in preparation of the ITER baseline scenarios. EAST achieved for the first time Edge Localised Mode suppression by n = 4 resonant magnetic perturbations with relevant ITER baseline parameters of q95 ~ 3, βN ~ 1.8-2.0.


HL-3 Tokamak

Following the first plasma in the HL-3 Tokamak on 4th April 2020, initial research in 2023 achieved plasma current above 1 MA, H-mode discharges, advanced divertor control and snowflake configuration.  The snowflake configuration at a plasma current of 0.5 MA showed a broadened wet area observed in the visible and infrared cameras, with the wider distribution of ion saturation current and heat flux on the dome due to magnetic surface broadening, therefore, effectively reducing the peak heat loads.


HL-2A tokamak

Experimental observations of the bifurcation of turbulence spreading at O-point and X-point of the magnetic island were compared with numerical simulations and it was found that the vortex flow associated with the magnetic island plays an essential role in inducing turbulence phase locking.



Korean Superconducting Tokamak Advanced Research (KSTAR)

In KSTAR, the new tungsten divertor was installed after 3 years of preparation together with start-up scenario modifications with changes in the poloidal field coils bias voltages, plasma control systems upgrades with a new strike point control segment and an update on strike point control. The goal of achieving 50 seconds operation at high beta has been guided by predictive modelling, with plasma pressure feedback control in the Neutral Beam auxiliary heating, and additional Electron Cyclotron and Neutral Beam heating for improved MHD stability. Operation at βN > 3 was achieved for 15 seconds with fully non-inductive operation in the upper single null carbon divertor and for 50 seconds in the lower single null tungsten divertor.


United States

Doublet III D-shaped Tokamak (DIII-D)

DIII-D completed in 2023 a 40-week run period with many results obtained during a negative triangularity campaign. Key results in preparation of ITER include operation at low power compared with the power required for the L-H transition, QH regimes, tungsten dynamics studies, the use of novel alloys and the use of artificial intelligence driven plasma control.


National Spherical Torus Experiment Upgrade (NSTX-U)

In NSTX-U the recovery project is now 80% complete with the prototyping of the new central magnet fabrication process ongoing in parallel with the conductor fabrication. The next steps include installing the new plasma facing components on the casing and the delivery of the new bundle which is forecasted for 2024.  




ADITYA-UG experiments include runaway mitigation by Neon impurity seeding as the Neon puffing increases the electron density which increases the required Dreicer electric field substantially reducing the runaway electron energy. Wall cleaning by electron cyclotron resonance sweep was performed in ADITYA-UG showing potential for localised wall cleaning of plasma facing components and wall coating with homogenous plasma combined with ion cyclotron resonant heating. ADITYA-UG studied Argon transport by simulating the measured emissivity in the visible and ultra-violet range using the code STRAHL, and intrinsic toroidal rotation of Carbon ions measured by Doppler shift spectroscopy using a high-resolution multi-track spectrometer.


Steady State Superconducting Tokamak (SST-1)

SST-1 could not operate in 2023 due to further hardware failures.


Future plans


HL-3 Tokamak

HL-3 tokamak integrated research plan aims at operating above the current of 2 MA and at ion temperature of 10 keV in advanced scenarios with heat, particle and configurations control. The extended research plan foresees the development of fusion reactor grade plasmas with the triple product n t T >  1020 m-3 keV s for fusion reactor related integration and critical component tests. 


Joint European Torus (JET)

With the end of operations in December 2023 and the start of decommissioning from 2024, JET studies focus on laser induced break down spectroscopy and post-mortem analysis of wall tiles with more focus on Tungsten surfaces.


ASDEX Upgrade

Extension and maintenance work is ongoing at ASDEX Upgrade with the installation of an upper divertor with two in-vessel coils and a cryopump for alternative divertor configurations and several new diagnostics; and neutral beam injection upgrades by changing box I from arc to radio frequency sources, and box II equipped with variable gaps for the decoupling of acceleration voltage and power.





Korean Superconducting Tokamak Advanced Research (KSTAR)

The main short-term goals for the experiments at KSTAR are to reproduce the plasma performance of the Carbon divertor with the Tungsten divertor and extend the plasma performance to longer pulses. A major upgrade in 2024 includes the installation of full tungsten plasma facing components.


Doublet III D-shaped Tokamak (DIII-D)

DIII-D will operate in 2024 during the Spring and Summer with plans to exploit a new divertor and a new lower hybrid current drive system.  The divertor allows for larger, more strongly shaped plasmas that operate on the stability boundary for peeling modes, while the lower hybrid system will inject power from the high field side of the plasma where the edge plasma turbulence is significantly reduced.  Several joint international task forces will focus on long pulse compatibility with tungsten walls and high beta poloidal scenarios.