A. Bhattacharjee(a), B. Allen(g), C.-S. Chang(a), H. Chen(f), Y. Chen(f), J. Cheng(f), E. D’Azevedo(b), P. Davis(e), J. Dominski(a), M. Dorf(c), M. Dorr(c), S. Ethier(a), A. Friedman(c), K. Germaschewski(h), R. Hager(a), A. Hakim(a), G. Hammett(a), J. Hittinger(c), S. Janhunen(d), F. Jenko(d), S. Klasky(b), S. Ku(a), R. Kube(a), L. LoDestro(c), N. Mandell(a), G. Merlo(d), A. Mollen(a), M....
A new era of predictive integrated modeling has begun. The successful validation of theory-based models of transport, MHD stability, heating and current drive, with tokamak measurements over the last 20 years, has laid the foundation for a new era where these models can be routinely used in a "predict first" approach to design and predict the outcomes of experiments on tokamaks today. The...
Since the last IAEA-FEC, the EAST research programme has been, in support of ITER and CFETR, focused on development of the long-pulse steady-state (fully non-inductive) high beta H-mode scenario with active control of stationary and transient divertor heat and particle fluxes $^{[1]}$. The operational domain of the steady-state H-mode plasma scenario on EAST has been significantly extended...
Construction of JT-60SA is progressing on schedule towards completion of assembly in March 2020 and the first plasma in September 2020. As of January 2020, manufacture and assembly of all the main tokamak components have been successfully completed satisfying technical requirements including functional performances and dimensional accuracies. Development of plasma actuators and diagnostics is...
DIII-D physics research addresses critical challenges for operation of ITER and the next generation of fusion energy devices through a focus on innovations to provide solutions for high performance long pulse operation, development of scenarios integrating high performance core and boundary plasmas, and fundamental plasma science and model validation. Substantial increases in off-axis current...
We present here recent highlights from Wendelstein 7-X (W7-X), the most advanced and largest stellarator in the world, in particular stable detachment with good particle exhaust, low impurity content, and energy confinement times exceeding 100 ms, maintained for tens of seconds, as well as proof that the reduction of neoclassical transport through magnetic field optimization is successful....
Experiments on ST40 towards burning plasma conditions
M. P. Gryaznevich for TE.Ltd team
Tokamak Energy Ltd, 173 Brook Drive, Milton Park, Abingdon, OX14 4SD, UK
e-mail: mikhail.gryaznevich@tokamakenergy.co.uk
Spherical Tokamak (ST) path to Fusion has been proposed in R Stambaugh et al, Fus. Tech. 33 (1998) 1, and experiments on STs have already demonstrated feasibility of this approach....
Introduction: The stellarator is unique among magnetic confinement concepts in that the plasma performance is mostly determined by externally applied magnetic fields. There is considerable opportunity to improve the stellarator through increased understanding of how 3D fields impact important plasma physics processes, enabling innovation in configuration design. We review recent...
Operating a full tungsten actively cooled tokamak:
overview of WEST first phase of operation
J. Bucalossi and the WEST Team (http://west.cea.fr/WESTteam)
CEA, IRFM, F-13108 St-Paul-Lez-Durance, France.
E-mail: jerome.bucalossi@cea.fr
WEST is a MA class superconducting, actively cooled, full tungsten (W) tokamak. Equipped with two up-down symmetric divertors, it operates at 3.7T, up...
Spherical tokamak (ST) research in Japan [1] is being conducted as a nationally coordinated program of university-scale ST devices under the ST Research Coordination Subcommittee organized by National Institute for Fusion Science (NIFS). The roles of university ST research include: (1) unique and challenging research through creativity and innovation which might be considered too risky for...
The report provides an overview of the results obtained at the upgraded Globus-M2 spherical tokamak [1] since the last IAEA conference. The tokamak was designed to reach the toroidal magnetic field as high as BT =1 T and the plasma current Ip = 0.5 MA having a small plasma minor radius a = 0.22-0.23 m. Currently 80% of highest magnetic field and plasma current value are reached, so during the...
The 2019-2020 scientific and technological programme exploits JET’s currently unique capabilities: Tritium handling and ITER-like wall (ILW: Be wall and W divertor). It is the culmination of years of concerted scientific and engineering work, with the ILW installation in 2010, improved diagnostic capabilities, now fully available, a major Neutral Beam Injection (NBI) upgrade providing record...
KSTAR$^{1,2}$ program has been focused on resolving the key physics and engineering issues for ITER and future fusion reactors utilizing unique capabilities of KSTAR. First of all, a new advanced scenario was developed targeting steady-state operation based on the early diverting and heating during the ramp-up phase of plasma current and significant progress has been made in shape control to...
The ADITYA Upgrade (ADITYA-U) is a medium sized (R0 = 75 cm, a= 25 cm) tokamak having toroidal graphite limiter, configured to attain shaped-plasma operations with an open divertor in single and double-null configurations [1]. The foremost objective of ADITYA-U is to prepare the physics and the technological base for future larger tokamaks by expanding the ADITYA-U operating space and by...
The TCV tokamak continues to leverage its unique shaping capabilities, flexible heating systems and modern control system to address critical issues in preparation for ITER and a fusion power plant. For the 2019-20 campaign its configurational flexibility has been enhanced with the installation of divertor gas baffles and its diagnostic capabilities with an extensive set of upgrades....
Using its unique flexibility and advanced plasma diagnostics, the TJ-II stellarator is contributing to the understanding and solution of critical challenges in fusion plasmas. Next, we highlight some of the most relevant recent results in the framework of its research programme.
Towards validation of gyrokinetic and neoclassical simulations. Aiming at the validation of the instability...
ITER Organization, CS 90 046, 13067 St. Paul lez Durance Cedex, France
Significant progress has been made in the fabrication of the tokamak components and the ancillary systems of ITER and in the finalization of the plant infrastructure at the ITER site since the 2018 Fusion Energy Conference. By an agreed measure, over 2/3 of the work scope required for First Plasma has been accomplished....
Plasmas in the ASDEX Upgrade (AUG) tokamak can match a large number of fusion
relevant parameters simultaneously. With a tungsten wall and ITER-like
magnetic and divertor geometries, high values of the plasma $\beta$, the
normalized confinement time, Greenwald fraction, and power densities $P/R$
are reached under detached divertor conditions. The synopsis first addresses
the integration...
Achieving ignition and high fusion yield in the laboratory is a central goal of the U.S. Inertial Confinement Fusion (ICF) Program. Three major and credible approaches are currently being pursued: laser indirect-drive (LID), laser direct-drive (LDD), and magnetic direct-drive (MDD). While the three approaches use very different means for driving a spherical or cylindrical implosion that can...
As an important part of the fusion research program in China, the key missions of the HL-2A and HL-2M tokamak programs are to explore physics and technology issues and provide research basis in support of ITER and fusion reactors. This overview reports the latest progresses in HL-2A programs, including high performance scenarios for the study of advanced plasma physics, ELM control physics and...
The mission of the spherical tokamak NSTX-U is to advance the physics basis and technical solutions required for optimizing the configuration of next-step tokamak fusion devices, and to advance the development of the ST concept towards a compact, low-cost Pilot Plant [1]. NSTX-U will operate at up to 2 MA and 1 T with up to 10 MW of Neutral Beam Injection (NBI) power for 5 seconds and 4 MW of...
In the recent deuterium experiment on the Large Helical Device (LHD), we have succeeded to expand the temperature domain to higher region both in electron and ion temperatures as shown by the red region in Fig.1. We found a clear isotope effect in the formation of Internal Transport Barrier (ITB) in high temperature plasmas. In the deuterium plasmas, we have also succeeded to realize the...
As of a long-term research program, the J-TEXT [1] experiments aim to develop fundamental physics and control mechanisms of high temperature tokamak plasma confinement and stability in support of success operation of the ITER and the design of future Chinese fusion reactor, CFETR. Recent research has highlighted the significance of the role that non-axisymmetric magnetic perturbations, so...
Inertial confinement fusion (ICF) aims to assemble and confine a dense, high pressure fusion fuel over a relatively short timescale (≪1 μs) compared to magnetic confinement fusion (> 1 s). This is typically accomplished by imploding a spherical capsule at high implosion velocities (>350 km/s) to obtain the fuel temperatures (>4 keV) and areal densities (ρR >0.3 g/cm2) required for...
In order to meet the commitments for the first plasma at ITER, all the domestic agencies are putting in considerable efforts to ensure the manufacturing and delivery of their commitments. Many of them are first of its kind components in terms of the sizes, technologies involved, performance requirements, compliance with ITER’s nuclear safety requirements and the need to survive the lifetime...
The EUROfusion Work Package PFC (Plasma-Facing Components) focuses on critical plasma-surface interaction studies and components qualification in view of upcoming ITER operation and in preparation for DEMO exhaust solutions. This poster gives an overview of the latest main results in WP PFC, as well as their implications for ITER and DEMO.
Helium-Tungsten Interaction
The...
Abstract
The COMPASS tokamak, operated in the Institute of Plasma Physics of the Czech Academy of Sciences in 2009 – 2020, is one of few devices with an ITER-like plasma shape. Its flexibility, extensive set of diagnostics, and NBI heating allow to address key issues in the fusion research in support of ITER and DEMO, such as edge and SOL physics, the L-H transition, runaway...
Since the 2018 IAEA FEC Conference, FTU operations have been devoted to several experiments covering a large range of topics, from the investigation of the behaviour of a liquid tin limiter to the runaway electrons mitigation and control and to the stabilization of tearing modes by pellet injection and electron cyclotron heating. Other experiments have involved the spectroscopy of heavy metal...
Achieving net energy production in magnetic confinement fusion devices is a key milestone in the quest for fusion energy. With the mission of demonstrating net fusion energy, the SPARC tokamak is being designed jointly by the MIT Plasma Science and Fusion Center and Commonwealth Fusion Systems. Its study of reactor-relevant, alpha-heating-dominated scenarios and high power density regimes will...
The ADITYA/ADITYA-U tokamaks are equipped with state-of-art spectroscopic diagnostics in the visible and vacuum ultraviolet (VUV) region of the spectra. These spectroscopic systems are used to study several physics problems in ADITYA tokamak as well as in ADITYA-U, which is an upgraded version of ADITYA, having capability of producing shaped plasmas. The physics studies addressed in this paper...
The Keda Torus eXperiment (KTX) is a new built middle-size reversed field pinch (RFP) device at the University of Science and Technology of China. The mission of KTX is complementary to the existing international Revered Field Pinch (RFP) facilities. The plasma wall interactions, transport in different boundary conditions, the single helicity (SH) state are the main physics aspects of KTX. The...
