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....
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...
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...
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...
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....
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...
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...
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...
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...
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...
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...
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...
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....
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...
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...
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...
One of the major challenges in magnetic confinement thermonuclear fusion research concerns the confinement, inside the reaction chamber, of the energetic particles (EPs) produced by fusion reactions and/or by additional heating systems, as, e.g., electron and ion cyclotron resonant heating, and neutral beam injection. In such experiments, EPs, having their velocities of the order of the Alfven...
Using modern deep neural network architectures, accurate disruption predictions on the DIII-D tokamak are made possible with the raw data from a single, high temporal resolution diagnostic which contains multi-scale, multi-physics sequences of events. This work illuminates a path forward to meet the disruption prediction requirements of devices such as ITER by using raw diagnostic data...
Recent analysis leveraging the broad array of measurable plasma parameters on the DIII-D tokamak has been used to elucidate the physics underlying detachment processes in the divertor and reveal the 2D nature of detachment important for design of detachment scenarios for next step devices. The dominant role of EUV/VUV radiation for radiative power exhaust has been established experimentally...
Recent tungsten (W) divertor experiments in the DIII-D tokamak have made significant progress elucidating key mechanisms responsible for high-Z erosion, re-deposition, leakage, and scrape-off-layer (SOL) transport. These results have important implications for ITER and other next-step fusion devices, including insight into W sourcing during mitigated edge localized modes (ELMs), diagnosing and...
In tokamak discharges there are often saturated Alfven modes (1). They are destabilized by gradients of the high energy particle population, so are to be expected in discharges with a significant alpha particle population, such as expected in ITER (2) or any fusion device. These modes may produce only a small local flattening of the particle distribution, or if the number and amplitude be...
A portable and interpretable data-driven algorithm for disruption prediction has been developed and installed in DIII-D and EAST Plasma Control Systems (PCS). The Disruption Prediction via Random Forest (DPRF) algorithm guarantees explainable predictions in real-time thanks to the feature contribution analysis [Rea2019], which is able to identify the main drivers of the disruptivity – the...
New studies identify the critical parameters and physics governing disruptive neoclassical tearing mode (NTM) onset. A m/n=2/1 mode in DIII-D begins to grow robustly only after a seeding event (ELM Fig. 1, or sawtooth precursor and crash Fig. 2) causes the mode rotation to drop close to that of the plasma’s Er=0 rest frame; this condition opens the stabilizing ion-polarization current “gate”...
Detachment control tests at DIII-D and EAST have expanded to new sensors and integration with high confinement ($H_{98,Y2}$≈1.5, $\beta_N$=3) core scenarios (see $^1$ for details on core performance). Active detachment control protects the divertor target from extreme heat fluxes and temperatures which might otherwise cause melting and erosion while minimizing fuel or impurity seeding commands...
ELM control and mitigation efforts are of highest priorities for ITER and beyond. However, often such efforts with RMPs and pellets do not focus on the pedestal turbulence and transport and their effect on ELM characteristics. Detailed analysis of pedestal turbulence as a function of the heating mix may provide crucial insights as well as additional handles for ELM control. It is found that...
In this paper we report on experimental and modeling results concerning the energetic particle (EP) dynamics in plasma scenarios with off-axis neutral beam (NB) injection at ASDEX Upgrade (AUG). The tools validated in this processes are applied to selected scenarios at JT-60SA and ITER pre-fusion plasmas.
Off-axis NB injection is an important tool to control and optimise the current profile...
Controlled mitigation of heat load has been demonstrated for the first time by doping a closed divertor$^{1}$ plasma at DIII-D with low Z impurity powders. Injection of low-mid Z impurities is a technique under investigation to address the issue of power exhaust in ITER and next-step fusion reactors$^{2,3}$. The use of non-recycling impurities in powder form provides a new capability extending...
Energetic particles (EP) represent the main source of heat and momentum for burning plasmas. However, EPs can drive instabilities that, in turn, can cause redistribution and loss of EPs. The reduced physics, energetic particle kick model for EP transport enables interpretive and predictive capabilities for time-dependent integrated tokamak simulations including the effects of EP transport by...
The understanding and control of runaway electrons (RE) is a top priority of the nuclear fusion program because, if not avoided or mitigated, RE can severely damage the plasma facing components of a tokamak. Two key open problems are the generation and the impurity-based mitigation of RE. The first problem requires the computation of the production rate of RE. That is, given a plasma state,...
This work reports on a breakthrough on the way to a comprehensive modelling of burning fusion plasmas. For the first time, global electromagnetic gyrokinetic PIC simulations of Alfvénic modes have been successfully performed for a high beta ITER plasma.
This finally gives us the ability to model alpha particle driven modes self-consistently in the non-linear regime and to predict the related...
Recent experiments on DIII-D have utilized the new off-axis neutral beam injection (NBI) power to achieve $\beta_N$ = 3.8 with n = 1 ideal stability limits up to $\beta_N$ = 6. The NBI upgrade adds two additional co-current, off-axis, beams giving a total of 8 MW of on- and 7 MW of off-axis NBI power for advanced tokamak (AT) scenario development in these experiments. In addition, 1.6 MW of...
In this paper the integrated modelling for the steady-state regime of a fusion neutron source DEMO-FNS (R/a=3.2m/1m, B=5T, Ip=4-5MA) [1] is complemented by the helium balance in the divertor and core plasma. The model describes the power and particle balances consistently in the divertor and core plasmas according to approach [2] and finds the condition to fulfill the global requirements on:...
Contrary to previous thinking, recent experiments on DIII-D suggest that the low-frequency instability known as the beta-induced Alfvén-acoustic eigenmode (BAAE)$^1$ will not degrade high-energy fast-ion confinement on future devices. On the other hand, another low-frequency instability, the beta-induced Alfvén eigenmode (BAE)$^2$, interacts strongly with the high-energy fast-ion population...
Unstable Alfvén eigenmodes (AEs) are a key issue in magnetically confined fusion, both for currently operating machines (JET, AUG, etc) and for next step devices such as JT60-SA and ITER, due to their potential to cause energetic-ion (heated by NBI/ICRH or fusion alphas) redistribution and losses [1,2]. Toroidicity-induced AEs (TAEs), resulting from the coupling of two shear-Alfvén waves, are...
Energetic particle physics is a crucial issue in burning plasmas such as the International Thermonuclear Experimental Reactor (ITER). Instabilities driven by energetic particles, such as fishbones and various Alfvén eigenmodes, can induce the transport and loss of energetic particles, degrade fast particle confinement, and even lead to serious wall damage. A non-monotonic safety factor...
Hybrid simulations for energetic particles interacting with a magnetohydrodynamic (MHD) fluid were conducted using the MEGA code [A, B] to investigate the spatial and the velocity distributions of lost fast ions due to the Alfvén eigenmode (AE) bursts in the Large Helical Device (LHD) [C, D]. It is found that the spatial distribution of lost fast ions in the divertor region during the AE burst...
Experiments on DIII-D and C-Mod show that high neutral opacity is compatible with a steep density gradient at the plasma edge [1,2]. Future reactors, including ITER, will operate at high neutral opacity, which will strongly limit direct fueling of the pedestal structure inside the Last Closed Flux Surface (LCFS) through ionization from edge sources in comparison with current existing...
Impurity seeding studies were performed for the first time in the slot divertor at DIII-D, showing that with suitable use of radiators, full detachment is possible without degradation of core confinement. First ever multi species SOLPS-ITER simulations including full cross-field drifts and neutral-neutral collisions in DIII-D demonstrate the importance of target shaping and plasma drifts on...
Improving Energetic Particle Confinement in Stellarator Reactors
A. Bader$^1$, M. Drevlak$^2$, D.T. Anderson$^1$, C.C. Hegna$^1$, S.A. Henneberg$^2$,
T.G. Kruger$^1$, A. Ware$^3$
1: University of Wisconsin-Madison, WI, USA,
2: Max-Planck Institut fur Plasmaphysik, Greifswald, Germany,
3: University of Montana, MT, USA
Energetic particle confinement is a key issue for the scalability...
A novel control architecture for simultaneous regulation of several plasma scalar variables, such as the thermal stored energy ($W$), the bulk toroidal rotation ($\Omega_\phi$), and the safety factor at various spatial locations (e.g., $q_{95}$, $q_0$), and for active suppression of Neoclassical Tearing Modes (NTMs) by means of ECCD {1} is shown to improve plasma performance even in the...
Internal magnetic fluctuation measurements identify magnetic turbulence in the DIII-D ELMy H-mode pedestal as micro-tearing modes (MTM) and mode growth accompanied by degraded plasma confinement is observed. This work provides the first direct measurement of internal magnetic fluctuations supporting the prediction of gyro-kinetic simulations$^1$ that MTM exist in the H-mode pedestal. Using a...
New DIII-D results may explain why achieving ELM suppression by resonant magnetic fields (RMPs) remains elusive in double null (DN) diverted configurations: the lack of ELM suppression in DN correlates with a damped high-field side response of field-aligned structures that could be indicative of a missing resonant tearing needed to stop inward growth of pedestal. This is found despite...
Improved understanding of the mechanisms that govern thermal transport in the pedestal region is crucial for determining the fundamental processes behind the L-H transition and pedestal structure, and providing a foundation for predicting and optimizing the pedestal and performance of future devices such as ITER. We report world first inferred ion and electron heat fluxes in the pedestal...
The radial width of heat flux flowing into the DIII-D divertor is found to expand beyond that of the established empirical scaling (1) for conditions of high input power and high plasma density. This expansion is consistent with a scrape-off-layer (SOL) radial pressure gradient limited by the MHD ballooning stability limit, but does not inherently result in a degradation of edge pedestal...
New power law scalings of the error field (EF) penetration thresholds across a wide range of tokamaks have been developed for toroidal mode numbers n=1 and 2 and project values for ITER that the construction tolerances and correction coils satisfy. This paper presents a multi-variable n=2 threshold regression across a wide range of densities, toroidal fields, and pressures in 3 machines...
The DIII-D tokamak has developed a new regime for high-beta hybrid plasmas where the broad current profile is achieved with strong off-axis electron cyclotron current drive (ECCD) rather than anomalous poloidal magnetic flux pumping. The high-beta hybrid regime with $q_{min}$ slightly above 1 and without sawteeth is a candidate for the $Q=5$ steady-state scenario on ITER$^{1-3}$, but the...
Energetic particle (EP) instability models based on gyro-Landau closure techniques (1) have addressed important nonlinear simulation and linear stability survey challenges that will be critical for the understanding and control of burning plasmas in ITER and the next generation of fusion systems. The long-term intermittency and frequency spreading characteristics of saturated Alfvén...
Energetic-particle (EP)-driven instabilities such as Toroidal Alfvén Eigenmodes (TAEs) can be responsible for the effective ion heating via collisionless EP energy channeling. Although the quantitative estimation of the EP transport by instabilities has been actively conducted [1–3], studies on the energy channeling have been limited [4,5]. It is important to estimate collisionless energy...
In this work we show that nonlinear MHD plasma response simulations are essential for understanding and predicting accurate heat and particle flux striations in DIII-D during ELM suppression. Understanding the nature of heat and particle distributions on the divertor plates due to splitting of the separatrix by 3D magnetic perturbation in RMP ELM suppressed discharges is an important issue for...
Energetic particles (EPs) including fusion-alpha particles related physics are expected to play important roles in magnetic confinement fusion devices as EPs contribute significantly to the total power density [1,2]. In particular, two important aspects are heating of thermal plasmas and excitation of symmetry breaking collective modes, e.g., shear Alfvén wave (SAW) instabilities. SAWs could...
The transport consequences of the nonlinear trapping in wave-particle interactions, including collisions, in tokamaks are investigated for the first time. The perturbed distribution is flattened in the vicinity of the resonance by the nonlinearly trapped particles. Particles trapped or barely circulating diffuse radially as a result of collisions. The transport fluxes, scale as the square...
Novel internal measurements and analysis of ion cyclotron frequency range fast-ion driven modes in DIII-D are presented which advance understanding of the dynamics controlling mode stability and thereby the physics basis for prediction of fast-ion (e.g. alpha) transport in burning plasmas. Observations, including internal density fluctuation ($\tilde{n}$) measurements obtained via Doppler...
Runway electrons (REs) [1] in a tokamak is of great concern irrespective of the size of the machine. Such runway electrons carries significant amount of plasma energy of several MeV can severely damage the first wall and in-vessel components of the tokamaks [2] as well as can interfere with the complex plasma phenomena like plasma equilibrium, MHD instabilities and plasma disruption. Therefore...
Tokamak devices aim for magnetically confined burning plasmas in order to reach steady state operations and produce economically exploitable fusion energy. One of the main issues are the strong levels of transport due to the highly nonlinear turbulent plasma behaviour, which causes an increment of heat fluxes with respect to neoclassical theory. It is believed that microturbulence,...
Off-axis Neutral Beam Current Drive (NBCD) physics has been tested on DIII-D for Advanced Tokamak (AT) operation with increased off-axis injection power ($P_{OANB}\simeq7$ MW) by using the newly available, toroidally steerable co/counter off-axis neutral beam (CCOANB) injection capability. DIII-D experiments confirm that the new CCOANB drives current as predicted by the classical model NUBEAM...
Introduction
A powerful method for diagnosing runaway electrons in tokamak experiments is to measure the synchrotron radiation that the relativistic electrons emit. The radiation intensity depends on the electron energy, pitch angle as well as the background magnetic field, and using a spatially resolved diagnostic, such as a camera, allows the spatial distribution of electrons to be...
Background. Alpha channelling [1] is a mechanism to deposit the energy of the fusion-generated alpha particles directly into the bulk ion population through wave-particle interaction. The alpha-channelling mechanism relies on the interaction between the fusion alphas and a high-frequency wave (typically an ion Bernstein wave (IBW) obtained via mode conversion of a Fast Wave injected by an...
Initial experimental evidence shows that the L-H transition power threshold P_th can be reduced via Neoclassical Toroidal Viscosity (NTV) associated with applied n=3 non-resonant magnetic fields (NRMF) [Figs. 1 and 2], and, independently, via a fast reduction or reversal of intrinsic edge toroidal (co)rotation [Fig. 3]. It is also demonstrated that a small/moderate increase in lower...
Measurements of pellet-triggered edge localized mode (ELM) heat fluxes are presented here from experiments in ITER-relevant low collisionality pedestals (normalized pedestal collisionality $\nu^*_{ped}$ < 1) on DIII-D. These measurements demonstrate a reduction of peak ELM energy fluence at the inner strike point as compared to natural ELMs by as much as $\sim$50%. The inner strike point is...
Reliable whole device modeling (WDM) of present and future burning plasmas critically depends on correct treatment of the auxiliary heating by energetic particles (EP) introduced into the plasma externally or by the fusion alpha particles. These energetic, or super-thermal, ions can effectively resonate with Alfvénic plasma oscillations, lead to EP losses and modify the profiles of the...
Introduction: Disruptions due to tearing mode locking are one of few potential obstacles remaining for successful ITER fusion reactor and beyond. Here, we report the experiments of locking avoidance, but also contributing to the H-mode recovery and sustainment by slowly rotating edge-localized tearing mode (TM) layers by applying 3D external field. The process is expected as a non-linear...
The Kinetic Orbit Runaway electrons Code (KORC) {1} has been extended to model post-disruption runaway electron (RE) dissipation by impurity injection incorporating state-of-the-art collisional models for partially ionized impurities {2} and models of thermal electron and impurity spatiotemporal dynamics. We fit these models to data from the DIII-D and JET tokamaks, exploring the role of...
We report the achievement of a world unique capability of high power co/counter steerable off-axis neutral beam injection on a major tokamak, which widens the broad pressure and current profile parameter space for high beta steady-state advanced tokamak (AT) scenarios on DIII-D, while retaining the ability to balance the injected torque for low rotation studies. The unique steering capability...
Analysis of “super H-mode” experiments on DIII-D shows that high rotation, not high pedestal, plays the essential role in achieving very high energy confinement quality ($H_{98y2}$ > 1.5) $^1$. While the stored energy increases as expected with higher pedestal, the energy confinement quality mainly depends on the toroidal rotation (figure 1). At moderate rotation, similar to levels expected in...
The high-power helicon antenna has been extensively tested at low power with excellent results, and the full installation in the DIII-D vessel was successfully completed in February 2020, as shown in Fig. 1. High-power striplines feed the antenna from either end (not visible in Fig.1). Commissioning of the 1.2 MW 476 MHz klystron source and associated high voltage power supply and switching...
Understanding impurity transport within fusion research plasmas is of critical importance as progress is made toward burning plasmas. High core impurity concentration can have a deleterious effect on plasma performance: fuel dilution (for fully stripped low Z impurities), triggering of MHD instabilities, or increased core radiated power (for high Z impurities), thus hindering the achievement...
We present an overview of theory and simulation of low-frequency drift Alfvén waves (DAW) in toroidal fusion plasmas based on the framework of the general fishbone like dispersion relation (GFLDR) [1, 2, 3]. In addition to recovering various limits of the kinetic MHD energy principle, this approach can also be applied to general e.m. fluctuations characterized by a broad range of spatial and...
Energetic particle (EP) physics in fusion research are excepted to play crucial roles in the next generation of tokamak burning plasma experiments, e.g., ITER. Energetic fusion alpha particle heating of fuel ions through collisional and collisionless channels is crucial for achieving self-sustained burning. On the other hand, free energy associated with EPs pressure gradient, can drive...
The (2,1) neoclassical tearing mode (NTM) has been proposed as a candidate to explain the larger than expected losses of high energy ions produced by neutral beam injection observed in DIII-D and ASDEX-U [1-4]. Although the numerical simulations performed so far to study the effect of NTMs on energetic ions have reproduced some features of the experimental results, the situation is not...
Turbulence driven shear flow through Reynolds stress associated with the coexistence of multiple edge instabilities lowers the L-H power threshold ($P_{LH}$) across multiple parameters on DIII-D: $q_{95}$ (Fig. 1), ion ∇B drift direction (Fig. 2), plasmas with and without resonant magnetic perturbation (RMP) (Fig. 3), as well as ion isotope mass {1}. Application of RMP raises turbulence...
Future DT operation in ITER and DEMO will face a significant number of challenges. From the physics point of view, the change from DD to DT plasmas is poorly understood. There are indications that the core confinement, the ELM behavior, the pedestal confinement and the Scrape of Layer behavior can significantly change from DD to DT. From the operational point of view, scenarios with high...
1.Introduction
The inductive goal of ITER is to produce 500s long burning plasmas with $Q=P_{fus}/P_{aux}\geq$10[1]. This requires the development of operationally robust scenarios that span the whole plasma discharge from start-up to termination not only in Deuterium Tritium (DT) but also in the Pre Fusion-Plasma Operation (PFPO) phase in Hydrogen (H) and Helium (He). In the PFPO phase,...
Recent EAST experiment has successfully demonstrated long pulse steady-state high plasma performance scenario with core-edge integration since the last IAEA in 2018 $[1]$. A discharge with a duration over 60s with $\beta_P$ ~2.0, $\beta_N$ ~1.6, $H_{98y2}$~1.3 and internal transport barrier on electron temperature channel is obtained with multi-RF power heating and current drive, i.e. ~2.5 MW...
Nonlinear two-fluid MHD simulations reveal the role of resonant field penetration in ELM suppression and density pump-out in low-collisionality ITER-Similar-Shape (ISS) plasmas in the DIII-D tokamak$^1$. The operational window for ELM suppression in DIII-D ISS plasmas coincides with calculations for magnetic island formation at the pedestal top ($n_e<3\times10^{19}m^{-3}, B_r/B_t >10^{-4},...
Edge Localized Modes (ELMs) comprising of repetitive plasma eruptions from the edge of a tokamak plasma are a very common feature of high confinement mode (H-mode) operation in advanced tokamak devices. Each ELM outburst is associated with an expulsion of a large amount of energy and particles in a very short time that can potentially cause serious damage to plasma facing components and...
EAST is typically operated on the radio-frequency (RF) waves heating scenarios, which is also highly related to ITER and CFETR. For the recent experiments of EAST H-mode operations, RF waves are found to be efficient to mitigate and suppress the edge instabilities, such as ELMs [1,2]. From the simulation point of view, the RF heating and driving effects can be studied from 2 aspects: directly...
E-mail: ysna@snu.ac.kr
Edge Localized Modes (ELM) are rapid MHD events occurring at the edge region of tokamak plasmas, which can result in damages on the divertor plates. Therefore, to fully suppress ELM via resonant magnetic perturbation (RMP) [1-4] is of great help to reach and sustain high-performance H-mode plasmas. It was found that certain conditions must met for the RMP-driven ELM...
Highlight of this work: This work predicts the optimal coil phasing, semi-empirical threshold coil current and ‘favorable’ $q_{95}$ window for ELM mitigation for HL-2M 1MA discharge scenario. It is found that pressure gradient may play an important role on determining the peeling-tearing displacement near X-point, due to the curvature effect (GGJ effect) of equilibrium magnetic field....
A significant advancement of RF modeling was achieved by realizing for the first time the entire 3D full torus plasma simulation with detailed 3D realistic antenna and SOL plasma. Many experiments in different fast wave (FW) heating regimes, such as hydrogen minority heating and high harmonic fast waves (HHFW), have found strong interactions between RF waves and the SOL region. In a...
Access to high current (IP >~ 1 MA) relativistic electron (RE) beams in the DIII-D and JET tokamak reveal excitation of current-driven (low safety factor) kink instabilities that promptly terminate the RE beam on an Alfvenic time-scale [Ref. 1], a phenomenon first observed during the early JET carbon wall operation period. Unlike past results however, this phenomenon when combined with the...
Joint research on the tokamaks DIII-D and EAST demonstrates a successful integration control of divertor detachment with excellent core plasma performance, a milestone towards solving the critical Plasma-wall-interaction (PWI) issues for ITER and future reactors. DIII-D has achieved actively controlled fully detached divertor with low plasma temperature ($T_{e,div} \le $ 5 eV across the entire...
Novel disruption prevention solutions spanning the range of control regimes have been developed and tested on DIII-D to enable ITER success. First, the disruption risk during fast, emergency shutdown after large tearing and locked modes can be significantly improved by transitioning to a limited topology during shutdown. More than 50% of limited shutdowns reach a final normalized current $I_N$...
New scaling laws and modeling, developed at DIII-D and benchmarked with data from JET and KSTAR, provide a path for projecting Shattered Pellet Injection (SPI) performance to ITER, while improved understanding of higher-order effects such as asymmetries better constrain the expected behavior. In the limit of radiative shutdown by high-Z impurity injection, the volume-averaged performance of...
ITER adopts a strategy that distributes radiated power evenly during the disruption mitigation and reduces the time to prepare pellets, using simultaneous multiple shattered pellet injections (SPIs)$^1$. However, since there were no existing devices with perfectly symmetric SPIs, as planned in ITER$^2$, sufficient studies have not been conducted on the effects of simultaneous multi-injections....
Diverted discharges at negative triangularity on the DIII-D tokamak (figure [1]a) sustain normalized confinement and pressure levels typical of standard H-mode scenarios ($H_{98,y2}\simeq 1$, $\beta_N\simeq 3$) without developing an edge pressure pedestal (figure [1]b), despite the auxiliary power far exceeding the L→H power threshold expected from conventional scaling laws. The power...
Recent EAST experiment has successfully demonstrated long pulse steady-state high plasma performance scenario with core-edge integration since the last IAEA in 2018 $[1]$. A discharge with a duration over 60s with $\beta_P$ ~2.0, $\beta_N$ ~1.6, $H_{98y2}$~1.3 and internal transport barrier on electron temperature channel is obtained with multi-RF power heating and current drive, i.e. ~2.5 MW...
Next step fusion devices such as ITER will need a reliable method for controlling the quasi-periodic expulsion of a large amount of heat and particles onto the plasma-facing components caused by edge-localized modes (ELMs). Several options are currently being considered to achieve the required level of ELM-crash control in ITER; this includes operation in plasma regimes which naturally have no...
Full suppression of Edge Localized Modes (ELMs) by using n=4 resonant magnetic perturbations (RMPs) has been demonstrated for ITER for the first time (n is the toroidal mode number of the applied RMP). This is achieved in EAST plasmas with low input torque and tungsten divertor, thus also addressing significant scenario issues for ITER. In these conditions energy confinement does not drop...
1.Introduction
The inductive goal of ITER is to produce 500s long burning plasmas with $Q=P_{fus}/P_{aux}\geq$10[1]. This requires the development of operationally robust scenarios that span the whole plasma discharge from start-up to termination not only in Deuterium Tritium (DT) but also in the Pre Fusion-Plasma Operation (PFPO) phase in Hydrogen (H) and Helium (He). In the PFPO phase,...
Recent experiments in the DIII-D tokamak have shown that a broadened fast-ion pressure profile enables better control of Alfvén Eigenmodes (AEs), improves fast-ion confinement, and allows access to new regimes. New discharges reach 15% higher normalized plasma beta ($\beta_N$) than previously achieved in steady-state scenarios with negative central shear and $q_{min}>2$ at high field...
ITER high Q operation requires the integration of high performance plasmas with high plasma density, good fast ion confinement with acceptable stationary and transient power fluxes to plasma facing components (1). To control transients associated with high performance plasmas (Edge Localized Modes or ELMs), ITER is equipped with a set of in-vessel coils that modify the edge magnetic field...
In the study of burning plasmas it is important to understand multi-scale interactions between energetic-particle-driven MHD mode and drift-wave turbulence for establishing good confinement of both energetic particles and bulk plasmas simultaneously. We investigate nonlinear multi-scale interactions between TAE, which is unstable at low $n$, and drift-wave turbulence, which is driven by...
Detailed new H-mode pedestal measurements in the inter-ELM periods of DIII-D discharges find that ion temperature gradient (ITG)-scale density fluctuations (ñ) can explain anomalous ion heat flux (Q$_{i}$) during the ELM (Fig. [1]), followed by Q$_{i}$ becoming neo-classical$^{1}$ until the next ELM, and with trapped electron mode (TEM) and microtearing mode (MTM) like ñ and magnetic...
An accurate and predictive model for turbulent transport fluxes driven by microinstabilities is a vital component of first-principle-based tokamak plasma simulation. However, tokamak scenario prediction over energy confinement timescales is not routinely feasible by direct numerical simulation with nonlinear gyrokinetic codes. Reduced order modelling with quasilinear turbulent transport models...
A predictive 3D optimizing scheme in tokamaks is revealing a robust path of error field correction (EFC) across both resonant and non-resonant field spectrum. The new scheme essentially finds a way to deform tokamak plasmas in the presence of non-axisymmetric error fields while restoring a quasi-symmetry in particle orbits as much as possible. Such a “quasi-symmetric magnetic perturbation”...
A small angle slot (SAS) divertor concept [1] with a closed slot structure and appropriate target shaping in the near SOL has been developed in order to explore a potentially robust boundary solution with acceptable plasma surface interaction which is essential for fusion reactor plasma conditions, in particular for high-power steady-state operation. Recent experimental tests in DIII-D have...
KSTAR has clarified a set of unresolved 3-D physics issues that could be addressed in the ITER-like in-vessel 3-row, resonant magnetic perturbation (RMP) configurations. In particular, considering that one of the most critical metrics of RMP ELM-crash control would require the compatibility with the divertor heat fluxes under the given material constraints, a series of intentionally misaligned...
For the first time, the progress in RF full wave modeling allows for simulating the wave field in arbitrary 3D antenna/first wall geometry together with the scrape-of-layer (SOL) and the entire tokamak/stellarator core plasmas in an integrated manner. Universal observation among many RF heating and current drive (H/CD) experiments in the ion cyclotron (IC), high-harmonic fast waves (HHFW),...
ITB formation due to energetic particles. The performance of present-day and future fusion devices is largely determined by turbulent transport generated by plasma turbulence. Any mechanisms able to reduce the overall radial propagation of energy and particles is, therefore, crucial in view of scenario optimization. This contribution presents numerical results of turbulence suppression by...
Most of high-performance discharges based on the advanced scenario have shown the active generation of the Alfvén eigenmodes (AE), driven by enhanced fast-ion pressure gradient and broad current density profile in the core region$^{1,2}$.
Among various AE control tools, it has been found that the ECCD and ECH are able to mitigate or suppress the toroidal Alfvén eigenmodes (TAE) and the...
The neutron and tritium production during the pre-DT, Pre-Fusion Plasma Operation, (PFPO) phase of ITER need to be quantified in view of the plans for commissioning and operation of the heating systems in hydrogen and helium plasmas, as discussed in the ITER Research Plan (IRP) [1]. In the assessment presented here, we consider a number of tritium and neutron sources of different origins. In...
The primary goal of Tokamak based fusion reactors is to achieve a self-sustained plasma satisfying ignition criterion by maximizing the product of plasma density, temperature and energy confinement time. There are limits on achievable plasma density and temperature due to current driven or pressure driven instabilities. Also, the Tokamak performance can be enhanced by operating with advanced...
The steady-state superconducting tokamak (SST1) [ref1], having major and minor radius of 1.1m and 0.2 m respectively, equipped with a Copper-based central solenoid to provide the required loop voltage. Like in other superconducting tokamaks [ref2-3], SST1 tokamak also relies on ECR breakdown technique to form Ohmic plasmas. Plasma preionization and current start-up is assisted by the 42 GHz...
Regulation of the $q$ profile via feedback control has been recently demonstrated in EAST in both L-Mode and H-mode experiments. Extensive studies have shown that the $q$ profile, which is closely related to poloidal magnetic flux profile, is a key factor to achieving advanced-tokamak operating scenarios that are characterized by improved confinement and the non-inductive sustainment of the...
A novel integrated model GOTRESS+ has been developed, which consists of the iterative transport solver GOTRESS as a kernel of the integrated model$^{1,2}$, the equilibrium and current profile alignment code ACCOME and the neutral beam heating/current-drive code OFMC. GOTRESS is able to find out an exact steady-state solution using global optimization techniques, enabling us to robustly deal...
Experiments have been carried out to explore quiescent H-mode (QH-mode) $[1]$ scenario under the condition of low torque and pure RF injection (true zero torque) with ITER like tungsten divertor on EAST, which will facilitate the eventual use of this scenario on ITER and other reactors. EAST achieved the stationary QH-mode over 50-80 energy confinement time or 3-6 current relaxation time with...
Simultaneous control of large ELMs and divertor heat load in a metal wall environment is crucial for steady-state operation of a tokamak fusion reactor. A new scenario for ELM suppression compatible with radiative divertor has been demonstrated, for the first time, in the EAST superconducting tokamak. An n = 1 mode [FIG 1(f)] along with its harmonics, initiating from the oscillation of a...
For future fusion devices including ITER and DEMO, it is crucial to handle the power flux to the divertor targets in order not to exceed the steady-state material limit, q$_{t,⊥}$≤ 10 Wm$^{-2}$ [Ref. 1]. One of the viable effective methods to achieve it under development is divertor detachment via impurity gas seeding or deuterium injection in the divertor region in order to dissipate the...
The configuration of reversed shear q(r)-profile is existence for the burst of sawteeth-like events, and the collapse is triggered periodically by the magnetic reconnection of double tearing modes (DTM) [1], with abbreviation of DTRC (double tearing reconnection crash). The excitation conditions of DTRC has strong relationship with the impurity ions, the power threshold of ECRH, the influence...
Successfully confining, and heating D and DT plasmas in tokamaks results in the emission of neutrons, which carry a part of the excess energy produced in the fusion of fuel ions. Upon escaping the magnetic confinement neutrons effectively transfer fusion energy generated in the plasma to the device’s first wall and other tokamak components, such as plasma diagnostics. In future fusion devices...
Recent control advances at KSTAR enabled us to not only sustain the ITER-similar shape (ISS) in a stationary manner but also experimentally demonstrate the ISS-compatible RMP-ELM control in KSTAR for the first time, using the n=2, +90-deg phasing RMP, matching the ITER-like dimensionless parameters in lower single null (LSN) configuration with vastly contrasting upper/lower triangularities....
The Divertor Tokamak Test facility (DTT) [1-3] is a D-shaped superconducting tokamak (R=2.14 m, a=0.65 m, BT≤6 T, Ip≤ 5.5 MA, pulse length ≤ 100 s, auxiliary heating ≤ 45 MW, W first wall and divertor), whose construction is starting in Frascati, Italy. Its main mission is to study the controlled exhaust of energy and particle from a fusion reactor, which is a top priority research item in the...
Due to edge opacity, future tokamaks will rely on injection of cryogenic pellets for plasma fuelling. Maintaining the desired density and isotope composition crucially depends on the transient plasma response to pellet injection, motivating dedicated studies for increased understanding and predictive capabilities. In a recent experiment at the Joint European Torus (JET), Deuterium pellets...
Introduction and motivation.
Numerical simulations with the EIRENE [1] code are indispensable for both understanding and predicting the fuel and impurity transport in the edge and divertor areas of fusion devices including ITER. The transport determines impurity penetration towards the core, plasma exhaust and plasma-surface interaction (PSI) issues. The insight into the interplay of...
The toroidal rotation without any external momentum sources known as an intrinsic rotation has been focused an important topic since the most promising toroidal rotation source driven externally from a neutral beam injection may not successful for the future burning devices like ITER and DEMO. The toroidal rotation in pure ohmic plasmas is self-generated and it is considered as one of the most...
The helium plasmas have been demonstrated for the first time on EAST under the condition of pure RF-heating and ITER-like tungsten divertor, which advances physical understanding in support of the ITER non-nuclear operational phase $[1]$. Concentration of helium ($C_{He}$) in the plasma is confirmed to play a critical role in H-mode operation, as higher concentration raises the H-mode...
Hybrid scenarios are under development in KSTAR which are defined as “stationary discharges with β_N ≥ 2.4 and H_89 ≥ 2.0 at q_95 < 6.5 without or very mild sawtooth activities”. β_N≲3.0, H_89≲2.4 and G-factor (≡β_N H_89/q_95^2) ≲0.46 has been obtained simultaneously at ne/nGW~0.7 and sustained for ≳40 τ_E during the main heating phase as shown in figure 1.
![A representative hybrid...
Development of high performance operation regimes in the magnetic confinement fusion devices has been of great interest in the fusion community for decades as it is critical to accomplish efficient steady-state operations of fusion reactor. Since the first discovery of high confinement mode (H-mode) in tokamaks that claimed enhancement of energy confinement by more than a factor of 2 compared...
Plasma-wall interaction (PWI) processes are important in long-pulse plasma operation of fusion devices due to the main issues of increased fuel retention, material erosion and redeposition which are induced by a large increase in particle fluence to the wall compared to present experiments. The in situ approach is an urgent requirement to be utilized to real-time measure the fuel content on...
Leading edge induced material damages are very critical in future fusion devices which may have cassette structure for plasma facing components [1]. The upper tungsten divertor in EAST is the first application of active cooled ITER-like W/Cu monoblocks modules in a tokamak [2]. The misalignment between neighboring monoblocks was formed inevitably during fabrication and assembly processes,...
The high-performance operation is one of the major missions of HL-2M [1, 2] for supporting ITER, CFETR and future fusion reactors. Notice that, high-performance scenarios with the large plasma current (2.5-3MA) and the high elongation (1.8-2.0) are normally accompanied by the potential VDEs risk. This requires an efficient and reliable feedback control system, which is under construction. This...
The Integrated Modelling & Analysis Suite (IMAS) is the software infrastructure that is being developed building upon the modelling expertise from across the research facilities within the ITER Members to support the execution of the ITER Research Plan [1]. It is built around a standardised representation of data described by a Data Dictionary that is both machine independent and extensible. ...
Disruption prediction and avoidance is a high-priority challenge for tokamaks to sustain long pulse and high performance plasmas that are critical for ITER and next-step devices for fusion generation. Disruption-free, continuous operation of high performance plasmas over long pulse is a main goal of modern superconducting tokamak devices such as the Korea Superconducting Tokamak Advanced...
Control of impurity species in fusion plasmas is one of the main issues for long and stable plasma operation. The impurities can be generated by interaction of the plasma with its facing materials and/or by intentional injection for the purpose of maintaining radiative mantle. These impurities can cause detrimental radiation cooling and fuel dilution as they become accumulated inside the...
Repetitive buildup and collapse of the edge confinement barrier (called pedestal) in H-mode plasmas can seriously damage the plasma-facing components in tokamak fusion devices [1]. Hence, an accurate understanding of the underlying mechanism of the collapse is essential for the safe operation of fusion devices. We have proposed the formation of a solitary perturbation (SP) in the edge as a...
Kinetic plasma control based on extremely simple data-driven models and a two-time-scale approximation has been developed and validated on non-linear plasma simulations in recent years. Both in these models and in the associated control algorithms, the fast component (kinetic time scale) of the plasma dynamics is considered as a singular perturbation of a quasi-static magnetic and thermal...
Introduction
Compass-Upgrade tokamak, which is being constructed in the Institute of Plasma Physics of the Czech Academy of Sciences in Prague, will replace existing Compass tokamak $[1]$. It will be a compact, medium-size ($R=0,89 m, a=0,3 m$), high-magnetic-field ($5 T$) device. The tokamak is expected to operate with plasma densities up to $n_e=10^{21} m^{-3}$. Plasma heating will be...
Successful operation of ITER will require robust regulation of the plasma temperature and density despite the plasma's nonlinear dynamics and various uncertainties. In this work, a burn-control algorithm was designed to determine control efforts that will drive the plasma to desired targets. In order to effectively achieve the commanded control efforts, control allocation modules were...
$\quad$Plasmas with a high runaway electron (RE) current fraction, $f_{RE}$ > 0.5, have been achieved during the flat-top of EAST Ohmic discharges with both a circular limited and an X-point diverted configuration. Low toroidal mode number Alfvén eigenmodes (AE) in the frequency range of 100-300kHz including TAE, KTAE and GAE, which are excited by low-energy REs, are clearly identified in the...
Helium (He) operation has recently been successfully performed on EAST equipped with an upper ITER-like, water-cooled, tungsten (W) monoblock divertor. The main plasma-wall interaction issues in He plasmas have been studied and compared with those in deuterium (D) plasmas, such as divertor detachment, W erosion, material migration, ELM characteristics and control, etc. Studying the impact of...
Possible ways to suppress anomalous absorption at ECRH
E.Z. Gusakov, A.Yu. Popov
Ioffe Institute, St.-Petersburg, Russia
Possible approaches which allow reducing of anomalous absorption rate associated with the low-power-threshold two-UH-plasmon parametric decay instability, which is excited by the extraordinary pump wave in the X2 ECRH experiments in the vicinity of the plasma...
Preparation of the ITER experimental campaigns will require use of verified and validated models for the prediction of plasma response to actuators and of fusion performances. Several simulation codes have been developed by the fusion community for modelling of plasma equilibrium, transport processes, MHD stability, heating and current drive and fusion reactions. The integration of these codes...
The practical and economic viability of tokamak fusion reactors depends, in a significant way, on the efficiency of radio frequency (RF) waves to deliver energy and momentum to the plasma in the core of the reactor. The RF electromagnetic waves, excited by antenna structures placed near the wall of a tokamak, have to propagate through the turbulent edge plasma along their path to the core of...
KSTAR has a mission such as achieving a pulse length for more than 300 seconds and achieving a high-performance plasma[1]. The pulse length of the KSTAR discharge has increased each year gradually. Assigned research on the long pulse operation in KSTAR has been conducted since 2015. The pulse length of 90 seconds is achieved in the 2018 KSTAR experimental campaign.
The high $\beta_{P}$...
The electron cyclotron heating (ECH) is one of the intense methods of non-inductive plasma current drive (CD). The ECH waves accelerated the electrons with the Doppler-shifted electron cyclotron resonance (ECR) interactions, and effectively ramped and sustained the plasma current non-inductively to achieve long discharge duration. The plasma sustainment with ECH waves is a key issue for the...
Plasma waves naturally occur in various forms in magnetically confined plasmas. With a broad categorization into cold waves, hot (kinetic) waves, and coupled cold-hot waves, the plasma waves play integral functions in fusion plasma physics, such as particle confinement, fluid instabilities, and radiative processes. Some plasma waves, in particular low-frequency electrostatic (ES) waves, affect...
Chinese Fusion Engineering Testing Reactor (CFETR) aims to bridge the technical gaps between ITER and the first commercial fusion power plant [1]. The physics design of the operation scenarios should provide an integrated solution for the plasma to meet the key mission goals for CFETR subject to engineering constraints. For this purpose the optimization of scenarios and the exploration of...
Studies are carried out examining the dependence and sensitivity of fusion power production, temperature and density pedestals on edge density fueling strength, current density profile, alpha heating, and magnetic field strength. The goal of these integrated ITER simulations is to identify dependencies that can impact ITER fusion performance.
The self-consistent predictive core-pedestal...
The Advanced FRC is a Field Reversed Configuration maintained by neutral beam (NB) injection and electrode biasing (EB), with scrape-off-layer (SOL) pumping and electron heat confinement provided by expander divertors. This alternate magnetic confinement system has been developed at TAE Technologies, Inc in the C-2 [1,2], C-2U [3,4], and C-2W [5] series of devices. In this paper we summarize...
Tearing modes with poloidal/toroidal mode numbers $m$/$n$ = 2/1 have been routinely observed in KSTAR [1] with modes having notably large amplitude leading to significantly reduced both normalized beta, $\beta_N$, and plasma stored energy. Global kink/ballooning or resistive wall modes (RWMs) are observed to be stable at high $\beta_N$ above the $n$ = 1 ideal MHD no-wall stability limit,...
Recently, improved high-performance plasma operation has been significantly extended towards more ITER and CFETR related high beta steady-state regime with optimization of current profile (βP ~ 2.5 & βN ~ 1.9 with ITB +ETB of using RF & NB and βP ~ 1.9 & βN ~ 1.5 with eITB + ETB of using pure RF) on EAST [1]. The ITB formation and sustainment company with optimization of the current profiles,...
The KSTAR uses the NBI (neutral beam injection) as a majority of heating and current drive and has been exploring the inboard-limited ITB (Internal Transport Barrier) as an alternative candidate to achieve a high performance regime since 2016. The approach with the inboard limited configuration to avoid the H-mode transition prior to the formation of the ITB was effective at a given L-H...
Lower hybrid waves (LHW) are absorbed in the scrape-off layer (SOL), and then the heated plasma follows the magnetic field lines in the co-current and counter current directions, which intercepts the LHW antenna limiter and divertor plate [1,2]. Hot spots are observed on the guard limiter, and heat flux striations on the divertor plate are observed on both the ion and electron drift sides. In...
Of the three additional heating methods envisaged for ITER, waves in the Ion Cyclotron Range of Frequencies (ICRF) are attractive as the only one capable of ion heating and central deposition at high density. Yet, since their first use in magnetic fusion devices, the non-linear interaction of ICRF waves with the Scrape-Off Layer (SOL) plasma has attracted attention. This interaction is now...
Disruption prediction and avoidance is critical for ITER and reactor-scale tokamaks to maintain steady plasma operation and to avoid damage to device components. The present status and results from the disruption event characterization and forecasting (DECAF) research effort (1) are shown for multiple tokamak devices. Access to the full KSTAR, MAST, NSTX, AUG, TCV, and DIII-D databases is...
To predict energy and particle transport in future tokamaks we cannot use experimental measurements as boundary condition. Therefore, we need integrated modelling from the SOL to the plasma center.
On the other hand, transport in the various plasma regions is known to different degree. In order to increase our confidence on the transport predictions, we need to validate the available...
The development of tokamak start-up operation scenario often relies on operator’s experiences, rather than more robust approach based on numerical modelling. Such a trial-and-error approach has fortunately worked to find start-up recipes in small or medium size devices, but increases the risk of delays to experiments. Moreover, it would not be appropriate anymore for a large superconducting...
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....
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 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...
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 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 a recently conducted test for assessing compatibility of accelerator grid of Neutral beam [1] for their performance at 150 C, failure has been evidenced across an electrodeposited (ED) bond layer, which forms a vacuum boundary with cooling medium. This happens to be the first instance where an electrodeposited bond has been subjected to Hot Helium Leak test under operational conditions of...
The present contribution is devoted to the neutral beam injectors (NBIs) for ITER heating and current-drive. First, updated information is provided about the development status of the entire NBI prototype (MITICA); starting in 2021, the first experiments will be dedicated to high-voltage holding tests in vacuum. Then the contribution describes the full-scale prototype of the NBI ion source...
Introduction
Electron cyclotron systems of fusion installations are based on powerful millimetre wave sources – gyrotrons, which are capable to produce now megawatt microwave power in very long pulses. Gyrotrons for plasma fusion installations usually operate at frequencies 40-170 GHz. Requested output power of the tubes is about 1 MW and pulse duration is between seconds and thousands...
This paper presents a progress of the achievement of performance tests of ITER-gyrotrons developed in QST and design of dual-frequency (170 GHz and 104 GHz) gyrotron to enhance various operation scenarios in ITER such as characteristics studies of H-mode/ELM at low magnetic field. Major achievements of the ITER gyrotron developments are as follows: (i) Manufacturing of 6 out of 8 sets of ITER...
Shattered pellet injection (SPI) systems that form cryogenic pellets of low and high-Z impurities in a pipe-gun [1] for injection to mitigate disruptions have been fabricated and installed for use in thermal mitigation and runaway electron dissipation experiments on JET and KSTAR. These systems are to support disruption mitigation research for ITER and are based on an ORNL 3-barrel design for...
The mitigation of thermomechanical and runaway loads during disruptions and Vertical Displacement Events (VDEs) in ITER is essential for the project to execute the ITER Research Plan culminating (1) in the demonstration of the fusion power production goals (Q = 10 inductive operation for 300-500 s and Q = 5 for 1000 s and in steady-state up to 3000 s). To mitigate these loads ITER is equipped...
The Final Design Review (FDR) of the ITER Plasma Control System (PCS) for First Plasma will be held in July 2020 following the conceptual and preliminary designs [1,2] to prepare for First Plasma operation scheduled for the end of 2025. ITER operation follows the Staged Approach of the ITER Research Plan (IRP) [3]. The main goals of the First Plasma campaign include achieving a plasma current...
The Chinese Helium Coolant Ceramic Breeder (HCCB) Test Blanket Module (TBM) and its ancillary systems (together called Test Blanket System or TBS) is one of important steps for the China magnetic confinement fusion development, which will contribute to validate the key tritium breeding blanket technologies under the burning plasma environment, including tritium extraction, heat removal,...
The 3D nonlinear equilibrium and its associated magnetic topology are investigated on EAST for the first time for future understanding on the mechanism of how the Resonant Magnetic Perturbation (RMP) mitigates or suppresses the Edge-Localized Mode (ELM). Recently, a nonlinear transition from mitigation to suppression of the ELM by using RMP on EAST is observed $^{1}$. To understand the RMP...
For tokamaks to be attractive as the core of future fusion based power plants, it must operate in steady state or at least quasi-steady state without plasma current disruptions. As is evident from the predictions for ITER based on present day tokamak research, a major challenge would be to avoid disruptions in majority of plasma discharges with full plasma parameters achieving >99% good shots....
Shattered pellet injection (SPI) systems that form cryogenic pellets of low and high-Z impurities in a pipe-gun [1] for injection to mitigate disruptions have been fabricated and installed for use in thermal mitigation and runaway electron dissipation experiments on JET and KSTAR. These systems are to support disruption mitigation research for ITER and are based on an ORNL 3-barrel design for...
The Chinese Helium Coolant Ceramic Breeder (HCCB) Test Blanket Module (TBM) and its ancillary systems (together called Test Blanket System or TBS) is one of important steps for the China magnetic confinement fusion development, which will contribute to validate the key tritium breeding blanket technologies under the burning plasma environment, including tritium extraction, heat removal,...
Core transport in present tokamaks is mostly ascribed to micro-turbulence driven by the non-linear saturation of ion-scale ITG-TEM [1] instabilities ($k_\theta\rho_i\le1$, where $k_\theta$ is the poloidal wave number and $\rho_i$ the ion Larmor radius). It has been shown that electron-scale ETGs [2] ($k_\theta\rho_e\le1$) can also impact the heat transport, also exchanging energy with ITG-TEM...
Recent experiments in JET-ILW have been successfully exploring a high-performance H-mode scenario with no gas dosing at low $q_{95}$ ($I_p=3$ MA, $B_t$=2.8 T, $q_{95}=$ 3.2) and low triangularity, with peak neutron rates reaching values of 3.6$\times 10^{16}$ s$^{-1}$. This was enabled by operation at very low gas fueling, which is challenging in JET with the metal wall due the need to control...
State-of-the-art deep-learning disruption prediction models based on the Fusion Recurrent Neural Network (FRNN) (1) have been further improved. Here we report the new capability of the software to output not only the “disruption score,” as an indicator of the probability of an imminent disruption, but also a “sensitivity score” in real-time to indicate the underlying reasons for the imminent...
In recent years the multi-scale interaction between large-scale tearing modes and micro-scale turbulence has been found to be of paramount importance for thoroughly understanding the tearing mode physics and the island-induced transport, which will ultimately lead to developing a more effective method of the tearing mode control and optimizing the plasma performance in fusion devices, such as...
Future DT operation in ITER and DEMO will face a significant number of challenges. From the physics point of view, the change from DD to DT plasmas is poorly understood. There are indications that the core confinement, the ELM behavior, the pedestal confinement and the Scrape of Layer behavior can significantly change from DD to DT. From the operational point of view, scenarios with high...
The Final Design Review (FDR) of the ITER Plasma Control System (PCS) for First Plasma will be held in July 2020 following the conceptual and preliminary designs [1,2] to prepare for First Plasma operation scheduled for the end of 2025. ITER operation follows the Staged Approach of the ITER Research Plan (IRP) [3]. The main goals of the First Plasma campaign include achieving a plasma current...
$\qquad$ Characterizing and understanding the power threshold conditions for ITER to achieve H-modes ($P_{LH}$) is a major goal of a series of L-H transition experiments undertaken at JET since the installation of the ITER-like-wall (JET-ILW), with Beryllium wall tiles and Tungsten divertor [1,2,3,4]. In this contribution we report on results from L-H transitions studies in H, D and new almost...
Introduction
Electron cyclotron systems of fusion installations are based on powerful millimetre wave sources – gyrotrons, which are capable to produce now megawatt microwave power in very long pulses. Gyrotrons for plasma fusion installations usually operate at frequencies 40-170 GHz. Requested output power of the tubes is about 1 MW and pulse duration is between seconds and thousands...
This paper presents a progress of the achievement of performance tests of ITER-gyrotrons developed in QST and design of dual-frequency (170 GHz and 104 GHz) gyrotron to enhance various operation scenarios in ITER such as characteristics studies of H-mode/ELM at low magnetic field. Major achievements of the ITER gyrotron developments are as follows: (i) Manufacturing of 6 out of 8 sets of ITER...
The mitigation of thermomechanical and runaway loads during disruptions and Vertical Displacement Events (VDEs) in ITER is essential for the project to execute the ITER Research Plan culminating (1) in the demonstration of the fusion power production goals (Q = 10 inductive operation for 300-500 s and Q = 5 for 1000 s and in steady-state up to 3000 s). To mitigate these loads ITER is equipped...
In a recently conducted test for assessing compatibility of accelerator grid of Neutral beam [1] for their performance at 150 C, failure has been evidenced across an electrodeposited (ED) bond layer, which forms a vacuum boundary with cooling medium. This happens to be the first instance where an electrodeposited bond has been subjected to Hot Helium Leak test under operational conditions of...
Nonlinear two-fluid MHD simulations reveal the role of resonant field penetration in ELM suppression and density pump-out in low-collisionality ITER-Similar-Shape (ISS) plasmas in the DIII-D tokamak$^1$. The operational window for ELM suppression in DIII-D ISS plasmas coincides with calculations for magnetic island formation at the pedestal top ($n_e<3\times10^{19}m^{-3}, B_r/B_t >10^{-4},...
Since the initial JET operations with the metal wall (JET-ILW), the experimental results have shown a pedestal pressure in baseline plasmas that tends to be 10-20% lower than in the corresponding earlier carbon wall operations (JET-C) [1]. While this degradation seems mainly correlated with the high fueling rates typical of JET-ILW [2,3] and/or the lack of carbon impurity [4,5], an exhaustive...
Good confinement of the fusion-born alpha particles is essential to ensure adequate burning plasma performance in next-step fusion devices. Among the processes determining this confinement, instabilities triggered by energetic particles (EPs) may play a major role, and are currently being studied in various tokamaks using auxiliary power sources to sustain EP populations. Instabilities...
A series of experiments have been executed at JET to assess the efficacy of the newly installed Shattered Pellet Injection (SPI) system in mitigating the effects of disruptions. In this contribution, the results from these JET SPI experiments are presented and their implications for the ITER disruption mitigation scheme discussed.
An effective Disruption Mitigation System (DMS) that minimizes...
The present contribution is devoted to the neutral beam injectors (NBIs) for ITER heating and current-drive. First, updated information is provided about the development status of the entire NBI prototype (MITICA); starting in 2021, the first experiments will be dedicated to high-voltage holding tests in vacuum. Then the contribution describes the full-scale prototype of the NBI ion source...
Significant stabilizing effect of kinetic thermal ions is found for the LHD plasmas. The kinetic MHD simulations for the LHD plasmas at high magnetic Reynolds number show that the high beta plasmas can be maintained since the saturation level of the pressure driven MHD instabilities is significantly reduced by the kinetic thermal ions. This results from the fact that the response of the...
Disruption in the TOKAMAK device is generally known as one of the most harmful events. The subsequent event of the thermal quench and the current quench cause collateral heat-damage and structural damages. These two potential sources of danger are relatively well known because it is easy to conceive that the confined thermal energy and the magnetic field energy associated with the plasma...
The JT-60 Super Advanced (JT-60SA) tokamak construction have been achieved respecting the requirements of very tight tolerance for the assembly and by handling very heavy components in a very close space environment. The construction of this large superconducting tokamak represents a big step forward in the world nuclear fusion history, opening the road for ITER and DEMO. Precise assembly is...
In magnetically confined fusion devices, nonlinear wave-wave interaction has been noticed to play important roles in the production of new modes. On NSTX, nonlinear interactions among low-frequency energetic particle modes (EPMs) and high-frequency toroidal Alfvén modes (TAEs) have been reported $[1]$. On JET, a 3/2 neoclassical tearing mode (NTM) is stabilized through the nonlinear coupling...
In this paper, the JT-60SA cryogenic system and the results of the commissioning and annual operations are summarized. During the commissioning of the cryogenic system, performances for each component and the automatically controlled operation sequence have been confirmed. Notably, mitigation of cryogenic heat load fluctuations in large superconducting tokamak machines is essential. The...
In this paper, we present the recent experimental results of cross phase influence on turbulent momentum and particle transport in the edge of HL-2A tokamak. The mathematical expressions for cross phases are derived in Fourier domain. The fluctuations and turbulent flux are measured by Langmuir probes. For Reynolds stress, prominent phase scattering in the strong shear layer is found, which...
Core density profile peaking and particle transport have been recently extensively studied on several tokamaks [1,2,3]. In JET, the earlier research of the significance of the NBI fueling on density peaking [4] was recently confirmed when thanks to the development of the gas puff modulation technique it was found that in ITG dominated plasmas, NBI is responsible for typically half of the...
Assembly of 1 MV power supply (PS) components to produce 1 MeV negative ion beams have been completed for the ITER neutral beam test facility (NBTF). To realize 1 MV insulation after the assembly of long and complicated components, (i) dust and particle during the assembly were controlled, (ii) the transmission line (TL) with a total length of 100 m composed of 22 series connected vessels was...
The Lithium Tokamak eXperiment-$\beta$ (LTX-$\beta$), the upgrade to LTX, is designed to utilize low recycling walls and the resultant gradient-free temperature profiles [D. P. Boyle et al., Phys. Rev. Lett. 119, 015001 (2017)] to robustly stabilize ion and electron temperature gradient-driven modes. Low recycling and the resultant low collisionality in the scrape-off layer (SOL) plasma [R....
Introduction
In ITER the density of deuterium and tritium will be controlled by injection of cryogenic pellets, with D and T isotopes separately to allow active isotope ratio control. Due to technical limitations and high pedestal temperatures the pellets will be ablated at the plasma periphery, about at the pedestal top. For deeper penetration one have to invoke curvature drift and...
The SST-1 machine is the first medium size superconducting tokamak operational at Institute for Plasma Research India. It comprises of set of toroidal (TF) as well as poloidal field (PF) superconducting coils system. In order to cool and maintain these magnets under superconducting state, a dedicated specially designed helium cryogenic system of cold capacity 1350 W at 4.5 K along with its...
This report focuses on the development of the thermal insulation devices including thermal shield (TS) and cryostat for the superconducting tokamak JT-60SA.
- Design, manufacturing and acceptance test were successfully completed
by 2019 and installation will be done by March 2020.
- The technique and knowledge to realize high accuracy manufacturing
and short time installation of...
The 2018 ITER Research Plan states that “Operation of ITER will have to strongly focus on avoiding disruptions with a high success rate and on mitigating those in which avoidance techniques fail” (1). The development of a disruption mitigation system for ITER will not suffice. We discuss a nonlinear effect that can allow the RF current drive stabilization of larger islands than would otherwise...
The disruption mitigation technology remains the key issue of safe and reliable device operation in future large tokamaks including ITER [1,2]. Several approaches have been proposed and experimentally tested in contemporary devices, which demonstrate opportunities of massive gas, pellets, dust and liquid gets injection in preventing the avalanche as the most dangerous mechanism of the runaway...
The control of edge localized mode is crucial for protecting the plasma facing components in the magnetic fusion reactor. In the investigation of ELM control, it has been commonly observed that the pedestal turbulence enhances during edge localized mode (ELM) mitigation with supersonic molecular beam injection (SMBI) [1], lower hybrid current drive (LHCD) [2] and impurity seeding [3][4]. The...
Pedestal formed in the plasma edge of high confinement mode (H-mode) strongly affects edge localized modes (ELMs) burst. Thus, understanding the physics of the pedestal instabilities is key to reducing the uncertainties associated with the realization of burning plasma conditions and the appropriate control of ELMs. Fusion power depends strongly on the pedestal pressure. RF waves, such as...
Runaway electrons (REs) are a crucial issue for future large tokamaks, especially during disruptions, due to the local impact of RE beam and large thermal loads they can place on the plasma facing components [1]. Therefore, a very active field of research has been opened up in the past decades on RE dynamics during disruptions [2]. Utilizing the newly developed key systems in the HL-2A tokamak...
N. Zhang$^1$, Z. C. Yang$^1$, Y. Liu$^1$, Y. Q. Liu$^2$, T. F. Sun$^1$,X. Q. Ji$^1$, P. Piovesan$^3$, V. Igochine$^4$, D. L. Yu$^1$, S. Wang$^1$, G. Q. Dong$^1$, R. Ke$^1$ , J. M. Gao$^1$, W. Deng$^1$, N. Wu$^1$, Q. W. Yang$^1$, M. Xu$^1$ and X. R. Duan$^1$, The HL-2A Team, The ASDEX Upgrade Team and The EUROfusion MST1 Team
$^1$ Southwestern Institute of Physics, P. O. Box 432, Chengdu...
Pellet injection is used in tokamaks and stellarators for fueling, ELM pacing and disruption mitigation. Injection of shattered pellets is a critical part of the envisaged ITER disruption mitigation system. Rapid deposition of a large amount of material is expected to result in a quick cooling of the entire plasma. However, it has recently been demonstrated that a considerable transfer of...
In magnetically controlled fusion devices, improving plasma performance is crucial for enhancing the confinement efficiency. An improved regime, high confinement mode (H-mode), has been chosen as the standard operating scenario for the international thermonuclear experimental reactor (ITER). Recently, the energy confinement improvement by externally seeded low or medium Z impurities has been...
The mechanism of excitation of beta-induced Alfv\'{e}n eigenmodes (BAEs) with magnetic island larger than a threshold without energetic ions is studied. It is found that the nonlinear coupling between Geodesic acoustic mode and magnetic island can drive the pair of BAEs. To excite the BAEs, the phase of BAEs to island should be $\pi/2$ and the magnetic island is larger than a threshold. The...
The nonlinear interactions between energetic particle (EP) and Alfvén waves are very important for astrophysics and high temperature plasma physics, especially for magnetically confined fusion plasma, because they will affect the redistribution and transport of EPs significantly. When the EPs have sufficiently strong pressure gradient, they can excite a non-normal mode, named energetic...
Alpha particles are the key players of a burning plasma as they provide the self-heating required for the sustainment of the fusion burn. At the same time, however, there is only little experimental knowledge on their properties, mostly because of the limited availability of deuterium-tritium (DT) plasmas. Among the challenges that the scientific program of the Joint European Torus (JET) is...
Here we report recent progresses of laser fusion energy research in Japan, especially on the fast ignition scheme. For the fast-track to the laser fusion energy, we are investigating the fast-ignition plasma physics to realize optimal compression of a fusion fuel as well as efficient heating of the compressed fuel. In this scheme, we have demonstrated the efficient heating of high density...
Experimental investigations of frequency slowly-sweeping Alfvenic modes have been carried out on the HL-2A Tokamak. There are two different kinds of instabilities in the neutral beam heated plasma, i.e., the typical reversed shear Alfven eigenmodes (RSAEs) and the high modes with frequency slowly sweeping from 500kHz to 100kHz. On one hand, the RSAEs are driven unstable by the passing fast...
Gamma-ray spectrometry of the plasma [1] is one of the tools giving information on the heating efficiency. The source of gamma-ray is nuclear reactions between energetic confined ions and plasma impurities, i.e. Be and bulk plasma ions. Gamma-ray diagnostics allow monitoring the energy distributions of the fusion products, ions accelerated during ICRF heating and plasma fuel ions and provide...
To verify the CarMa0NL modelling for COMPASS-U, the numerical results are cross-validated with general analytical predictions [Pustovitov V. D., Nuclear Fusion 55 113032 (2015)]: the computed vertical force on the tokamak wall is found to be almost zero during fast transients, as it should be. This test proves the credibility of the model and computational method. The role of poloidal eddy...
Achieving high-$\beta_N$ for current and future tokamaks is a challenging and important issue, where \beta_N is the normalized toroidal beta. High-$\beta_N$ is beneficial for the ignition and fusion reaction, as well as the ratio of bootstrap current is proportional to $\beta_N$. Recently, on HL-2A a high-performance region, combining edge and internal transport barriers (double transport...
The report presents the results of the design development of the equatorial diagnostic port #11 of the final project level and the upper diagnostic ports ## 02,07,08 of the preliminary design level. Ports are being developed at the Budker Institute of Nuclear Physics of the Siberian Branch of the Russian Academy of Sciences (BINP SB RAS or shot BINP) in close cooperation with the ITER central...
Recent experiments at the JET tokamak with ITER-Like-Wall studied intrinsic rotation in a large tokamak, addressing questions related to the effects of collisionality and hydrogen isotope type on the amplitude of the measured toroidal rotation and rotation reversals of Ohmic plasmas. The isotope effect on the intrinsic rotation was investigated by comparing the rotation of the main ion in...
Abstract
The nuclear fusion reactor ITER (International Thermonuclear Experimental Reactor) foresees a Pressure Suppression System (PSS) in order to manage a Loss of Coolant Accident (LOCA) or other over pressurization accidents in the Vacuum Vessel (VV) which has a pressure limit fixed at 150 kPa (abs).
This system (VVPSS) has a key safety function because a large internal...
Abstract In the frame of the EUROfusion breeding blanket research activities, two reference blanket concepts are developed, the helium cooled pebble bed and the water cooled lead lithium (WCLL) blankets, which represent the most attractive designs for a DEMO reactor $\left[ 1\right] $. Test Blanket Modules (TBMs) derived from these concepts will be tested in ITER.
The design of breeding...
Abstract. The study is devoted to theoretical analysis of the models for calculating the disruption forces in tokamaks. It is motivated by the necessity of reliable predictions for ITER. The task includes the evaluation of the existing models, resolution of the conflicts between them, elimination of contradictions by proper improvements, elaboration of recommendations for dedicated...
The intensive experimental and theoretical study of the Edge Localized Modes (ELMs) and methods of their control is of great importance for ITER [1]. The application of small external Resonant Magnetic Perturbations (RMPs) has been demonstrated to be efficient in ELMs suppression/mitigation in present day tokamaks [2]. RMPs are foreseen as one of the methods of ELMs control in ITER [3]. ...
A transition from an interchange mode with high mode numbers $(m,n)$ for $m$ (poloidal) and $n$ (toroidal) to a non-resonant $(m,n)=(1,1)$ mode is found in the nonlinear magnetohydrodynamic (MHD) simulation for a Large Helical Device (LHD) plasma with net toroidal current. This transition occurs when the rotational transform is closed to unity in the core region. Because partial collapses...
A helical coil designed to passively generate non-axisymmetric fields during a plasma disruption has been shown (via electromagnetic circuit and linear MHD modeling) to be effective at deconfining runaway electrons (REs) before the RE beam current grows to dangerous levels. Magnetic equilibria from DIII-D RE experiments were used to calculate the toroidal electric field generated during the...
The extreme energy content in ITER makes the disruptions a matter of grave concern. The current strategy of disruption mitigation in ITER relies on the injection of cryogenic pellets into the disruptive plasma [1]. Pellet ablation is an essential factor in disruption mitigation, which calls for dedicated theoretical support in modeling this process in the plasma and establishing related...
Plasma rotation in thermonuclear fusion plasma plays an important role for particle and energy confinement and it can stabilize magnetohydrodynamic instabilities if some rotation level is achieved. In early studies, the relation between plasma confinement and radial electric field E_r was studied via measurements of the electric potential distribution and also via spectroscopic measurements of...
Plasma detachment is the desired operational regime for ITER baseline scenario and in next-step fusion reactors, as it allows to reduce the heat fluxes impacting onto the divertor plasma-facing components (PFCs) below their material limits. It is typically characterized by a reduction of plasma pressure between the upstream separatrix and the divertor targets, which is caused by dissipation of...
The commissioning of the power supplies for superconducting coils in JT-60SA has been done with dummy load (Inductance: 7.64 mH, Resistance: 6.995 m$\Omega$) since June 2019. The most important results are that (i) Integrated operation of the different PS components was completed successfully, (ii) High voltage generation of the rated voltage of 5 kV by Switching Network Unit (SNU) was...
The paper summarizes the studies carried out on a novel arrangement of the core X Ray Crystal Spectroscopy (XRCS) system for ITER, particularly with respect to physics analysis and system integration in a different equatorial port in ITER. The XRCS Core diagnostic is the only one available for ITER at PFPO-1 (Pre-Fusion Power Operation-1) phase, and offers key parameters like core ion...
In this presentation, we report recent advances in the development of the CFQS quasi-axisymmetric stellarator as a joint project of National Institute for Fusion Science, Japan and Southwest Jiaotong University, China. The quasi-axisymmetric stellarator (QAS) offers good plasma confinement properties with low aspect ratio, giving a prospect to become a compact fusion reactor. MHD equilibrium...
Ion Cyclotron Resonance Frequency (ICRF) heating plays an important role in many present day experiments and it is one of the auxiliary heating methods that will be used in ITER. In this contribution, we will review the recent key ICRF results from the JET and ASDEX Upgrade (AUG) tokamaks in preparation of ITER.
In the recent JET campaigns, the focus has been in the preparation of...
Shattered pellet injection (SPI) has been selected as the baseline disruption mitigation (DM) system for ITER. SPI utilizes cryogenic cooling to desublimate low pressure (<100 mbar) gases onto a cold zone within a pipe gun barrel, forming a cylindrical pellet. Pellets are dislodged from the barrel and accelerated using either a gas driven mechanical punch or high-pressure light-gas delivered...
This paper presents numerical modelling to assess the fast shutdown scenario by shattered pellet injection (SPI) in a baseline strategy of the ITER Disruption Mitigation System (DMS). A new versatile 1.5D disruption simulator INDEX is applied for this purpose and a SPI ablation/assimilation model has been implemented. Pre-thermal quench (pre-TQ) H$_2$/D$_2$ SPI is proposed as a promising...
![Impurity Density and Total Pressure Profiiles and 3D Visualization][1]
The efficacy of ITER's DMS must be evaluated within the next several years and, since in-situ evaluation is impossible, verified and validated simulations are critical. High fidelity 3D initial value simulations of Shattered Pellet Injection and Dispersive Shell Pellet (SPI and DSP) simulations show favorable...
Plasma Exhaust and Plasma Wall Interaction are subjects of intense studies in fusion energy research for the understanding of the amount of heat loads and the lifetime of Plasma Facing Components. In order to ensure reliable predictive edge modeling in this context, it is mandatory to determine the transport properties of the Scrape Off Layer (SOL), a region largely influenced by the presence...
MHD stability at edge pedestal in QH-mode plasmas in DIII-D and JT-60U was analyzed by considering plasma rotation and ion diamagnetic drift effects. It was found for the first time the coupled rotation and ion diamagnetic drift effects can stabilize a kink/peeling mode in both experiments in case the rotation direction is counter to the plasma current direction although it has been recognized...
ITER Council has decided to reduce the number of port from three to two for testing TBM due to the introduction of a Disruption Mitigation System during November 2018. Since then, India is proposing Helium Cooled Solid Breeder (HCSB) concept to be tested in one half of the two ports available for the same in ITER. Indian HCSB concept is having Reduced Activation Ferritic Martensitic Steel...
In Wendelstein 7-X, the vacuum rotational transform, $\bar\iota$, has a rather small shear and does not cross any major rational surfaces. Nevertheless, during plasma operation it can be modified by electron cyclotron current drive (ECCD) in such a way that the resulting iota profile passes through low-order rational values, potentially triggering magnetohydrodynamic (MHD) events.
Indeed,...
The termination of high performance plasmas in tokamak devices with high Z metal plasma facing components presents challenges related to the influx of heavy impurities which, if not kept under control, cause an increase of the radiative losses, radiative cooling and high probability of disruption.
A number of key players in these dynamics have been identified by intensive research performed...
![(top left) growt rate of the ideally unstable $2/3$ mode. (top right) radial displacement of this mode. (bottom) Stochastization of the core region due to the perturbed field associated with this mode.][1]
We present a new model of the sawtooth oscillation that can explain direct measurements which show $q_0$ well below 1 during the entire cycle, and observations that indicate the $q=1$...
Jing Wu1*, Jiming Chen1, Liman Bao2, Pinghuai Wang1, Kun Wang3, Stefan Gicquel2, Xiaobo Zhu1, Qian Li1, Hui Gao1, Weishan Kang1, Rene Raffray2
1Southwestern Institute of Physics, Chengdu 610041, China
2ITER Organization, 13115 St Paul Lez Durance, France
3China Int’l Nuclear Fusion Energy Program Execution Center, Beijing 100862, China
*E-mail: wj@swip.ac.cn, Tel:...
Jiming Chen1*, Pinghuai Wang1,Kun Wang2, Liman Bao3, Xiaobo Zhu1, Jing Wu1, Stefan Gicquel3, Qian Li1, Hui Gao1, Jialin Li1, Rene Raffray3, Ming Xu1, Xuru Duan1
1Southwestern Institute of Physics, Chengdu 610041, China
2China Int’l Nuclear Fusion Energy Program Execution Center, Beijing 100862, China
3ITER Organization, 13115 St Paul Lez Durance, France
*E-mail: chenjm@swip.ac.cn, Tel:...
It is an extreme challenge to reduce the transient peak of the heat load on the plasma facing components (PFC) in tokamak plasma$[$1$]$. One effective way is to increase the wetted area on the divertor target by splitting the strike point. Most of the experimental results show that the change of magnetic topology induced by RMP and LHCD is responsible for the strike point splitting$[$2-4$]$....
This contribution presents a numerical assessment of the impact of density fluctuations on the electron cyclotron (EC) wave in view of neoclassical tearing mode control in European DEMO. We show that, using the current design for the EC system launching the EC wave from equatorial outboard plane, the quality of the EC current profile is severely affected by the density fluctuations located at...
Parametric decay instability (PDI) is a kind of nonlinear wave-wave interaction, which significantly influence the wave accessibility and heating in plasmas. In fusion plasmas, the parametric process is typically displaying as quasi-mode decay, such as nonlinear landau damping or ion cyclotron harmonic decay. [1] For these quasi-mode decays, the previous kinetic theory [2] for PDI, where...
At the present time, the preparing for physical start-up of tokamak T-15MD is completed in the National Research Center “Kurchatov Institute”. Tokamak T-15MD has the following parameters: R=1.48 m, a=0.67 m, B=2.0 T, Ipl= 2.0 MA [1]. The electromagnetic system is capable of maintaining without overheating (more 60ºC) the plasma current of 1MA for 40s, 700 kA for 120 s, 500 kA for 250 s and 300...
We report recent advances in reducing the coil complexity for optimized stellarators. Three efforts have been dedicated. First, the FOCUS code which uses fully 3-D representations for coils and employs analytically calculated derivatives has been applied in designing coils for new stellarators. FOCUS allows searching for more design space and thus is able to find more possible solutions....
In JET plasma with a carbon wall (JET-C), and most other existing tokamaks, exceeding a critical density results in a H-L back transition, even when well above the H-mode power threshold. In contrast, at high density, JET plasma with a Be/W ITER-like wall (JET-ILW) always enter a ‘dithering’ phase before the H-L back transition, which enables a ($\approx20\%$) higher H-mode density limit (HDL)...
- Introduction : Disruptions are one of the major concerns in ITER and other future tokamaks [1]. A particularly troublesome type of disruption is a vertical displacement event (VDE) where control of the vertical position of the plasma column is lost. In addition to heat, particle flux, and energetic electrons impacting the first wall, significant electromagnetic loads will arise.. For...
A promising repeatable laser system producing multi-kilojoule of pulse energy has been basically designed for realization of the fast-ignition-based inertial fusion energy (IFE) reactor. Two cultivated core key technologies ensure high reliability of the proposed design. First, using our novel bonding technology, a cryogenic active-mirror amplifier has been developed to enable 100 Hz...
Inertial Confinement Fusion (ICF) schemes are designed to heat and compress DT fuel to conditions exceeding the Lawson criterion ($p \tau$) using implosion, which greatly amplifies the pressure of a driver (~100 MBar) to the conditions necessary for laboratory-scale ICF (~100s GBar). The National Ignition Facility (NIF) focuses on the laser indirect drive approach to ICF, in which laser energy...
TAE Technologies, Inc. (TAE) is a privately funded company pursuing an alternative approach to magnetically confined fusion, which relies on field-reversed configuration (FRC) plasmas composed of mostly energetic and well-confined particles by means of a state-of-the-art tunable energy neutral-beam (NB) injector system. TAE’s current experimental device, C-2W (also called “Norman”) shown in...
The capability to suppress edge localized modes (ELMs) is crucial for the success of ITER because the transient heat loads on the divertor due to ELMs would reduce the lifetime of plasma facing components to unacceptable levels. ELMs can be suppressed with the application of resonant magnetic perturbations (RMPs). But a side effect of RMPs is enhanced particle flux, or density pump-out, that...
Introduction. Extrapolations to ITER and DEMO from existing smaller experiments alone are unreliable, especially for turbulent transport - requiring the aid of predictive simulations. The 3D fluid turbulence code GRILLIX [1–4] is used to study confinement improvement through turbulence suppression that is compatible with power exhaust. This contribution describes the validation against...
Outline. We report on major progress regarding simulations of edge localized modes (ELMs). First of a kind simulations of realistic repetitive type-I ELM cycles are presented, reproducing in particular the explosive onset of the ELM crashes for the first time. Key to this achievement were numerical improvements, fully realistic plasma parameters and flows, a self-consistent evolution of...
In this paper we will present nonlinear full-$f$ electromagnetic gyrokinetic simulations of turbulence in the pedestal and scrape-off layer (SOL) region of a tokamak. The algorithms in the Gkeyll code solve the electromagnetic gyrokinetic equations using a continuum high-order discontinuous Galerkin scheme. The equations are written in a sympletic form in which the particle parallel momentum...
During burning plasma operation on ITER, extrinsic impurity seeding will be mandatory for heat flux control at the tungsten (W) divertor vertical targets [1]. A very extensive database of SOLPS plasma boundary code simulations has been compiled for ITER [1], including the most recent advances, obtained with the SOLPS-ITER version, in which for the first time, fluid drifts have been included...
Introduction – Achieving safe power and particle exhaust compatible with a high-performance core plasma is one of the main challenges towards commercial fusion power. Currently, the most promising solution is to operate a diverted tokamak in detached conditions with high divertor neutral pressure, high volumetric power dissipation, and a strong temperature and pressure gradient along the...
Power exhaust is one of the big challenges for future fusion reactors. In the EU programme, both conventional and alternative divertor approaches are studied. For a conventional divertor in EU-DEMO (1), more than 80% of the exhaust power needs to be dissipated before entering the SOL to keep the divertor in the detached regime, where the interaction of the plasma with the wall is significantly...
A promising repeatable laser system producing multi-kilojoule of pulse energy has been basically designed for realization of the fast-ignition-based inertial fusion energy (IFE) reactor. Two cultivated core key technologies ensure high reliability of the proposed design. First, using our novel bonding technology, a cryogenic active-mirror amplifier has been developed to enable 100 Hz...
Successful operation of ITER depends critically on disruption management for the Pre-Fusion Power Operation (PFPO) phase up through Fusion Power Operations (DT). The power-handling capabilities of the beryllium (Be) first-wall panels (FWP) and other plasma-facing components (PFC) must be preserved in the face of disruptions and vertical displacement events (VDE). This need should account for...
Positive plasma potential was observed for the first time in a core tokamak plasmas, conventionally characterized by negative potential. Direct measurement of the electric potential in the core plasma is of paramount importance for the understanding of the role of radial electric field E_r in the mechanisms regulating the toroidal plasma confinement. New experimental observations and...
High performance advanced tokamak scenarios are very attractive for future burning plasmas. They can be achieved by elevating the central $q$-profile to values around unity to stabilize the sawtooth instability, which would otherwise reduce performance and could trigger deleterious instabilities. High-$\beta$ plasmas can develop such a flat elevated central $q$-profile in the presence of MHD...
Here we report recent progresses of laser fusion energy research in Japan, especially on the fast ignition scheme. For the fast-track to the laser fusion energy, we are investigating the fast-ignition plasma physics to realize optimal compression of a fusion fuel as well as efficient heating of the compressed fuel. In this scheme, we have demonstrated the efficient heating of high density...
Inertial Confinement Fusion (ICF) schemes are designed to heat and compress DT fuel to conditions exceeding the Lawson criterion ($p \tau$) using implosion, which greatly amplifies the pressure of a driver (~100 MBar) to the conditions necessary for laboratory-scale ICF (~100s GBar). The National Ignition Facility (NIF) focuses on the laser indirect drive approach to ICF, in which laser energy...
During burning plasma operation on ITER, extrinsic impurity seeding will be mandatory for heat flux control at the tungsten (W) divertor vertical targets [1]. A very extensive database of SOLPS plasma boundary code simulations has been compiled for ITER [1], including the most recent advances, obtained with the SOLPS-ITER version, in which for the first time, fluid drifts have been included...
TAE Technologies, Inc. (TAE) is a privately funded company pursuing an alternative approach to magnetically confined fusion, which relies on field-reversed configuration (FRC) plasmas composed of mostly energetic and well-confined particles by means of a state-of-the-art tunable energy neutral-beam (NB) injector system. TAE’s current experimental device, C-2W (also called “Norman”) shown in...
Outline. We report on major progress regarding simulations of edge localized modes (ELMs). First of a kind simulations of realistic repetitive type-I ELM cycles are presented, reproducing in particular the explosive onset of the ELM crashes for the first time. Key to this achievement were numerical improvements, fully realistic plasma parameters and flows, a self-consistent evolution of...
Introduction. Extrapolations to ITER and DEMO from existing smaller experiments alone are unreliable, especially for turbulent transport - requiring the aid of predictive simulations. The 3D fluid turbulence code GRILLIX [1–4] is used to study confinement improvement through turbulence suppression that is compatible with power exhaust. This contribution describes the validation against...
Introduction – Negative triangularity discharges were first studied on TCV to examine the effect of plasma shaping on energy confinement in ohmic, L-mode discharges [1]. Subsequent experiments, using ECRH to stabilize MHD instabilities, showed an improvement of energy confinement in negative triangularity as compared to similar positive triangularity discharges [2]. Modulated ECRH to allow...
We propose a novel heating mechanism for ions in overdense plasmas by introducing two whistler waves along a strong magnetic field in the counter-beam configuration [A]. The essential process is the collapse of standing whistler waves within a short timescale comparable to the wave oscillation period. During the collapse, ions are accelerated by a static electric field and acquire a large...
The uncertainties surrounding the physics of plasma exhaust and its centrality in reactor design require a thorough evaluation of promising alternatives as a precautionary measure to avoid delays in DEMO, if the ITER solution for the divertor could not extrapolate to reactor relevant machines. In this contribution, we review the physics and engineering work carried out within EUROfusion’s work...
The Japan Establishment for a Power-laser Community Harvest (J-EPoCH) is proposed as a next generation laser facility having multi-purpose high repetition laser beams at the maximum rate of 100 Hz. The omnidirectional 12 laser beams with 8 kJ would yield ~$10^{13}$ neutrons with a Large High Aspect Ratio Target (LHART) (1). As one of the applications of J-EPoCH, a laser fusion subcritical...
Divertor detachment is a scenario characterized by the dominance of neutral interactions to mitigate the extreme plasma heat flux that would otherwise be incident upon solid walls of fusion reactors. Despite the critical role theory will play in predicting divertor performance, rigorous modelling of neutrals is plagued by the difficulty of directly solving the nonlinear Boltzmann equation....
A world-class ultraintense laser LFEX at ILE, Osaka University directly heated a CD shell target, imploded by GEKKO XII(GXII) laser. Illuminating LFEX energy of 246 J increased the core internal energy by $23\pm 3$ J, leading to the conclusion that the heating efficiency is $9\pm 0.8$ %. The results encourage the fast ignition scheme fusion as a hopeful candidate of the fusion...
Introduction
An interesting phenomenon known as density incrustation at the interface of high Z / low Z plasmas was recently reported in high energy density (HED) systems like inertial confinement fusion (ICF) [1]. Radiation transport and hydrodynamic motion of materials are intricately coupled in HED systems and their interplay gives rise to several interesting phenomena. Density...
The divertor target is the most intense plasma-surface interaction area in tokamaks. Currently, the control of power load on the targets becomes to one of the most important issue for high performance long-pulse discharges. EAST has achieved over 100 s high performance operation, however, its lower graphite divertor prevents its achievement of further high-power long-pulse discharges $[1]$....
We developed a hydrogen population code by combining the Neutral-Transport code with the rovibrationally resolved Collisional-Radiative model (NT-CR), EMC3-EIRENE code, and the Molecular Dynamics (MD) simulation of carbon (C) and tungsten (W) divertor plates. Using this code, (i) we successfully treated hydrogen molecule H$_2$ reactions relating to the molecular assisted recombination (MAR),...
Here we report the mechanism of plasma heating by magnetized fast-isochoric (MFI) heating scheme (1). The mechanism was visualized experimentally by combining spectroscopic and spatially- and temporally-resolved X-ray imaging techniques. The MFI scheme employs an external magnetic field for guiding a high-intensity relativistic electron beam (REB) generated by relativistic laser-plasma...
Here we report theoretical and numerical studies for efficient plasma heating by high intensity lasers with a lateral confinement of the laser-accelerated fast electrons in the laser spot area. Recent experiments using kilojoule (kJ) petawatt lasers show efficient particle accelerations and plasma heating, indicating that the lateral loss of fast electrons is considerably small. We here found...
Laser Magneto-Inertial fusion is a recently developed approach for the thermonuclear fusion. It consists in applying to the laser inertial fusion plasma a strong magnetic field whose the role is to limit the diffusion of the formed plasma during the impact of an intense laser pulse with a target containing the thermonuclear fuel, as well as the confinement of produced alpha particles by the...
Erosion of plasma-facing components (PFC) due to sputtering by impinging ions and neutrals is one of the key challenges on the road to fusion power. Erosion will be a contributor to the overall PFC lifetime estimation, and to impurity production that can potentially lead to radiative collapse. Moreover, PFC erosion is directly linked to the key issues of fuel retention by co-deposition and...
Anomalous plasma transport in the boundary region of a tokamak plasma is normally associated with density structures. These density structures are commonly termed as plasma blobs. Recently, a theory for a universal mechanism of plasma blob formation has been put forward that is based on a breaking process of a radially elongated streamer [1] in the presence of poloidal and radial velocity...
In magnetic confinement fusion devices, the plasma density is largely self-sustaining through internal recycling processes. While refueling the plasma, the recycling neutrals can also considerably affect the energy transport in both the plasma confinement and exhaust regions. A quantitative understanding of the neutral-related effects first requires a precise knowledge of the origin and source...
Tungsten (W) and tungsten based alloys are candidate materials for plasma-facing components (PFCs) in future fusion reactors, largely due to their low erosion yield by physical sputtering and low retention of plasma fuel in them. The present work concentrates on erosion results obtained from two European tokamaks, ASDEX Upgrade (AUG) and WEST, which have operated with full-W first walls since...
Introduction – The H-mode confinement regime will be the main operational scenario on ITER and also the current foreseen scenario for fusion reactors. A continuous effort towards better predictive capabilities of H-mode confinement is being pursued on both experimental and theoretical fronts. The H-mode is characterised by the formation of a pedestal near the plasma edge and as the fusion...
Introduction -- Power exhaust solutions for a next step device must be compatible with good plasma performance. To reach sufficiently low divertor power loading impurity seeding is necessary. The amount of injected impurities required depends critically on both the maximum achievable separatrix density and the scrape-off layer width. Transient power loads due to type-I edge localised modes...
The high coupling efficiency ($\eta_c$) can be achieved by using a shell target with holes in the direct fast ignition. Here, we have made it clear that a diffusive heating is not negligible compared with a collisional and a resistive heating. In order to obtain high $\eta_c$, it is important to maintain the low effective hot electron temperature ($T_{eff}$). In the shell target with holes,...
Numerical simulations by the integrated divertor code SONIC show that the screening effect on the seeded high-Z impurity in the SOL plasma is improved through the enhancement of plasma flow induced by additional low-Z impurity injection. A single impurity injection of Ar into a steady-state high-beta plasma of JT-60SA results in a high Ar density at the top of SOL plasma, leading to an...
For a laser fusion reactor of fast ignition$^{1,2}$, we design an optimum implosion with solid spherical target and inserted conical gold target, where we achieve the maximum areal density $\rho R_{\rm max}=0.46$ g/cm$^2$. According to the hydro-equivalent, the results correspond to the re-quired laser energies for the implosions are 82 kJ for the ignition-scale-target ($\rho R_{\rm max}$=1.1...
Electric fields in plasma plays a key role in understanding many plasma phenomena from confinement to particle flows. In fusion machines like tokamak, changes in the edge radial electric field are also correlated with changes in many edge phenomena such as L-H transition, rotation, transport, and the suppression of large magnetohydrodynamic (MHD) instabilities called ELMs (Edge Localized...
Present and future long-pulse tokamaks such as JT60-SA, ITER and DEMO will require increasingly advanced control methods to maximize the plasma performance and pulse duration while avoiding plasma disruptions. Given the cost and complexity of a single discharge, maximal use of automated approaches is strongly preferred above costly and error-prone trial-and-error discharge development...
Tokamak plasmas are mostly optically thin for visible radiation-emitting out of the plasma as the density of the emitter in the plasma is not sufficiently high to produced self-absorption of spectral lines in the visible region. However, in certain conditions, such as during pellet ablation inside plasma and massive gas injection signatures of absorption in the emission spectra might be...
In the ADITYA-U tokamak [1], impurity seeding experiments were carried out to achieve transitions to radiative improved (RI) modes [2], which is usually characterized by the increased confinement along with increased plasma density, temperature and toroidal velocity shear profile as compared to the Ohmic mode. In this type of discharges, radiation from the edge region of the plasma is...
EX
The new results of the experimental study of the ion cyclotron emission (ICE) characteristic features in the NBI heated plasma of the TUMAN-3M tokamak [1] are presented. Figure 1 shows an example of the NBI ICE spectrum in deuterium plasma with deuterium NBI, comprising fundamental cyclotron resonance (IC) frequency and its harmonics. For the first time, a dispersion relation for the...
The paper reports unique observation of an ion cyclotron emission (ICE) from the ohmically heated plasma in the absence of energetic ions. The possibility to use the ohmic ICE for determination of the hydrogen isotopes ratio is discussed.
Described experiments were performed in the compact circular shaped limiter tokamak TUMAN-3M $(R(0)=0.53\:m$, $a=0.22\:m$, $B_T=1\:T$, $I_p=150\:kA$,...
In contrast to theory expectations, in numerous experiments the isotope effect results in the improvement of tokamak energy confinement as the hydrogen isotope mass increases [1]. This effect is beneficial and important for the success of Iter, where a mixture of heavy hydrogen isotopes will be used as a fuel.
The influence of the plasma isotope content on turbulence parameters and,...
The theoretical model of the feedback instability is proposed to explain the mechanism of the correlation between the detachment and the cross-field plasma transport. It is shown that (1) the feedback instability on the detached divertor plasma can be induced in a certain condition in which the recombination frequency $\nu_{\textrm{rec}}$ is larger than the ion cyclotron frequency...
ADITYA tokamak has been upgraded to ADITYA-U tokamak which is equipped to have shaped plasma operations. The main structural modification is the replacement of an old rectangular cross-section vacuum vessel by a new circular cross-section vacuum vessel. New poloidal coils (six) have been installed in between new vessel and toroidal field coils [1]. Additionally, as per the design requirement,...
In the last few years it was demonstrated experimentally on tokamak ASDEX Upgrade that with a big amount of seeded radiant (nitrogen) the outer target fully detaches, and the high radiation zone appears in the confined region up to 10 cm above the X-point. The location of such a spot may be controlled in real time by variation of the impurity seeding rate [1]. Such regimes might be promising...
Magnetic equilibrium modeling using the FIESTA code shows that steady-state snowflake (SF) divertor (1) configurations can be created and maintained with the existing poloidal field coil set in the MAST-U tokamak. A full multi-fluid plasma transport model with a computational grid encompassing two poloidal magnetic field nulls, with charge-state-resolved carbon impurities sputtered at material...
After the IAEA-FEC 2016 presentation (1) showing that the midplane-mapped heat-load width $\lambda_q^{XGC}$ predicted by the XGC1 edge-gyrokinetic-code (2) for the full-current ITER is $\stackrel{>}{\sim}6\times$ wider than the experimental-data based formula $\lambda_q^{Eich}$ (PRL 2011) while the same code reproduces $\lambda_q^{XGC}\simeq\lambda_q^{Eich}$ for the present experiments, a...
Tokamak is a toroidal device where plasma is confined by means of appropriate magnetic field configurations. Tokamak plasma is home to several magnetic Instabilities, which can lead to loss of confinement and even termination of plasma delivering very high heat load on the plasma facing components [1]. As these magnetic instabilities are dependent on the magnetic field configurations inside...
Implementation of suitable disruption mitigation technique remains the topmost priority for larger tokamaks including the ITER. The spontaneous disruption in ITER may probably be an unavoidable part, while operated with high performance D-T fuel [1]. Disruptions in ITER could produce very large heat loads on divertor targets and other Plasma Facing Components (PFC), and large electromagnetic...
Control and/or mitigation of runaway electrons (REs) is necessary for the operation of larger fusion devices including ITER [1]. The disruption generated REs, in particular, pose a serious threat for the plasma facing components in ITER as they are predicted to acquire very energy (~ tens of MeV). Many RE mitigation mechanisms like Massive Gas Injection (MGI) [2] and Resonant Magnetic...
The high-power reconnection heating of merging spherical tokamak (ST) plasmas has been developed by TS-3U, TS-4U$[1]$, UTST, MAST$[2]$ and ST-40 experiments and PIC simulations$[3]$, leading us to direct access to burning and high-beta ST often with absolute minimum-B without using any additional heating like neutral beam injection (NBI). All of them confirmed, (i) the promising scaling of ion...
Disruption instability and formation of the accelerated electron beams represent one of the main problems in design of the cost-effective fusion reactor. To minimize consequences of the disruptions in tokamaks, several methods for predicting disruption, controlling plasma discharge at the initial stage of the instability, and fast quenching (stopping) a plasma discharge are considered...
New concept of the longitudinal losses suppression for linear magnetic traps have experimentally demonstrated the reduction of the plasma flow by the factor of 2–2.5. This factor is in a good agreement with the theory. Preliminary scalings show the possibility of the further improvement of the suppression efficiency.
High relative pressure (β ≈ 60%), mean energy of hot ions of 12 keV and...
This paper presents solutions for critical problems in Japan’s DEMO (JA DEMO), which include common DEMO design issues beyond ITER-relevant technologies. The highlights of this design study are (i) system design for electric power generation, (ii) remote maintenance concept to attain a high plant availability, and (iii) novel concept for water-cooled pebble bed blanket and tritium recovery....
Fundamental mechanisms governing the prompt redeposition of tungsten impurities sputtered in tokamak divertors have been identified and analyzed to enable quantitative estimations and in-situ monitoring of the net erosion and lifetime of tungsten divertor plasma-facing components in ITER [1]. The net erosion of tungsten divertor PFCs is primarily determined by the prompt redeposition of...
A mini-reactor called CANDY that based on kJ-class diode-pumped solid-state laser (DPSSL) is proposed to perform feasibility studies of the power plant in fast ignition scheme fusion. In order to implement CANDY, we have adressed two issues. First is a repetitive operation of pellet injection and laser illumination, second is target physics related to inertial confinement fusion plasma. First...
Pulsed power technology of switched-mode has been employed to design and build a high-
voltage pulsed-power supply as an efficient pulse drive unit for low energy plasma focus
devices (PFDs). The plasma focus has been widely investigated as a radiation source,
including as ion-beams, electron-beams and as a source of x-ray and neutron production,
providing considerable scope for use in...
Reaching good-quality H-mode and the development of ELM control techniques are among the main priorities for ITER during its non-active operations [1']. Recent encouraging experimental results at JET-ILW demonstrated a significant reduction in the H-mode power threshold for NBI-heated plasmas when a small amount of $^{4}{\rm He}$ ions, $n(^{4}{\rm He})/n_{e} \approx 10\%$, was added to...
Scoping the possible operational regimes of a DEMO reactor requires reliable models of power exhaust processes. The scrape-off layer (SOL) and divertor in DEMO will need a higher radiated power fraction than present-day devices, and strong asymmetries in the SOL power fluxes entering the divertor regions may reduce the operational window in which the targets can be protected from excessive...
Sustained fusion reactions have been measured in a quiescent deuterium Z-pinch plasma, wherein unity beta was achieved, and sheared flows alone provided stability. Measurements from multiple scintillator neutron detectors demonstrated that 2.45 MeV neutrons were emitted uniformly along the majority of the 50-cm plasma column and that the neutrons were produced from a thermonuclear process with...
We have recently developed a kinetic neutral code for understanding the hydrogen neutral gas dynamics in the divertor region. The Linear Divertor Analysis with fluid model (LINDA) code [A] has been successfully coupled with Kinetic Monte-Carlo Neutral Code (KMNC) in the present study. This is the first attempt for the GAMMA 10/PDX detached plasmas to perform such a self-consistent and detailed...
It is critical to solve the power exhaust issue for future fusion devices, such as China Fusion Engineering Test Reactor (CFETR) [1]. In case of the steady-state operation, the divertor plasma is expected to be in the detachment regime. Thus, the heat load onto the targets can be reduced within the engineering limits and the sputtering can also be mitigated. Radiative snowflake divertor [2] is...
Power exhaust is considered as one of the most critical issues for future fusion devices. It is considered as an indispensable way for reducing divertor heat load to a tolerable level by seeding impurity and achieving a radiative divertor, especially for the future tokamaks with full metal walls. For China Fusion Engineering Test Reactor (CFETR) [1], which aims to achieve a fusion power ~1 GW,...
Hailong Du1*, Guoyao Zheng1, Xuru Duan1, Houyang Guo2, Jiaxian Li1, Lei Xue1 and Yue Zhou1
1Southwestern Institute of Physics, Chengdu, 610041, China
2General Atomics, PO Box 85608, San Diego, CA 92186-5608, United States of America
Email: duhl@swip.ac.cn
A new mid-size device, HL-2M $[$ 1$]$, is being installed at SWIP to address critical physics and technology issues towards advanced...
During non-local (“global”) L-H transitions found in various regimes of JET and JT-60U tokamaks earlier [1-3], the rise of Te,i and ne starts simultaneously in the spatial zone ≈0.3< r/a <1, while heat and density fluxes fall simultaneously in the same region. At the ITB-events in JT-60U and T-10 (circular tokamak with a limiter =30 cm, R=150 см, Вт= 2.5 Т), heat and density fluxes fall in a...
Finding an optimized Inertial Confinement Fusion1–3 experimental design is a challenge due to the large number of physical parameters that can be modified from experiment to experiment, and the inability of simulations to accurately
and rapidly a priori predict experimental results when these changes are made. Recently, a novel method[4] has been developed to address this issue by...