IEA Technology Collaboration Programme on Tokamak Programmes (CTP TCP)

Annual Briefing 2021


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 in 2021 due to the covid-19 pandemic, but, many activities took place mostly via remote participation tools, covering the participation in the International Tokamak Physics Activities topical group meetings, workshops and joint experiments. These included, the successful remote participation in EAST (CN), HL-2A (CN), KSTAR (KO), DIII-D (US), JET (EU), TCV (EU), WEST (EU) experiments and simulation activities.


2.     Chair’s report 2021


The participation of Thailand is under discussion and observers from Thailand participated in the 10th Meeting of the Executive Committee. However, there were no observers from Thailand at the 12th Meeting of the Executive Committee meeting and that no contacts have been made for a long time regarding Thailand joining the CTP TCP. The discussion on the accession of Thailand will be resumed if further contacts with Thailand take place.      



·       12th Executive Committee Meeting via video conference, Thursday 9th December 2021

·       24rd meeting of the International Tokamak Physics Activity Coordinating Committee (ITPA CC)

and 12th ITPA meeting for Joint Experiments via video conference, 7th-8th December 2021

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





Status report and milestones achieved



The ITER project has continued making steady progress in many areas including machine assembly and manufacturing of the core tokamak components. The construction has continued with the installation and welding of the lower cryostat cylinder, cryostat thermal shields and supports for the Toroidal Field coils. The in-pit installation of the poloidal field coil 5 and 6 was completed. The sector sub-assembly in the assembly hall of the Toroidal Field coils (TF), thermal shields and Vacuum Vessel (VV) started in 2021, with the VV Sector #7 (KO-DA), TF Coil #6 (EU-DA), Central Selonoid Modules #1 and #2 (US-DA), and TF Coils #2 and #10 (JA-DA) delivered to the ITER site and VV Sector #6 (KO-DA) and TF Coils #12 and #13 successfully positioned in the sector sub-assembly. Busbars and piping networks were installed in the tokamak building. The commissioning of the cooling water plant has started and the commissioning of the Cryoplant is near completion. The work on Heat Rejection System, Reactive Power Compensation and Harmonic Filtering System has been completed. Significant progress has taken place in key issues for the successful implementation of the ITER Research Plan including the area of Disruption Mitigation, Strategy for Tokamak Assembly and Strategy for Edge Localised Modes (ELMs) control to ensure adequate divertor lifetime.


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


Operation in the MAST Upgrade Tokamak with the Super X divertor has shown a 10 times reduction in the divertor peak heat flux and a 2 times reduction in the upstream density to detach outer divertors when compared with conventional divertors. Both baseline and hybrid scenarios have reached stationary conditions on JET at high confinement and controlled impurity behaviour in deuterium plasmas. Tritium experiments on JET have showed higher density at the pedestal in tritium plasmas in the baseline scenario with stronger hollow density profile during the H-mode entry in tritium plasmas, earlier density increase, rapid increase in edge radiation suggesting loss of impurity screening, pedestal cooling and compounding loss of screening. The core thermal energy confinement improvement in tritium has been broadly confirmed. High frequency modes, compatible with the Toroidal Alfvén Eigenmode frequency were observed in DT plasmas with no Ion Cyclotron Resonant Heating injection.  Comparison of the pedestal stability on AUG, JET, and TCV has shown consistent dependence of the density pedestal scale length on the separatrix density, that the pedestal width is wider than the EPED1 model assumption and that the pedestal width correlates with the divertor neutral pressure. Hydrogen experiments on AUG have shown that the density pump-out leads to pedestal densities significantly below the ELM suppression threshold. Detachment control schemes were implemented in AUG, JET, and TCV showing the impact of divertor closure on detachment and neutral pressure. A common framework for thermal event detection was developed for WEST, W7-X and ITER with the development of a common machine learning framework for real-time thermal event detection at WEST and W7-X, real-time heat load estimation with Neural Networks and reflection modelling for metallic and non-metallic walls.




After the completion of the construction of the JT-60SA tokamak, the integrated commissioning has started in 2020, after evacuation of both the vacuum vessel and the cryostat, the cool down of the superconducting coils and the thermal shields, and all the coils have reached the superconducting state. Following that, in 2021, the coil energisation tests had started. Energisation of the Toroidal field (TF) coils achieved up to 25.7kA of current. Electron Cyclotron heating at 82GHz and 760kW and Hydrogen gas was injected during Toroidal Field energisation at 25.7kA (2.25T). An Electron Cyclotron Resonant plasma at the fundamental resonance was successfully produced. Energisation of all the poloidal field (PF) coils was done up to ± 5kA. Voltage control tests at ± 5kV was done on all the PF coils but one equilibrium field (EF) coil. The integrated commissioning was suspended due to an incident on that EF coil. Arcing marks were found on the coil joints. It is identified that the cause is due to insufficient voltage holding capability at the joints. Repairs and reinforcement work is under way, not only for the damaged joints but also all the other joints to prevent further incidents. On the basis of the trilateral agreement among the ITER organisation (IO), F4E and QST, information sharing with the IO on the assembly and the integrated commissioning is 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)

The extension of the EAST steady-state Operational regime towards the China Fusion Engineering Test Reactor (CFETR) with relevant performance with enhanced capabilities was achieved with high density, dominant electron heating, zero torque small Edge Localised Modes (ELMs) with Improved confinement.  Experiments in the operational regime towards the ITER baseline-like scenario have showed the importance of the Resonant Magnetic Perturbations (RMPs) to alleviate high-Z accumulation, ELM mitigation and density pumping out. Successful demonstration of active detachment control was achieved with compatible core and edge integration in long pulse H-modes.




HL-2A tokamak

The confinement was improved with Ne and Ar impurity seeding with the decoupling of ion thermal transport contributing to an improved energy confinement. An extended turbulence spreading model with nonlocal effect for edge cooling at high density provided understanding of the edge cooling mechanism in high density plasmas. New electron-Beta Alfvén Eigenmodes driven by counter Lower Hybrid Current Drive was found affecting the parallel momentum transport.


HL-2M tokamak

The preparation for plasma experiments on HL-2M are in progress including the upgrade and installation of the divertor, first wall, cryopump, inner wall Resonant Magnetic Perturbation coils, diagnostics, development of new plasma control system, vertical stability control of elongated plasmas, with experiment campaigns expected in 2022-2023.


Fusion Research in Universities

Particle fuelling by compact torus injection was demonstrated on the Keda Torus eXperiment. Dual shattered pellet injection was successfully demonstrated on J-TEXT in HUST. Successful equilibrium reconstruction at the very beginning of the discharge with 30 MA/s ramp rate in the presence of large eddy currents effects was achieved on SUNIST in Tsinghua University with the response function method.



Korean Superconducting Tokamak Advanced Research (KSTAR)

In KSTAR, Hybrid scenarios are being developed for reactor-relevant operation with q95 < 6.5 in stationary conditions and high performance long-pulse hybrid scenarios are pursued by controlling heating power, plasma shaping and real-time equilibrium correction. Development of the high performance long-pulse discharges included high performance with low qmin discharges with Ohmic or Electron Cyclotron pre-heating frontend, efficient central Electron Cyclotron on-axis Current Drive, good confinement (both thermal and beam) and no sign of performance degradation for long pulses of 9 s.  The longest pulse length of 91.6 s in KSTAR was achieved with 90 s in H-mode. A pre-emptive Resonant Magnetic Perturbation control eliminates the first ELM-crash leading to the longest ELM suppression period of 40 s. The Divertor thermal heat load was managed in ITER-like conditions with Resonant Magnetic Perturbation ELM suppression with control of gas fuelling and stationary detachment was achieved with N seeding and real-time feedback control.  Real time event determination was used for plasma control development, and disruption mitigation actuation.




United States

Doublet III D-shaped Tokamak (DIII-D)

DIII-D completed 18.7 weeks of operation in 2021 with high system availability including Hydrogen experiments and commissioning of the Helicon system. Experiments in DIII-D showed that Helium seeding and n=3 magnetic resonant perturbation fields reduce the L to H power threshold by 30% in hydrogen plasmas with an ITER similar shape.


National Spherical Torus Experiment Upgrade (NSTX-U)

The NSTX-U recovery project is proceeding and is now 70% complete but there are technical issues related to the toroidal field bundle insulation that may delay the project. All production of the plasma facing component tiles have been delivered to PPPL and are being readied for installation. The delivery of the completed centre stack casing is expected in the spring of 2022.




ADITYA-UG continued to carry out important experiments relevant to small tokamaks, including   enhancement of parameters in circular plasmas with a current of 210 kA, duration up to 400 ms and toroidal field of 1.5 T; deuterium plasma operations, 42 GHz Electron Cyclotron Resonant heating two pulse operation, drift Tearing Mode rotation studies, gas-puff induced cold pulse propagation studies, Neon impurity seeding studies and attempts at Divertor operation.


Steady State Superconducting Tokamak (SST-1)

SST-1 was mostly in shutdown and maintenance during 2021. Nevertheless, some novel experiments of plasma start-up were initiated. In SST-1, the central solenoid is located outside the cryostat and therefore, resulting in low loop voltage operations. Therefore, Electron Cyclotron Resonant heating breakdown assist is critical for successful plasma current start-up with toroidal magnetic field Bt < 1.5 T. To operate SST-1 in the 1.5TBt 3T range, an alternative Radio Frequency based plasma current start-up system has recently been developed. This antenna system, made of a series of combinations of two flat spiral antennas has already been installed and tested without a background magnetic field within the frequency range of 35–60 MHz.


Future plans


Regarding the personnel assignments and remote participation in 2022, there is still some uncertainty regarding the possibility of travel, but the focus will remain on the participation in the International Tokamak Physics Activities topical groups and joint experiments. EU participation include joint experiments in DIII-D on the topics of fast particles, runaway electrons, integrated control, resonant magnetic perturbations, radiative scenarios, advanced tokamak physics and negative triangularity; joint experiments in KSTAR on the topics of shattered pellet injection, resonant magnetic perturbations and conditioning, joint experiments in EAST and HL-2A on the topics of current drive and long pulse operation. CN proposals centre on remote participation on the development of hybrid, steady state operation in EAST and studying the scrape of layer width in HL-2A, disruption mitigation experiments in EAST, analysis of the LH-transition and ELM mitigation on EAST. JP proposals focus on the experiments in DIII-D on plasma transport, MHD stability, heating and current drive, disruption mitigation; experiments in JET on MHD, energetic particles, disruption mitigation, edge turbulence; experiments on ASDEX Upgrade on current profile control, particle confinement, disruption mitigation; and simulations on EAST turbulence data. US proposals include the work on disruption mitigation on JET under the continuation of the present Shattered Pellet Injection agreement and IO proposals will continue focussing on Shattered Pellet Injection, Helium and Ion Cyclotron Heating experiments on JET.  AU proposals focus on the participation in KSTAR experiments on energetic particles.