TCP on Tokamak Programmes (CTP TCP)
Annual Briefing 2017
The objective of this agreement is to advance the physics and technologies related to toroidal plasmas by strengthening cooperation among tokamak research programmes, to enhance the effectiveness and productivity of the research and development (R&D) effort related to the development of the tokamak fusion concept, to contribute to and extend the scientific and technology database of toroidal confinement concepts, and to provide a scientific and technological basis for the successful development of fusion power. This includes experiments to contribute to the database for the next-generation tokamak devices including, but not limited to: plasma equilibrium and stability; energy and particle transport, plasma heating and current drive; plasma-wall interaction and divertor physics; pedestal physics including edge localised mode (ELM) control; energetic particle-driven instabilities, transport and confinement; integrated scenario development; plasma fuelling; plasma diagnostics; and other areas as mutually agreed.
2. Chair’s report
Governance: After review of the CTP TCP Supporting Documentation and the FPCC Working Party Feedback Form, the CERT approved the request for extension of the CTP TCP for the period 1 July 2017 to 30 June 2022 by written procedure ending 12 June 2017. The CERT's decision was subsequently recorded at the CERT meeting held 19-20 June 2017.
The CTP TCP Executive Committee (“ExCo”) the Contracting Parties (CPs) unanimously agreed to revise the term of the Chair from one year to three years. Tony Donne (EU) was elected as the Chair of the Committee from 1 March 2017 to 29 February 2020. A Vice-Chair will be elected in the last year of the Chair’s term in order to ensure a transition between the outgoing and incoming Chairs. The CTP TCP ExCo unanimously approved the change of the Japanese CP to "QST Japan". The 8th ExCo Meeting was held 9 November 2017 at ITER Headquarters.
Membership: At the 7th ExCo Meeting the CPs agreed to invite Australia to join the CTP TCP. An invitation letter was sent to Australia and an acknowledgement was received from Richard Garret on 12 September 2017. CPs also recognised once more the importance of Russia as a major partner in the ongoing collaborations and agreed that the outgoing Chair is to prepare a new letter of invitation to Russia (Kurchatov Institute) to join the CTP TCP.
Outreach: A representative of the CTP TCP attended the TCP Universal Meeting led by the IEA Executive Director in October 2017. The CTP TCP has generated articles for inclusion in the IEA biennial publication Technology Collaboration Programmes: Highlights and outcomes. Now that the website is active, the CTP TCP plans to announce developments through the OPEN Bulletin. The CTP TCP contributed to several expert events during 2017, including the 8th International Tokamak Physics Activities (ITPA) Joint Experiments Workshop (JEX); the 19th Meeting of the International Tokamak Physics Association (ITPA) Coordinating Committee; the CTP-ITPA JEX Planning Meeting; the KSTAR Conference; and a Princeton Plasma Physics Laboratory (PPPL) Workshop on Theory and Simulation of Disruptions.
Highlights and milestones achieved
ITER Tokamak: ITER construction continues to advance at pace, with the 50% construction completion milestone reached in late 2017, accompanied by a major press release which stimulated very significant interest in the world’s media. Supported by impressive achievements in fusion technology R&D, manufacturing of major ITER components, such as superconducting magnet systems, vacuum vessel and cryostat, is in full swing. A wide-ranging physics R&D programme, covered in many cases by the CTP TCP, is addressing key issues impacting on the finalization of the ITER design and preparations for operation.
EU Tokamaks (JET, ASDEX-Upgrade, TCV, MAST, WEST, COMPASS): Experiments in JET, ASDEX-Upgrade and MAST have developed a multi-machine scaling of the type-I ELM energy flux that was found to be proportional to machine size and pedestal pressure and that demonstrated that mitigated ELMs follow the same scaling. ELM suppression was achieved in ASDEX-Upgrade at low collisionality with high confinement. It was found that triangularity is a key ingredient to access ELM suppression under these conditions. Experiments in the TCV tokamak have studied the impact of the plasma geometry on the divertor power exhaust and SOL physics. The COMPASS tokamak has focused on understanding the importance of error fields using High-Field Side coils in support of ITER’s questions regarding the need for the lower Error Field Correction Coil, experimental study of the existence of eddy current generation during asymmetric Vertical Displacement Events and has been a major player in the investigation of castellated gap edge loading in support of ITER tungsten divertor design decisions. The WEST project that will focus on actively cooled ITER mono-block testing in a plasma environment has been delayed, but sustained plasma pulses were achieved in December 2017.
JT-60SA Tokamak (Japan): The fabrication and installation of components and systems for JT-60SA procured by EU and Japan are steadily progressing towards start of operation in 2020. Up to November 2017, 12 Toroidal Field coils (in total 18) have arrived at Naka from EU and assembled on the tokamak. Manufacture of all the six Poloidal Field coils and 3 Centre Solenoid modules (in total 4) have been completed with excellent accuracy of manufacture. Commissioning of the power supply system integrating EU and JA procurements is progressing.
EAST Tokamak (China): Active loop voltage control was obtained using Lower Hybrid waves, while the plasma current was controlled by the poloidal field coil current. Active feedback control of radiation power aimed at reducing the heat flux to the SOL was achieved, lowering the divertor temperature and peak particle flux effectively with only a slight degradation of plasma confinement. A stationary quasi-snowflake configuration consistent with steep edge density gradients and no ELMs was achieved with a peak heat load reduction with respect to the lower single null. ELMs were also eliminated using the PPPL Impurity Dropper in scenarios with the tungsten divertor.
HL-2A Tokamak (China): The first Experiments were carried out in H-mode with the Passive Active Multi-junction Lower Hybrid Current Drive Launcher, with 900kW of power coupled to an H-mode plasma for 400ms. The pedestal dynamics was studied across the L-H transition showing that the oscillatory flows can quench the edge turbulence and initiate the transition.
KSTAR (South Korea): Significant progress in plasma control was achieved in KSTAR to make a robust start-up. In exploring higher confinement operation, stationary high beta discharges were achieved above the no-wall limit, stationary Hybrid operation was achieved without sawteeth by adjusting the heating time and the plasma current. ELM suppression was expanded to operation at ITER baseline conditions. Castellated tungsten divertor tiles were exposed to higher heat fluxes in the leading edge. Tiles extracted after the 2017 campaign showed deep bulk melting of the first poloidal row due to accidental misalignment.
DIII-D (US): It was demonstrated that multi-mode Resonant Magnetic Perturbations (RMP) lower the threshold current for ELM suppression in DIII-D. The new multi-spectral tailoring of the applied field was made possible by the new power supplies from ASIPP/EAST in China. DIII-D Shattered Pellet Injection (SPI) tests provided insights for ITER Disruption Mitigation, in that a shallow trajectory reduces the SPI effectiveness versus core directed injection.
EU Tokamaks (JET, ASDEX-Upgrade, TCV, MAST, WEST, COMPASS, DTT): Euratom, the ITER Organization and the US Department of Energy are currently working on the installation of a Shattered Pellet Injector at JET, with planned operation starting in the second half of 2018. The present plan is to complete the system installation for JET Experimental Campaign in 2018. The objectives for the operation of JET in 2018 are to prepare scenarios for fusion performance and alpha particle physics, determine the isotope dependence of H-mode physics, SOL conditions and fuel retention and quantify the efficiency of SPI vs MGI on runaway electron and disruption energy dissipation and extrapolate to ITER. The coordinated operation of the Medium Size Tokamaks (MST) devices aims to demonstrate the compatibility of small, no/suppressed ELM regimes for ITER and DEMO, develop and characterize conventional and alternative divertor configurations for DEMO and develop/characterize methods to predict and avoid disruptions as well as control/mitigate runaway electrons and demonstrate their portability. At the end of the year the Czech colleagues obtained a national grant for upgrading COMPASS to a high field tokamak, with nitrogen cooled copper coils. Italy is intending to soon start the engineering design of a superconducting Divertor Test Tokamak.
KSTAR (South Korea): The near-term upgrade and research plan in KSTAR includes long pulse H-mode (>70s), ELM control (>30s) and the investigation of alternative operation modes. Advanced scenario research will be carried out at higher beta using the foreseen heating upgrade for transport and stability control studies.
DIII-D (US): The DIII-D program and facility enhancements aim at addressing key scientific issues for fusion energy, including the preparation of burning plasma scenarios, determining the path to steady state and plasma material boundary conditions. These foresee enhancements of the Electron Cyclotron Current drive with top launchers aiming at doubling the efficiency, Helicon current drive tests at high power and the development of High Field Side Lower Hybrid Current drive.
EAST (China): Future plans include the extension of the operation to demonstrate ITER and CFETR steady-state scenarios aiming at higher plasma pressure long pulse demonstrations (≥ 400 sec) with good confinement and high energy injection (>1 GJ).
Engineering Test Reactor (CFETR), China: The Engineering Test Reactor (CFETR), a device aiming at the demonstration of fusion energy production (50-200 MW for Phase I), of steady-state and high duty factor, 0.3 – 0.5, of tritium self-sufficiency with Tritium Breeding Ratio > 1, exploring options for DEMO blanket and divertor solutions is being developed.