IEA Technology Collaboration Programme on Tokamak Programmes (CTP TCP)

Annual Briefing 2019


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.


2.     ChairŐs report 2019


The participation of Thailand is under discussion and observers from Thailand participated in the 10th Meeting of the Executive Committee 



á       10th Executive Committee Meeting ITER HQ, France, Thursday 12th December 2019

á       22nd meeting of the International Tokamak Physics Activity Coordinating Committee (ITPA CC)

and 10th ITPA meeting for Joint Experiments ITER HQ, France, 10-12 December 2019

á       KSTAR Conference 20-22 February 2019, Coex, Seoul, Korea

á       PPPL Workshop on Theory and Simulation of Disruptions Princeton Plasma Physics Laboratory, Princeton, New Jersey 5-7 August 2019


Three projects have been submitted to IEA regarding the Today in the Lab - Tomorrow in Energy? Initiative on: JET TAE Amplifier and Control Upgrade; Event Detection Intelligent Camera (EDICAM) for the JT-60SA Tokamak and JET Shattered Pellet Injection for Disruption Mitigation in ITER. 



Status report and milestones achieved



The project continued to advance well in 2019 with 73% of construction (within First Plasma scope) completed on the platform and 67% of total First Plasma scope achieved (design, manufacturing, transport assembly).  Impressive progress has been made on the magnets, with two of the Toroidal Field coils nearly complete (Japanese coil 12 and EU coil 9), the 17m diameter Poloidal Field coil 5 in the final stages of completion in the on-site poloidal field winding facility, and the first coil to be installed in the tokamak pit, PF coil 6, complete and on the point of shipping from ASIPP in Hefei, China. The cryostat lower cylinder is complete and awaiting installation, with the base and upper cylinder in the final stages of assembly. In Korea, at Hyundai Heavy Industries, the first sector (#6) of the nine sector, 20 m high vacuum vessel is nearing completion and is expected on-site at ITER in the Spring of 2020. The ITER Organization underwent a very significant reorganization in 2019, with a new structure, better aligned to machine assembly and preparations for First Plasma, to be in place by 1 January 2020. Inside that structure, the ITER Science Division was reorganized to allow greater focus on First Plasma, preparations for the early years of non-active operation and an intense effort towards development of a disruption mitigation system based on shattered pellet injection.  This was accompanied by a strong drive to provide a new focus for the International Tokamak Physics Activity (ITPA), concentrating on R&D priorities for the ITER Research Plan, completely revised in 2018 to be aligned with the new Staged Approach for ITER operation.  A great many of these priorities are common to the efforts of the CTP-TCP programme partners. 



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


JET Experiments have begun to explore the high-performance domain at high power, where fusion power, impurity behaviour and MHD stability will be optimised in preparation for the DT experiments.  In these experiments an increase of performance in the neon seeded scenarios was observed. An ITER like tokamak agnostic control system was implemented on TCV and AUG and real-time handling of the H-mode density limit disruption path was achieved on AUG and TCV with the discharge trajectory followed in real-time in confinement and edge density space. Central heating was applied if the discharge approaches the unstable area and with similar stability boundary on AUG and TCV. Small Edge Localised Mode scenarios on AUG and TCV have shown the importance of magnetic shear at the magnetic separatrix and it was shown that the baffled TCV divertor increases the divertor neutral pressure and extends the detached regime, confirming model predictions. Experiments with the Shattered Pellet Injector on JET for ITER were carried out after successful commissioning of the system developed in collaboration with US and ITER. Deuterium, Neon and Argon pellets as well as Deuterium and Neon and Deuterium and Argon mixtures were used with all three barrels. Run-away electron beam suppression with shattered pellet was demonstrated and unexpected results were obtained with 12mm Deuterium pellets leading to suppression of fast electrons. More than 4 weeks of Helium plasmas were executed in the WEST tokamak equivalent to 2000 seconds of Helium plasma exposure and a fluence of 2.5 1025 He/m2.



Japan Torus 60SA

The fabrication of most of the key components has been finished. Assembly of JT-60SA is approaching the final stage towards its completion and the start of the integrated commissioning and the first plasma is foreseen in 2020. Assembly of the lower part of the cryostat has been finished. Thermal shields have also been assembled. Preparation and test of subsystems, power supplies, vacuum systems, cryogenic systems, heating systems, diagnostics, are proceeding towards completion. Work procedure documents of the integrated commissioning for each subsystem and plasma operation have been established. Lively discussion between Japanese and European physicists have been on going. The 8th Research Coordination Meeting (RCM-8) was held at QST Naka on 24 - 26 June 2019, in which about 60 experts participated including a representative from ITER, for the discussion on the research items on JT-60SA.




Experimental Advanced Superconducting Tokamak (EAST)

Fully non-inductive operation, steady-state scenarios using pure Radio Frequency heating have been developed in EAST as well as steady-state grassy Edge Localised Mode regimes and steady-state scenarios with detachment feedback control. Other techniques to control Edge Localised Modes included suppression by Boron Injection without confinement degradation and Resonant Magnetic perturbations in ITER relevant low safety factor - type I Edge Localised Modes regime. 


HL-2A tokamak

The mission of 2019 HL-2A experimental campaign addressed critical physics and technology issues in high beta and H-mode plasma scenarios. Plasmas with _N>3 operation have been achieved with high Neutral Beam Injection and Radio Frequency power for the first time. Type I Edge Localised Modes mitigation and control have been achieved by mixture of Super Molecular Beam Injection, Resonant Magnetic Perturbations and Lower Hybrid Waves.  In the low central magnetic shear configuration, fishbones plays an important role in the formation and sustainment of Internal Transport Barriers. It was observed that turbulence and turbulence spreading across the magnetic island are modulated by the rotation of tearing modes. Partial detachment was achieved in H-mode plasmas with Neon gas seeding.




HL-2M tokamak

The mission of the HL-2M tokamak, under construction by the Southwestern Institute of Physics (SWIP), is to develop the physics basis for advanced plasma scenarios with 3MA plasma current, high beta (_N > 3), high elongation (_ = 1.8-2), high triangularity (_ > 0.5), and snowflake divertor in support of future fusion machines (ITER, CFETR). The goal is to have the first plasma in May 2020. A several-stage development plan for the heating, control and diagnostic systems has been formulated. The physics of high confinement scenarios, advanced divertor configuration, heat and particle exhaust physics through active control of instabilities and disruption will be studied in detail.



Korean Superconducting Tokamak Advanced Research (KSTAR)

In KSTAR, important steps for advanced scenario development have been achieved with early divertor configuration access, high safety factor and high internal inductance. Long-pulse hybrid scenarios were explored in pulses of up to 9 seconds and double null plasmas were studied with significant increase in the stored energy. Disruption prediction and avoidance research on KSTAR is moving towards real-time applications. Nonlinear MHD modelling has reproduced the key elements of the Edge Localised Mode suppression observed by Electron Cyclotron Emission imaging. The measured electron temperature profile corrugation has been reproduced in Gyro-kinetic modelling, where the Zonal Flow staircase is the key mechanism. Dual Shattered Pellet Injector experiments have been performed with the two injectors in opposite sides of the tokamak allowing the studying of disruption speed dependence on impurity contents and radiation efficiency using a high-speed camera.

Spherical Torus at Seoul National University (VEST)

Plasma currents up to 170 kA have been achieved in VEST and an accelerated rotation in the opposite direction of the plasma current has been observed after strong ion heating during an Internal Reconnection event.


United States

Doublet III D-shaped Tokamak (DIII-D)

DIII-D successfully completed 12 run weeks in 2019 following the installation of worldŐs first toroidally steerable, off-axis neutral beam injector and demonstrated the top-launch electron cyclotron current drive system which works as intended. The helicon antenna modules have been fabricated and the design of the high-field-side lower hybrid system has been completed.


National Spherical Torus Experiment Upgrade (NSTX-U)

The NSTX-U Recovery Project is ongoing. Key project elements include replacement of the six inner-Poloidal Field magnets, numerous improvements to the structural and vacuum properties of the vacuum boundary, improve high heat flux plasma facing components, improvements to the test cell shielding, a new test cell access control system. The Project has a DOE-approved cost and schedule, having completed a successful Integrated Project Review in August 2019. The design phase of the Project is near complete, with the last Final Design Reviews to occur in the winter of 2020. Numerous components are under fabrication as approved long-lead procurements. These include six new inner-Poloidal Field magnets, as well as a new centre-stack casing; these are the two largest single fabrication activities on the Project. Improvements to the test cell nuclear shielding are in progress. All other major procurements and fabrication will be completed in 2020. The Machine reassembly is scheduled to commence in the summer of 2020. The early-finish date for the Project, followed by the resumption of operations, is May 2021. Preparations are underway to re-establish the Research Team and Program.




The ADITYA-UG tokamak achieved 177 kA of plasma current at the maximum toroidal field of 1.4T. The plasma pulse duration was extended to 335 milliseconds in 2019. Experiments with an Electromagnetic Pellet Injector have been performed and 200mg pellet of Lithium Titanate particles were injected during the plasma current flat-top. Pellet injection causes the fast disruption of the plasma current density.  The temperature reduces very rapidly due to increased plasma radiation after the injection.  Significant suppression of Runaway Electron Loss was achieved by suppression of edge density and potential fluctuations using periodic gas puffs.  


Steady State Superconducting Tokamak (SST-1)

Two experimental campaigns were carried out in SST-1 in 2019 performing ECRH assisted Ohmic breakdown studies and lower hybrid current drive experiments. Several technical issues related to cryogenic heat loads has been resolved with the Toroidal Field coils tested to the highest ever field of 2.7 T. Multiple gas puffs showed to suppress hard X-rays generated in low density discharges and recent lower hybrid current drive experiments showed significant improvement in the control of the operating pressure, resulting in the improved efficiency of Lower Hybrid Current Drive.


Future plans


Europe: Key Enhancements for the scientific exploitation in support to ITER are being planned for JET, including a second shattered pellet injector and associated diagnostics and a ITER relevant Laser Induced Desorption Spectroscopy. The participation in the JT-60SA Integrated Commissioning and first plasma operation is being prepared with the nomination of the EUROfusion on-site coordinator and deputies. 


Korea: KSTAR upgrades includes a plan for higher beta and steady-state operation with the installation of an actively cooled Tungsten Divertor and advanced current drive systems. These include the installation of a Helicon Advanced Current Drive System aiming at demonstrating a 0.15 Amperes / Watt of current drive with the injection of 4 MW of auxiliary power.


United States: The 2020 DIII-D program is focused on utilising several Heating and Current drive systems in steady-state scenarios and the installation of the helicon antenna. The NSTX recovery project is aiming at the completion date of July 2022. 


China: In 2020 EAST aims at long-pulse H-mode operation for longer than 400 seconds in a full metal wall environment, in scenarios relevant to CFETR and in support of ITER physics. 

The new tokamak under construction at SWIP, HL-2M tokamak, will have the first plasma in 2020. HL-2A tokamak at SWIP also plans to operate end of 2020, and the experimental compaign will continue into the Spring of 2021.