Since 18 of December 2019 conferences.iaea.org uses Nucleus credentials. Visit our help pages for information on how to Register and Sign-in using Nucleus.

# 26th IAEA Fusion Energy Conference - IAEA CN-234

Japan
Kyoto International Conference Center

#### Kyoto International Conference Center

Takaragaike, Sakyo-ku, Kyoto 606-0001 Japan
Description

The 26th IAEA Fusion Energy Conference is being organized by the IAEA in cooperation with the National Institute for Fusion Science, Japan. Previous conferences in this series were held in Salzburg (1961), Culham (1965), Novosibirsk (1968), Madison (1971), Tokyo (1974), Berchtesgaden (1976), Innsbruck (1978), Brussels (1980), Baltimore (1982), London (1984), Kyoto (1986), Nice (1988), Washington DC (1990), Würzburg (1992), Seville (1994), Montreal (1996), Yokohama (1998), Sorrento (2000), Lyon (2002), Vilamoura (2004), Chengdu (2006), Geneva (2008), Daejeon (2010), San Diego (2012) and Saint Petersburg (2014).
Support
• Monday, 17 October
• 08:30 10:15
Opening: O/1
Convener: Prof. Hiroshi Yamada (National Institute for Fusion Science)
• 08:30
Opening Address 10m
Speaker: Yukiya Amano
• 08:40
Welcome Address 15m
Speaker: Mr Host Country Representative
• 08:55
Opening remarks 5m
Speaker: Ms Meera Venkatesh (IAEA)
• 09:00
Fusion for Sustainable World Development 30m
It has been more than half a century since fusion energy research was disclosed at the 2nd Atoms for Peace conference, held in September, 1958 in Geneva. During the course of this period, DT-burning experiments were actually conducted in TFTR and in JET, both intended for energy breakeven: Q=1. This is a tremendous achievement of mankind, wishing to create a self-burning star on the Earth. The IAEA fusion energy conference was once held in Kyoto in 1986, so that this is the second time hosted in Kyoto. In the meantime, another IAEA-FEC was held in October, 1998 in Yokohama, which happened to be right after the DT-burning experiments, mentioned above. Interestingly, it was around that time the ITER-EDA came to a critical phase. As opposed to the rest of the world, from the beginning the Japanese fusion research community chose to explore multiple possibilities, including magnetic confinement by tokamak, helical and mirror configurations, and also laser-driven inertial confinement, each having made remarkable progress. In addition to achieving burning plasmas in tokamaks, LHD built at NIFS has been a unique effort in helical plasma studies, until recently W7-X in Germany has been put in operation. It is remarkable to find that these confinement facilities are making progress in their respective missions towards, the integration of which will hopefully lead to the realization of fusion energy. On its way, however, an experimental reactor, ITER being constructed in France, must be successful in sustaining the energy break even condition with Q>10, which will no doubt affect the design of the first DEMO reactor. For the rest of the process before fusion energy can be realized, all the governments in the fusion research community will hopefully provide continuous support for these confinement experiments, but basic research conducted in laboratory-scale facilities as well, which could end up with unexpected “spin-off” products, valuable for other communities. For example, the technology developed for superconducting magnets can be used for the long-distance DC-power transmission of solar energy. As such, one must remember that although the public acceptance of it may vary due to the socio-technical situation, fusion energy research and development can always contribute in many ways to the sustainable global development.
Speaker: Mr Atsuo Iiyoshi (Chubu University, 1200 Matsumoto, Kasugai, Aichi, 487-8501 Japan)
• 09:30
The Strategic Dimensions of the Fusion Energy Challenge 30m
Human beings have a short history in universe terms. Just 200.000 years old and no more than 100.000 years old out of Africa. As a consequence of climate changes those men decided to leave the continent looking for new land and new opportunities to prosper. It is impossible to understand the history of the humankind without keeping in mind its effort to understand the reality and to overcome the challenge to transform it. No other species has been able to advance so much in the knowledge of the environment; but those advances resulted, at the end, in radical changes in the environment, in our cultures and in our identity. Our civilization is built on the idea of permanent economic progress. We need to generate wealth as a premise to guarantee welfare, education, health, research, social cohesion… The economic activity requires energy. Our history shows a constant effort to improve our ability to generate more and more energy at minor possible cost. Economic revolutions are at the origin of social and political revolutions, but in many cases they are themselves a consequence of previous energy revolutions. If the capacity to generate energy is the precondition of economic progress and social welfare, we have to conclude that energy independence is a key goal for all the states. In advanced societies we have to face with growing demands of secure and clean energy sources at the same time. As a consequence nuclear fission, one of the cheapest an easiest ways to generate energy in large quantities, has been called into question. Without fission energy a large number of countries would have to largely depend on others to access to the necessary energy resources. Every day we see how energy dependence implies political dependence. A sovereign but energy dependent country has to condition its policy to the supplier’s interests. A free access to an unlimited and cheaper energy is the basis of political independence and equal opportunities in the global market. Research and production costs are conditioned by the price of energy. Political leaders try to confront these challenges through a diversification strategy. They are eventually succeeding in reducing the strategic limitations, but not the high costs of the energy generation and its effects. In some cases, like fracking (hydraulic fracturing), environmental impact remains a serious problem. In others, alternative ways of generating energy like Solar or Eolic power plants involve in many occasions a serious impact on the landscape and the environment. If we want a clean and cheap energy that guarantees not only respect for nature but also independence of states and enterprises, we need to direct our attention to fusion power. Nowadays fusion power is a major scientific challenge that requires the effort of everyone. A join action is required to overcome the challenges posed by its generation and this implies the involvement of states, the support of society and the determination of the scientific community. Success is just a matter of time and resources, and the consequences promise to be extraordinary. As a carbon-free energy source based on abundant fuels with no particular geographic distribution, controlled thermonuclear Fusion would eventually represent a breakthrough in the history of the mankind. Deuterium can be extracted from water and tritium is produced from lithium, which is found in the earth's crust. Significant progress is being done in making Fusion energy production a reality. ITER will be the biggest Fusion reactor on earth, but science and technology gaps still remain and need to be closed. Researchers and engineers together with policy makers require all needed tools to accomplish this new revolution in the history of the mankind, bringing the energy of the Sun and all other stars to the Earth to be used for peaceful purposes. Fusion power will give way to a new era where access to a clean, safe and unlimited energy will be a reality, an era with new challenges, such as the conquest of space. We will leave behind many of the problems that characterized 19th and 20th centuries. At last we will have at our disposal the power source required to navigate beyond our solar system.
Speaker: Mr Florentino Portero (Foundation Isaac Albéniz)
• 10:15 10:45
Coffee Break 30m
• 10:45 12:30
Overview 1: Magnetic Fusion: OV/1
Convener: Mr Kenichi Kurihara (NATIONAL INSTITUTES FOR QUANTUM AND RADIOLOGICAL SCIENCE AND TECHNOLOGY)
• 10:45
Extension of Operational Regime of LHD towards Deuterium Experiment 25m
Yasuhiko Takeiri for LHD Experiment Group The final goal of the LHD project is to obtain the high performance helical plasma relevant to the fusion reactor, i.e., ion and electron temperature T_i > 10 keV, volume averaged beta > 5 %, fusion triple product n_e tau_E Ti > 1020 keV m^-3 s, and long pulse length of more than 3600 s with heating power of 3 MW. In order to achieve this objective, the deuterium plasma is expected to have better energy and particle confinement than the hydrogen plasma, which is clearly seen in tokamaks, but is not always obvious in helical devices. As the finalization of the hydrogen experiment towards the deuterium phase, the exploration of the best performance of the hydrogen plasma was intensively performed in the Large Helical Device (LHD). High T_i, T_e, of more than 6 keV were simultaneously achieved by superimposing the high power electron cyclotron resonance heating (ECH) on the neutral beam injection (NBI) heated plasma. It was also demonstrated in hydrogen/helium discharges that experimental and numerical results imply the existence of the confinement improvement for heavier ions than proton. Another key parameter to present plasma performance is an averaged beta value. The high beta regime around 4 % was extended to an order of magnitude lower collisional regime than before. The pulse length has also become longer. In the last experimental campaign, the high performance plasma with e-ITB could successfully be maintained for more than 5 minutes. In such a long pulse discharge, it was found that the mixed-material deposition layer plays a key role in the wall retention and the particle recycling. These three results assured the start of the deuterium experiment from March 2017.
Speaker: Yasuhiko Takeiri (National Institute for Fusion Science)
• 11:10
Progress in ITER Construction, Manufacturing and R&D 25m
The ITER project is a critical step in the development of fusion energy: its role is to confirm the feasibility of exploiting magnetic confinement fusion for the production of energy for peaceful purposes by providing an integrated demonstration of the physics and technology required for a fusion power plant. Rapid progress is being made in the design, manufacturing, construction and R&D activities, and the facility is now taking shape at St-Paul-lez-Durance. Supported by impressive achievements in fusion technology R&D, manufacturing of ITER components is in full swing. The international collaboration formed around the production of superconducting magnets for ITER has produced over 600 t of Nb3Sn and almost 250 t of NbTi superconducting strand. 80% of the superconductors required for the ITER magnets are complete, and coil fabrication activities are underway in 6 of the 7 partners’ factories. Fabrication of the vacuum vessel is moving forward, with structures being manufactured under the responsibility of four contributing Domestic Agencies, manufacturing of the thermal shield is also in progress, and the first elements of the cryostat (~29 m diameter × ~29 m height) have been delivered to the ITER site. Substantial progress has also been achieved in prototyping and R&D activities in areas such as plasma facing components, in-vessel coils, H&CD systems, remote handling and power supplies in preparation for manufacturing. A wide-ranging physics R&D programme, closely integrated with the ITPA and the major fusion facilities in the ITER Members, is also addressing key issues impacting on finalization of the ITER design and preparations for operation. These R&D activities encompass studies of disruption mitigation, analysis of ELM control by magnetic perturbations, characterization of heat loads in stationary and transient plasma phases, plasma-wall interactions with all-metal PFCs, and studies of plasma scenarios for non-active and nuclear phases of the ITER experimental programme. The paper will review the progress made in developing the advanced technologies required for ITER and in the manufacturing activities for major tokamak components, discuss advances made in experimental and modelling studies of key physics issues, detail measures taken to establish a more effective project organization and present the status of construction of the ITER facility.
Speaker: Prof. Bernard Bigot (ITERFr)
• 11:35
DIII-D Research Advancing the Scientific Basis for Burning Plasmas and Fusion Energy 25m
The DIII-D tokamak has addressed key issues to advance the physics basis for burning plasmas for ITER and future steady-state fusion devices. Developments on ITER scenarios include the discovery of a new wide-pedestal variant of QH-mode where increased edge transport is found to allow higher pedestal pressure, consistent with peeling-ballooning theory, and complete ELM suppression in steady-state “hybrid” plasmas that is relatively insensitive to q95, having weak effect on the pedestal. Shattered pellet injection (SPI) has been shown an effective technique for runaway electron (RE) plateau dissipation. Mixed species shattered pellet injection (SPI) enabled control of disruption characteristics, while keeping the radiation fraction, divertor heat loads, and current quench times within ITER requirements. Reduced transport models such as TGLF reproduce the reduced confinement associated with additional electron heating in DIII-D ITER baseline plasmas. Density peaking can recover the performance, as can raising the pedestal density, which increases the pedestal pressure and can even give access to Super H-mode for ITER. Both high-qmin and hybrid steady-state plasmas have avoided fast ion instabilities and achieved increased performance by control of the fast ion pressure gradient and magnetic, and use of external control tools such as ECH. In the boundary, ExB drifts are found important for simulating observed asymmetries in divertor detachment, and the erosion rate of high Z materials is found to be reduced through control of the electric field in the presheath. Between-ELM heat flux asymmetries in the presence of RMP fields are eliminated in detached divertor conditions. Higher low-Z impurity concentrations in the background plasma are also found to reduce the net erosion rate of high-Z targets, even to the point of net deposition. These small-sample studies are being used to investigate high-Z impurity contamination efficiency from different divertor locations and impact on the core plasma performance, which in-turn inform the forthcoming metal divertor tile experiments.
Speaker: Dr Wayne M. Solomon (General Atomics)
• 12:00
Overview of the JET results in support to ITER 25m
The JET contributors Europe has elaborated a Roadmap to the realisation of fusion energy in which ‘ITER is the key facility and its success is the most important overarching objective of the programme’. We review the contribution of the recent JET experiments with the ITER first wall materials mix, and, the underlying physics understanding to mitigate the scientific risks identified in the ITER research plan. Indeed, together with the ITER scenario development, a strong focus on JET is pursued for addressing ITER needs and developing a sound physics basis for the extrapolation through first principle and integrated modelling: plasma wall interaction, disruption mitigation (installation of a third mitigation valve), H mode access, W-control with higher electron heating (ICRH ITER-like antenna re-instated), pellet ELMs pacing with the optimised vertical high field side track. The JET ITER-Like Wall experiment provides an insight in the coupling between tokamak-plasma operation and plasma-surface interaction in the unique Be/W material environment and acts as test-bed to verify models and modelling tools for ITER. Disruptions are considered as the highest programmatic risk in the ITER Research Plan and experimental and modelling effort in Europe and JET are reviewed. High spatial resolution Doppler backscattering measurements have revealed novel insights into the development of the edge transport barrier. The operational constraints of a metal wall can prevent reaching plasma energy confinement required for QD-T=10 on ITER. Progress on JET to mitigate this risk is reported aiming at maximizing the core and pedestal performance in stationary condition with the W divertor constrain. The measured D-D neutron fluence and gamma dose rates have been successfully compared with simulations performed with the codes used for ITER nuclear safety analyses. Finally, the benefit to further use JET beyond 2020 to train the international ITER team with an upgrade tungsten divertor and with the ITER control tools will be discussed. This work has been carried out within the framework of the EUROfusion Consortium and has received funding from the Euratom research and training programme 2014-2018 under grant agreement No 633053. The views and opinions expressed herein do not necessarily reflect those of the European Commission
Speaker: Dr xavier Litaudon (EUROfusion)
• 12:30 14:00
Lunch Break 1h 30m
• 14:00 16:10
Overview 2: Magnetic Fusion: OV/2
Convener: Predhiman Kaw (Institute for Plasma Research, India)
• 14:00
Overview of ASDEX Upgrade results 25m
The ASDEX Upgrade program is devoted to the preparation of ITER operation and the development of plasma scenarios and physics understanding for a future DEMO. Different scenario lines adapted to critical research tasks are developed and naturally integrated with the metallic, high-Z plasma facing components environment. The scenarios can be mainly divided into low core collisionality and high divertor collisionality conditions, which can be achieved simultaneously only in devices of ITER size. The development of non-inductive scenarios relies on low core collisionality and is performed with low neutral divertor pressure and an attached divertor. Fully non-inductive operation with a combination of NBCD and ECCD has been achieved at Ip= 0.8 MA and a safety factor q95= 5.4. The core W concentration is quite high as a consequence of the hot SOL and divertor which result in high W sputtering yields. A normalized exhaust power Psep/R of 10 MW/m has been achieved with nitrogen (N) seeding and a partially detached outer divertor at a total heating power of 25 MW. A high neutral deuterium divertor pressure was found to be essential for efficient divertor cooling. Confinement degradation connected to a high neutral pressure is partly compensated by improvement due to the effects of N. Investigations of the cause for the improved energy confinement with N seeding suggest that an inward shift of the pedestal density profile in relation to the temperature profile is the main driver for an enhanced pedestal stability which is associated with the improved confinement. The inward shift is attributed to a shrinking of a high density region on the high field side SOL due to the power reduction caused by the N radiation, which effects the fueling in the X-point region. The new pair of ICRF antennas with 3 straps has fulfilled the predicted reduction of tungsten release from connected limiters. The tungsten influx during antenna operation is similar or even slightly smaller compared to the 2-strap antenna pair with boron coated limiters. With the 3-strap antennas operated for central heating, generally a reduction of the central tungsten concentration is observed despite a still moderately increased W influx. A high local density in front of the antenna, achieved by well-tailored gas puffing, further optimizes ICRF operation with regard to coupling and low W release.
Speaker: Dr Arne Kallenbach (Max-Planck-Institut f. Plasmaphysik)
• 14:25
Overview of EAST Experiments on the Development of High-performance Steady-State Scenario 25m
EAST aims to demonstrate steady-state advanced high-performance H-mode plasmas with ITER-like configuration, plasma control and heating schemes. Since 2015, EAST has been equipped with all ITER-related auxiliary heating and current drive systems. Two NBI systems injected from Co- and Ctr-current directions, have been installed on EAST and allow the flexible study of the plasma rotation effect. A flexible in-vessel RMP coil system was installed in 2014 for active MHD instability control in order to achieve long-pulse steady-state operation in the EAST tokamak. Since then, EAST has been capable of investigating ELM control with most existing methods, including RMP, pellet-pacing, SMBI, LHW and Li-pellet injection. The exploration of fully non-inductive, high performance, upper single-null discharges with the tungsten divertor has been successfully demonstrated with upgrades of heating and current drive capabilities on EAST. A higher beta regime has been achieved with the 4.6 GHz LHCD and NBI. Experimental results show that LHWs at 4.6 GHz exhibit stronger current drive capability than at 2.45 GHz, in agreement with less pronounced parametric instability behavior with the 4.6 GHz LH wave. By means of the 4.6 GHz and 2.45 GHz LHCD systems, H-mode is obtained at relatively high density. A stationary ELM-stable H-mode regime has been achieved in EAST with 4.6 GHz LHCD. This regime allows nearly fully non-inductive long-pulse operations, exhibiting a relatively high pedestal and good global energy confinement with H98y2 near 1.2, good impurity control, and the capability of operation at relatively high density. Complete suppression of ELMs has been observed during the application of n = 1 and 2 RMPs on EAST. The experimental results show that the plasma response plays an important role in ELM control. Critical thresholds for the amplitude of the RMPs and the plasma rotation for this transition have been observed for the first time on EAST. The 3D edge magnetic topology has been applied for active control of heat and particle fluxes deposited on the divertor targets in steady-state operation on EAST. The impacts of the 3D magnetic topology on the edge plasma transport and heat flux distribution have been investigated using the EMC3-EIRENE code and found to be consistent with the experimental observation of strike-line splitting.
Speaker: Prof. Yunfeng Liang (Forschungszentrum Jülich GmbH, Germany)
• 14:50
Kinetics of Relativistic Runaway Electrons 25m
This overview talk covers recent developments in the theory of runaway electrons in tokamaks. Such electrons are known to be of serious concern with regard to safe operation of large-scale tokamaks in general and ITER in particular. They can quickly replace a large part of the bulk electron current during disruptions, and the corresponding magnetic energy exceeds the particle kinetic energy. This feature separates the time-scale of the runaway production from the time-scale of the current decay. The talk deals with the following physics aspects of the runaway evolution: (1) survival and acceleration of initially hot electrons during thermal quench, (2) effect of magnetic perturbations on runaway confinement, (3) multiplication of the runaways via knock-on collisions with the bulk electrons, (4) slow decay of the runaway current, and (5) runaway-driven micro-instabilities. Several theoretical groups internationally are currently addressing these aspects. The recent progress includes a first-principle description for the primary runaway electron production during the thermal quench, estimates of the runaway losses through partially destroyed magnetic flux surfaces, an improved description of fast electron collisions with heavy impurities within a Thomas-Fermi model for screening, a rigorous kinetic theory for relativistic runaways in the electric field that is close to the avalanche threshold, refined evaluation of the critical field for avalanche onset with a systematic description of knock-on collisions and radiative losses, demonstration of phase-space attractor that supports a peaked distribution function of the runaways, a model for current damping in a self-sustained regime of marginal criticality for the runaways, and reassessment of thresholds for the runaway-driven micro-instabilities. Some of these new theoretical findings are directly relevant to current experiments on DIII-D, ASDEX-U, and JET. They also provide an important input for ITER disruption modeling and runaway mitigation strategy. Work supported by the U.S. Department of Energy Contract No. DEFG02-04ER54742 and by ITER Contract No. ITER/CT/15/4300001178.
Speaker: Dr Boris Breizman (Institute for Fusion Studies, The University of Texas, Austin, Texas, 78712 USA)
• 15:15
Overview of the KSTAR Research in Support of ITER and DEMO 25m
The KSTAR device has been operated since the first plasma in 2008 with the mission of exploring the physics and technologies of high performance steady-state operation that are essential for ITER and fusion reactor. KSTAR has been focusing on maximizing performance and extending pulse length targeting H-mode dischage up to 300s at higher plasma current up to 2 MA, and at higher normalized beta (βN) up to ~5. In the 2015 campaign, various long pulse H-mode discharges have been operated after the improved plasma shape control, and a longest H-mode discharge was achieved up to 55 s at 0.6 MA in plasma current and 2.9 T in toroidal field utilizing 4.2 MW neutral beam and 0.65MW ECCD systems. This will be the longest H-mode discharge in tokamak devices. Fully non-inductive discharges have been carried out at the reduced plasma current of 0.4 MA to access the steady-state operation condition. The first fully non-inductive operation was achieved with relatively high plasma performances (βN ~ 2.1 and βP ~ 3.0). However the shot was terminated at about 16s due to the excessive heat in poloidal limiter. KSTAR device could be an ideal device to investigate the basic of the stability limits and confinement improvement utilizing unique features of KSTAR such as extremely low error field, versatile in-vessel control coils (IVCC), and advanced 2D/3D imaging diagnostics. In this paper, the progress of the KSTAR research to support ITER and DEMO will be reported.
Speaker: Dr Yeong-Kook Oh (National Fusion Research Institute)
• 15:40
Overview of High-Field Divertor Tokamak Results from Alcator C-Mod* 25m
C-Mod is the only divertor tokamak in the world capable of operating at B fields up to 8 T, equaling and exceeding that planned for ITER. C-Mod is compact, thus accessing regimes of extreme edge power density (1 MW/m2 average through the plasma surface). Scrape-off layer (SOL) power widths are of order of a few mm, with measured parallel power flows >1 GW/m2 at the divertor, surpassing the design for ITER, and approaching the levels envisioned in power plants. C-Mod results are particularly important for providing the physics basis of the high-field, compact tokamak approach, which can lead to a faster path in the development of fusion energy.[1] Results of experiments and related modeling, obtained since the last IAEA FEC meeting, span the topics of core transport and turbulence, RF heating and current drive, pedestal physics, scrape-off layer, divertor and plasma-wall interactions. ICRF has been successfully applied to control and reverse accumulation of high Z impurities in the core plasma. ICRF has also been employed to control and mitigate locked-modes induced by error fields. For the first time ever, feedback of low Z seeding for divertor power dissipation has been tied directly to real-time plasma power fluxes measured on the high-Z metal PFCs in the divertor, and used to mitigate those fluxes with no degradation of the pedestal pressure or core confinement. The naturally ELM-free I-mode regime has been up to BT=8T, and to double-null topology. I-mode threshold scalings show a weak dependence on B, yielding a significantly broader window for I-mode operation at high field. Quiescence of the high-field side scrape-off layer makes this a potentially attractive location for placement of RF actuators to ameliorate plasma interactions with launchers; the wave physics for penetration and damping, for both ICRF and LHRF appears very favorable for high-field side launch. BOUT++ edge plasma simulations are shedding important light on the nature of I-mode pedestal fluctuations which regulate impurity transport in this regime. LHRF has been employed as an actuator for controlling plasma rotation and rotation shear, critical parameters for turbulence control. A new disruption database has been populated, and used to identify key variables that could be used for disruption prediction and avoidance. [1] B. Sorbom et al. Fusion Eng. Design 100 (2015) 378.
Speaker: Dr Earl Marmar (Mass. Inst. of Technology)
• 14:00 18:45
Overview Poster: OV/P
• 14:00
3-D effects on transport and plasma control in the TJ-II stellarator 4h 45m
Recent improvements in diagnostics and operation have led to better understanding of 3-D effects on transport and plasma control in the TJ-II stellarator. Impurity transport: Direct measurments of electrostatic potential variations within the same magnetic flux surface in ECRH plasmas are presented. Calculations show that such asymmetries affect impurity accumulation. The asymmetry value and its observed dependency on the electric field are reproduced by neoclassical MC calculations. The dependence of the impurity confinement time on charge and mass has also been studied. Experiments have shown evidence of the influence of ECRH on turbulent mechanisms, increasing both the fluctuation level and the amplitude of Long-Range-Correlations as proxy of Zonal Flows (ZF), as well as affecting NC radial electric fields. Momentum transport and electromagnetic effects: Radial electric fields, ZF-like structures, time memory and radial correlations are modulated by low order rationals. It is shown that magnetic oscillations associated with rational surfaces play an key role in confinement transitions. Furthermore, evidence of the mutual interaction of NC and turbulent mechanisms in qualitative agreement with GK simulations is presented. Innovative power-exhaust scenarios using liquid metals: Novel solutions for plasma facing components based on liquid metals like Li and Sn/Li alloys have been developed. Biasing of Li limiters with respect to carbon ones has evidenced the role of the secondary electron emission of plasma exposed surfaces. Plasma stability studies: It has been shown that a reduction of magnetic well has a direct impact on fluctuations without reducing plasma confinement drastically, suggesting that Mercier stability calculations are missing some stabilization mechanisms. Plasma fuelling experiments and neutral dynamics: First core plasma fuelling experiments using a cryogenic pellet injector system are presented. The radial redistribution of particles can be understood qualitatively from NC predictions. First results on the impact of neutral fluctuations on the observed turbulent structures will be reported. Role of ECRH and iota profile on fast ion confinement: Results show that ECRH and iota-profile are potential tools for AE control. Coherent modes in NBI-heated plasmas are explained as global (GAE) and discrete shear-AEs induced by magnetic islands.
Speaker: Prof. Francisco Castejón (CIEMAT)
• 14:00
A Pathway to Laser Fusion Energy: Fast Ignition Realization EXperiment (FIREX) 4h 45m
Here we report recent progress of the fast ignition inertial confinement fusion demonstration. Fraction of low energy (< 1 MeV) component of the relativistic electron beam (REB), which efficiently heats the fuel core, increases by the factor of 4 by enhancing pulse contrast of heating laser and removing preformed plasma sources. Kilo-tesla magnetic field is studied to guide the diverging REB to the fuel core. The transport simulation of the REB accelerated by the heating laser in the externally applied and compressed magnetic field indicates that the REB can be guided efficiently to the fuel core. The integrated simulation shows >4% of the heating efficiency and > 4 keV of ion temperature are achievable by using GEKKO-XII and LFEX, properly designed cone-fuel and the external magnetic field.
Speaker: Prof. Hiroshi AZECHI (Institute of Laser Engineering, Osaka University)
• 14:00
DIII-D Research Advancing the Scientific Basis for Burning Plasmas and Fusion Energy 4h 45m
The DIII-D tokamak has addressed key issues to advance the physics basis for burning plasmas for ITER and future steady-state fusion devices. Developments on ITER scenarios include the discovery of a new wide-pedestal variant of QH-mode where increased edge transport is found to allow higher pedestal pressure, consistent with peeling-ballooning theory, and complete ELM suppression in steady-state “hybrid” plasmas that is relatively insensitive to q95, having weak effect on the pedestal. Shattered pellet injection (SPI) has been shown an effective technique for runaway electron (RE) plateau dissipation. Mixed species shattered pellet injection (SPI) enabled control of disruption characteristics, while keeping the radiation fraction, divertor heat loads, and current quench times within ITER requirements. Reduced transport models such as TGLF reproduce the reduced confinement associated with additional electron heating in DIII-D ITER baseline plasmas. Density peaking can recover the performance, as can raising the pedestal density, which increases the pedestal pressure and can even give access to Super H-mode for ITER. Both high-qmin and hybrid steady-state plasmas have avoided fast ion instabilities and achieved increased performance by control of the fast ion pressure gradient and magnetic, and use of external control tools such as ECH. In the boundary, ExB drifts are found important for simulating observed asymmetries in divertor detachment, and the erosion rate of high Z materials is found to be reduced through control of the electric field in the presheath. Between-ELM heat flux asymmetries in the presence of RMP fields are eliminated in detached divertor conditions. Higher low-Z impurity concentrations in the background plasma are also found to reduce the net erosion rate of high-Z targets, even to the point of net deposition. These small-sample studies are being used to investigate high-Z impurity contamination efficiency from different divertor locations and impact on the core plasma performance, which in-turn inform the forthcoming metal divertor tile experiments.
Speaker: Dr Wayne M. Solomon (Princeton Plasma Physics Laboratory)
• 14:00
Edge and divertor plasma: detachment, stability, and plasma-wall interactions 4h 45m
The processes involving edge plasma and plasma-material interactions in magnetic fusion devices are very multifaceted and include a wide spectrum of phenomena ranging from plasma turbulence and meso-scale stability, recycling and transport processes of hydrogen species in the wall material, to the modification of wall material properties. In many cases these processes are strongly coupled and exhibit synergistic effects. Here we present the results of our studies of a wide range of edge plasma related issues: Our numerical simulations solve a long standing dispute on the roles of impurity radiation loss, plasma volumetric recombination, and ion-neutral friction in the rollover of the plasma flux to the target, which is the manifestation of detachment. We show that the rollover is caused by the increase of the impurity radiation loss and volumetric plasma recombination while the ion-neutral friction, although important for establishing necessary edge plasma conditions, does not contribute per se . With numerical modeling and theoretical analysis we consider stability of detachment and show that the absorption/desorption of hydrogen and impurity species from the wall can be crucial for a global stability of detached plasma. We also identify different mechanisms of meso-scale thermal instabilities driven by impurity radiation and resulting in a self-sustained oscillations of edge plasma parameters. We consider a trapping of He, which is an intrinsic impurity of fusion plasmas, in the first wall tungsten material. Our newly developed model, accounting for the generation of additional He traps caused by He bubble growth, fits all available experimental data on the layer of nano-bubbles observed in W under irradiation of low energy He plasma. Finally, we report on an impact of sheared magnetic field on the dynamics of blobs and ELM filaments playing an important role in edge and SOL plasma transport. This material is based upon work supported by the U.S. Department of Energy, Office of Science, Office of Fusion Energy Sciences under Award Numbers DE-FG02-04ER54739, DE-FG02-06ER54852, DE-SC0010413, and through the Scientific Discovery through Advanced Computing (SciDAC) program on Plasma Surface Interactions, funded by U. S. Department of Energy, Office of Science, Advanced Scientific Computing Research and Fusion Energy Sciences under Award Number DE-SC0008660.
Speaker: Prof. Sergei Krasheninnikov (University California San Diego)
• 14:00
Extension of Operational Regime of LHD towards Deuterium Experiment 4h 45m
The final goal of the LHD project is to obtain the high performance helical plasma relevant to the fusion reactor, i.e., ion and electron temperature T_i > 10 keV, volume averaged beta > 5 %, fusion triple product n_e tau_E Ti > 1020 keV m^-3 s, and long pulse length of more than 3600 s with heating power of 3 MW. In order to achieve this objective, the deuterium plasma is expected to have better energy and particle confinement than the hydrogen plasma, which is clearly seen in tokamaks, but is not always obvious in helical devices. As the finalization of the hydrogen experiment towards the deuterium phase, the exploration of the best performance of the hydrogen plasma was intensively performed in the Large Helical Device (LHD). High T_i, T_e, of more than 6 keV were simultaneously achieved by superimposing the high power electron cyclotron resonance heating (ECH) on the neutral beam injection (NBI) heated plasma. It was also demonstrated in hydrogen/helium discharges that experimental and numerical results imply the existence of the confinement improvement for heavier ions than proton. Another key parameter to present plasma performance is an averaged beta value. The high beta regime around 4 % was extended to an order of magnitude lower collisional regime than before. The pulse length has also become longer. In the last experimental campaign, the high performance plasma with e-ITB could successfully be maintained for more than 5 minutes. In such a long pulse discharge, it was found that the mixed-material deposition layer plays a key role in the wall retention and the particle recycling. These three results assured the start of the deuterium experiment from March 2017.
Speaker: Yasuhiko Takeiri (National Institute for Fusion Science)
• 14:00
First plasma operation of Wendelstein 7-X 4h 45m
The main objective of the optimized stellarator Wendelstein 7-X (W7-X) is the demonstration of steady-state plasma operation at fusion relevant plasma parameters thereby verifying that the stellarator is a viable fusion power plant concept. The design of W7-X is based on an elaborate optimization procedure to overcome the shortcomings of the concept. After completing the main construction phase of W7-X and successfully commissioning the device, first plasma operation started in December 2015. Plasma operation of W7-X follows a staged approach according to the successive completion of the in-vessel components. During the first operational phase five inboard limiters defined the last closed flux surface. Subsequently, W7-X will be equipped with a test divertor unit and eventually with a steady-state capable high heat flux divertor including active water cooling of all in-vessel components. Integral commissioning of plasma start-up and operation using an electron cyclotron resonance heating (ECRH) and an extensive set of plasma diagnostics, and initial physics studies during the first operational campaign have been successfully completed. Both in helium and hydrogen, plasma break-down was easily achieved. Gaining experience with plasma vessel conditioning, discharge lengths could be continuously extended. Eventually, discharges lasted up to 6 sec, reaching an injected energy of 4 MJ which is twice the limit originally agreed for the limiter configuration. At higher powers of 4 MW and central electron densities of 4.5x10^{19} m^{-3}, central temperatures reached values of 7 keV for the electrons and just above 2 keV for the ions. Important physics studies during this first operational phase include the assessment of the heat load distribution over the inboard limiters changing the toroidal phase and amplitude of deliberately applied error fields, impurity injection and confinement experiments including the effect of the rotational transform, and ECRH power deposition and heat pulse propagation experiments. Also a first assessment of the central electron root confinement, 2nd harmonic O-mode ECRH using multi-pass absorption, and the investigation of confinement and stability of discharges with co- and counter current drive (ECCD) have been achieved. This paper will give an overview of the results of the first experimental campaign of W7-X.
Speaker: Prof. Robert Wolf (Max-Planck-Institute for Plasma Physics)
• 14:00
H-mode and Non-Solenoidal Startup in the Pegasus Ultralow-A Tokamak 4h 45m
Studies at near-unity aspect ratio offer unique insights into the high confinement (H-mode) regime and support development of novel startup scenarios. Ohmic H-mode operation has been attained at A < 1.3. Edge plasma parameters permit probe measurements of the edge pedestal, including the local current density profile, with high spatial and temporal resolution. H-mode plasmas have standard L-H transition phenomena: a drop in D_alpha radiation; the formation of pressure and current pedestals; field-aligned filament ejection during ELMs; and a doubling of energy confinement time from H_98 ~ 0.5 to ~1. The L-H power threshold P_LH increases monotonically with n_e, consistent with the ITPA08 empirical scaling used for ITER and the theoretical FM3 model. Unlike at high A, P_LH is comparable in limited and single-null diverted topologies at A ~ 1.2, consistent with FM3 predictions. The magnitude of P_LH exceeds ITPA scalings by an order of magnitude, with P_LH/P_ITPA08 increasing as A approaches 1. Multiple n modes are observed during two classes of ELMs, consistent with excitation of multiple peeling-ballooning modes. Small, Type III-like ELMs occur at P_OH ~ P_LH with n <= 4. Large, Type-I-like ELMs occur with P_OH > P_LH and intermediate 5 < n < 15. Helical edge current injection appears to suppress Type III ELM activity. J_edge(R,t) measurements across single ELMs show the nonlinear generation and expulsion of current-carrying filaments during the ELM crash. Local Helicity Injection (LHI) offers a nonsolenoidal tokamak startup technique. Helicity is injected via current sources at the plasma edge. A circuit model that treats the plasma as a resistive element with time-varying inductance reasonably predicts I_p(t). The electron confinement governs the power balance. Initial measurements show peaked T_e and pressure profiles, which are comparable to Ohmic-like transport or moderately stochastic confinement. Extrapolation suggests I_p ~ 1 MA may be achievable in NSTX-U. Resistive MHD simulations suggest I_p is built from current rings injected during reconnection between unstable helical current streams. Several experimental observations support this model: imaging of the merging current streams; n=1 MHD activity and discrete current stream localized in the plasma edge; and anomalously high impurity ion heating in the edge region.
Speaker: Prof. Raymond Fonck (University of Wisconsin-Madison)
• 14:00
Hysteresis and Fast Timescale in Transport Relation of Toroidal Plasmas 4h 45m
This article assesses the understanding of and impacts by the hysteresis of transport relation. The rapid changes of fluxes compared to slow changes of plasma parameters are overviewed for both edge barrier and core plasmas. The theoretical approaches to understand the direct influence of heating power on turbulent transport are addressed. The advanced data analysis method to search the hysteresis in gradient-flux relation is explained. Finally, the importance of transport hysteresis on the control system of fusion device is discussed. The modulation ECH experiment, in which the heating power repeats on-and-off periodically, revealed the hysteresis and fast changes in gradient-flux relation. The decisive progress is that both the hystereses in the gradient-flux and gradient-fluctuation relations were observed simultaneously. Analyses of observations that can be interpreted as the hysteresis have been undertaken on various experiments: LHD, ASDXE-U, DIII-D, HL-2A, JFT-2M, JT-60U, KSTAR, TJ-II, and W7-AS. Hysteresis with rapid timescale exists in the channels of energy, electron and impurity, and plausibly in momentum. The causes of hysteresis and fast timescale are discussed. The nonlocal-in-space coupling works here, but does not suffice. One mechanism for ‘the heating heats turbulence’ is that the external source S in phase space for heating has its fluctuation in turbulent plasma: S[f] = S[f0] + (dS/df) δf, where δf is the perturbation of distribution function. This coupling can induce the direct input of heating power into fluctuations. The height of the jump in transport hysteresis is smaller for heavier hydrogen isotope, and is one of origins of isotope effect on confinement. Advanced methods of data analysis are overviewed. The transport hysteresis can be studied by observing the higher harmonics of temperature perturbation δTm in heating modulation experiments. The hysteresis introduces the term δTm , which depends on the harmonic number m in algebraic manner (not exponential decay). The impacts of transport hysteresis on the control system are assessed. The control system must be designed so as to protect the system from sudden plasma loss. Thermonuclear instability due to fusion power is also discussed.
Speaker: Prof. Kimitaka Itoh (NIFS)
• 14:00
Implementation within the European Domestic Agency of the French nuclear safety Order of 2012, concerning Basic Nuclear Installation, applicable to ITER Project. 4h 45m
The ITER project is being undertaken at Cadarache, France, to construct and operate an experimental nuclear fusion facility. The aim of this paper is the description of the implementation of the French Order of February 7, 2012, concerning Nuclear Installation (called Installation Nuclear de Base, INB) in France within the European Domestic Agency (EU-DA). For protection of Public Safety, Health and Salubrity, and of Nature and Environment, the French order (INB Order 07/02/2012) establishes general rules relating to the Design, Construction, Operation, Final shutdown, Dismantling, Surveillance and Maintenance of Nuclear facilities during their full life cycle. The INB Order applies namely to the operator (ITER Organisation) and involves the whole supply chain. The EU-DA, as a tier 1 supplier, has duties regarding the compliance with the requirements propagated from the INB Order, mainly the dispositions to be propagated from the nuclear operator to the chain of suppliers performing Protection Important Activities (PIA), called external interveners in the INB order, in the contracts. Among other nuclear regulations in force in France, presently encoded in the French Environemental Code as, the Nuclear Pressurized Equipment regulation (ESPN), the INB Order addresses domains where the EU-DA shall play a prime role in providing to the nuclear operator reliable evidences and sound demonstration in organisation and responsibilities, nuclear safety demonstration, traceability, validation of methods, qualifications, calculations and modelling, … The EU-DA applies a Requirements Management and Verification (RMV) process in order to track, control and verify all technical requirements applicable to ITER components under the EU-DA responsibility. This process is applied to the nuclear safety defined requirements in a way that allows all defined requirements on ITER components to be recorded and controlled at all the different levels of the supply chain in a systematic way. Finally the communication performed within the EU-DA organization and the supply chain to continuously improve the nuclear safety culture, which is a first priority of the ITER project, will be presented.
Speaker: Mr Paul Wouters (Fusion For Energy)
• 14:00
Kinetics of Relativistic Runaway Electrons 4h 45m
This overview talk covers recent developments in the theory of runaway electrons in tokamaks. Such electrons are known to be of serious concern with regard to safe operation of large-scale tokamaks in general and ITER in particular. They can quickly replace a large part of the bulk electron current during disruptions, and the corresponding magnetic energy exceeds the particle kinetic energy. This feature separates the time-scale of the runaway production from the time-scale of the current decay. The talk deals with the following physics aspects of the runaway evolution: (1) survival and acceleration of initially hot electrons during thermal quench, (2) effect of magnetic perturbations on runaway confinement, (3) multiplication of the runaways via knock-on collisions with the bulk electrons, (4) slow decay of the runaway current, and (5) runaway-driven micro-instabilities. Several theoretical groups internationally are currently addressing these aspects. The recent progress includes a first-principle description for the primary runaway electron production during the thermal quench, estimates of the runaway losses through partially destroyed magnetic flux surfaces, an improved description of fast electron collisions with heavy impurities within a Thomas-Fermi model for screening, a rigorous kinetic theory for relativistic runaways in the electric field that is close to the avalanche threshold, refined evaluation of the critical field for avalanche onset with a systematic description of knock-on collisions and radiative losses, demonstration of phase-space attractor that supports a peaked distribution function of the runaways, a model for current damping in a self-sustained regime of marginal criticality for the runaways, and reassessment of thresholds for the runaway-driven micro-instabilities. Some of these new theoretical findings are directly relevant to current experiments on DIII-D, ASDEX-U, and JET. They also provide an important input for ITER disruption modeling and runaway mitigation strategy. Work supported by the U.S. Department of Energy Contract No. DEFG02-04ER54742 and by ITER Contract No. ITER/CT/15/4300001178.
Speaker: Prof. Boris Breizman (The University of Texas at Austin)
• 14:00
Overview of ASDEX Upgrade results 4h 45m
The ASDEX Upgrade program is devoted to the preparation of ITER operation and the development of plasma scenarios and physics understanding for a future DEMO. Different scenario lines adapted to critical research tasks are developed and naturally integrated with the metallic, high-Z plasma facing components environment. The scenarios can be mainly divided into low core collisionality and high divertor collisionality conditions, which can be achieved simultaneously only in devices of ITER size. The development of non-inductive scenarios relies on low core collisionality and is performed with low neutral divertor pressure and an attached divertor. Fully non-inductive operation with a combination of NBCD and ECCD has been achieved at Ip= 0.8 MA and a safety factor q95= 5.4. The core W concentration is quite high as a consequence of the hot SOL and divertor which result in high W sputtering yields. A normalized exhaust power Psep/R of 10 MW/m has been achieved with nitrogen (N) seeding and a partially detached outer divertor at a total heating power of 25 MW. A high neutral deuterium divertor pressure was found to be essential for efficient divertor cooling. Confinement degradation connected to a high neutral pressure is partly compensated by improvement due to the effects of N. Investigations of the cause for the improved energy confinement with N seeding suggest that an inward shift of the pedestal density profile in relation to the temperature profile is the main driver for an enhanced pedestal stability which is associated with the improved confinement. The inward shift is attributed to a shrinking of a high density region on the high field side SOL due to the power reduction caused by the N radiation, which effects the fueling in the X-point region. The new pair of ICRF antennas with 3 straps has fulfilled the predicted reduction of tungsten release from connected limiters. The tungsten influx during antenna operation is similar or even slightly smaller compared to the 2-strap antenna pair with boron coated limiters. With the 3-strap antennas operated for central heating, generally a reduction of the central tungsten concentration is observed despite a still moderately increased W influx. A high local density in front of the antenna, achieved by well-tailored gas puffing, further optimizes ICRF operation with regard to coupling and low W release.
Speaker: Dr Arne Kallenbach (Max-Planck-Institut f. Plasmaphysik)
• 14:00
Overview of DEMO Safety R&D and the Potential Future Role of IEA ESEFP IA 4h 45m
A fusion DEMO reactor, like other advanced nuclear energy systems, must satisfy several goals including a high level of public and worker safety, low environmental impact, high reactor availability, a closed fuel cycle, and the potential to be economically competitive. The experience of the ITER project will facilitate DEMO programs in developing a safety approach and a safety design, performing safety analyses under the scrutiny of a nuclear regulator, ensuring device availability, managing the radioactive waste, and conducting economic assessments. However, there are still large scientific and technological gaps between the current ITER and DEMO reactors. In this paper international fusion safety research relevant to DEMO will be summarized following the lessons learned from ITER. The main scientific and technological challenges will be presented by considering the differences between fission and fusion reactors as well as the corresponding implications on DEMO design and operation, with perspectives not only from fusion energy development but also from the development of Generation-IV fission reactors. Potential research topics for international collaboration will also be addressed with emphasis on the International Energy Agency (IEA) implementing agreement (IA) on a cooperative program on Environmental, Safety and Economic aspects of Fusion Power (ESEFP).
Speaker: Prof. Yican Wu (Institute of Nuclear Energy Safety Technology (INEST),Chinese Academy of Sciences)
• 14:00
Overview of EAST Experiments on the Development of High-performance Steady-State Scenario 4h 45m
EAST aims to demonstrate steady-state advanced high-performance H-mode plasmas with ITER-like configuration, plasma control and heating schemes. Since 2015, EAST has been equipped with all ITER-related auxiliary heating and current drive systems. Two NBI systems injected from Co- and Ctr-current directions, have been installed on EAST and allow the flexible study of the plasma rotation effect. A flexible in-vessel RMP coil system was installed in 2014 for active MHD instability control in order to achieve long-pulse steady-state operation in the EAST tokamak. Since then, EAST has been capable of investigating ELM control with most existing methods, including RMP, pellet-pacing, SMBI, LHW and Li-pellet injection. The exploration of fully non-inductive, high performance, upper single-null discharges with the tungsten divertor has been successfully demonstrated with upgrades of heating and current drive capabilities on EAST. A higher beta regime has been achieved with the 4.6 GHz LHCD and NBI. Experimental results show that LHWs at 4.6 GHz exhibit stronger current drive capability than at 2.45 GHz, in agreement with less pronounced parametric instability behavior with the 4.6 GHz LH wave. By means of the 4.6 GHz and 2.45 GHz LHCD systems, H-mode is obtained at relatively high density. A stationary ELM-stable H-mode regime has been achieved in EAST with 4.6 GHz LHCD. This regime allows nearly fully non-inductive long-pulse operations, exhibiting a relatively high pedestal and good global energy confinement with H98y2 near 1.2, good impurity control, and the capability of operation at relatively high density. Complete suppression of ELMs has been observed during the application of n = 1 and 2 RMPs on EAST. The experimental results show that the plasma response plays an important role in ELM control. Critical thresholds for the amplitude of the RMPs and the plasma rotation for this transition have been observed for the first time on EAST. The 3D edge magnetic topology has been applied for active control of heat and particle fluxes deposited on the divertor targets in steady-state operation on EAST. The impacts of the 3D magnetic topology on the edge plasma transport and heat flux distribution have been investigated using the EMC3-EIRENE code and found to be consistent with the experimental observation of strike-line splitting.
Speaker: Dr Baonian Wan (Institute of Plasma Physics, Chinese Academy of Sciences)
• 14:00
Overview of First Results from NSTX-U and Analysis Highlights from NSTX 4h 45m
The National Spherical Torus Experiment (NSTX) has undergone a major upgrade, and NSTX Upgrade (NSTX-U) is now the most capable Spherical Torus/Tokamak (ST) in the world program. NSTX-U mission elements include: exploring unique ST parameter regimes to advance predictive capability for ITER and beyond, developing solutions for the plasma-material interface challenge, and advancing the ST as a possible Fusion Nuclear Science Facility or Pilot Plant. NSTX-U has two major new tools including a new central magnet and new 2nd more tangential neutral beam injector (NBI). Plasma control commissioning and scenario development has proceeded rapidly on NSTX-U. Diverted plasmas with IP = 0.8MA, BT = 0.6T, and tau-pulse ~ 1s are obtained routinely, and sustained H-mode plasmas have been accessed with 2.5MW of NBI heating power. Peak parameters achieved during the first run-month of NSTX-U plasma operation include: NBI power ~4MW, IP = 1MA, stored energy ~ 200kJ, beta-N ~ 4, kappa ~ 2.2, tau-E-tot > 50ms, tau-pulse ~ 1.7s, and a 50% increase in pulse-length from n=1 error field correction. Expected results from the first run campaign include assessments of: core and pedestal confinement at lower collisionality via 60% higher field and current than NSTX, fast-ion confinement and current drive from the new 2nd NBI, and stability and control of high-kappa and high beta-N plasmas. Extensive analysis of NSTX results continued including novel analysis of: edge turbulence data during the L-to-H-mode transition, heat flux footprint narrowing with increasing amplitude of edge-localized modes, and gyrokinetic modeling of core turbulence from dissipative trapped electron mode and electron temperature gradient modes. Further, a unified kinetic resistive wall mode physics model has been developed, and Massive Gas Injection valves similar to proposed ITER valves will be tested on NSTX-U. Lastly, a new method for determining the saturation level for Alfvén Eigenmodes has been developed, and SOL power losses for RF heating modeled and interpreted with the AORSA code. Results from the first research campaign of NSTX-U will be presented, initial comparisons between NSTX-U and NSTX results described, and NSTX analysis highlights presented.
Speaker: Dr Jonathan Menard (Princeton Plasma Physics Laboratory)
• 14:00
Overview of High-Field Divertor Tokamak Results from Alcator C-Mod 4h 45m
C-Mod is the only divertor tokamak in the world capable of operating at B fields up to 8 T, equaling and exceeding that planned for ITER. C-Mod is compact, thus accessing regimes of extreme edge power density (1 MW/m2 average through the plasma surface). Scrape-off layer (SOL) power widths are of order of a few mm, with measured parallel power flows >1 GW/m2 at the divertor, surpassing the design for ITER, and approaching the levels envisioned in power plants. C-Mod results are particularly important for providing the physics basis of the high-field, compact tokamak approach, which can lead to a faster path in the development of fusion energy.[1] Results of experiments and related modeling, obtained since the last IAEA FEC meeting, span the topics of core transport and turbulence, RF heating and current drive, pedestal physics, scrape-off layer, divertor and plasma-wall interactions. ICRF has been successfully applied to control and reverse accumulation of high Z impurities in the core plasma. ICRF has also been employed to control and mitigate locked-modes induced by error fields. For the first time ever, feedback of low Z seeding for divertor power dissipation has been tied directly to real-time plasma power fluxes measured on the high-Z metal PFCs in the divertor, and used to mitigate those fluxes with no degradation of the pedestal pressure or core confinement. The naturally ELM-free I-mode regime has been up to BT=8T, and to double-null topology. I-mode threshold scalings show a weak dependence on B, yielding a significantly broader window for I-mode operation at high field. Quiescence of the high-field side scrape-off layer makes this a potentially attractive location for placement of RF actuators to ameliorate plasma interactions with launchers; the wave physics for penetration and damping, for both ICRF and LHRF appears very favorable for high-field side launch. BOUT++ edge plasma simulations are shedding important light on the nature of I-mode pedestal fluctuations which regulate impurity transport in this regime. LHRF has been employed as an actuator for controlling plasma rotation and rotation shear, critical parameters for turbulence control. A new disruption database has been populated, and used to identify key variables that could be used for disruption prediction and avoidance. [1] B. Sorbom et al. Fusion Eng. Design 100 (2015) 378.
Speaker: Dr Earl Marmar (Mass. Inst. of Technology)
• 14:00
Overview of MST Reversed Field Pinch Research in Advancing Fusion Science 4h 45m
The reversed field pinch (RFP) offers unique capabilities that could be essential to closing gaps to fusion power. The RFP has large plasma current and small toroidal field, with q(r)<1. Two key benefits arise: (1) the possibility for ohmic heating to ignition and (2) minimization of the field strength at the magnets. The material boundary can be made invisible to an inductive electric field, and the first-wall need not accommodate power injection ports or antennas. These features could help achieve a maintainable and reliable fusion power source. This overview summarizes MST results important for the advancement of the RFP as well as for improved understanding of toroidal confinement generally. Evidence for first observations of trapped-electron mode (TEM) turbulence in the RFP is obtained. Short-wavelength density fluctuations exhibit a density-gradient threshold, and GENE modeling predicts unstable TEM’s. Core-localized neutral beam injection stimulates bursty modes with both Alfvenic and EPM scaling. One mode agrees with a new analytic theory for the magnetic-island-induced Alfven eigenmode (MIAE), which conspires with an EPM to affect fast ion transport. At high current the RFP transitions to the quasi-single-helicity (QSH) state. A method to control the locked phase of QSH has been developed using resonant magnetic perturbations (RMP). Runaway electrons that appear without RMP are suppressed. An improved model for simultaneous interactions of multiple tearing modes and error fields has been developed. The RFP’s tearing-relaxation behavior together with well-developed theory and computation create a ripe opportunity for rigorous validation of MHD models. Integrated data analysis (IDA) complements validation by maximizing the information embedded in multiple diagnostics, which is essential for future fusion development steps having limited diagnostics. Using IDA methods, meta-diagnostics that combine charge-exchange recombination spectroscopy, x-ray tomography, and Thomson scattering yield more robust measurements of Z_eff and T_e, critical parameters for MHD. Nonlinear studies using an extended MHD model including drift and two-fluid physics in NIMROD show features similar to MST observations, including a tendency for the MHD and Hall emf terms to oppose each other in Ohm’s law, and opposition of the Maxwell and Reynolds stresses in momentum balance.
Speaker: Prof. John Sarff (University of Wisconsin-Madison)
• 14:00
Overview of progress in European Medium Sized Tokamaks towards an integrated plasma-edge/wall solution 4h 45m
Progress in tackling the edge challenge within the new European Medium Size Tokamak Task Force (EU-MST) will be reported. EU-MST coordinates research on ASDEX Upgrade, MAST and TCV supported by the local teams. The challenge is approached from two directions. On the one hand plasma regimes reducing the transient heat loads whilst trying to maintain high confinement are developed with active ELM control techniques and natural small ELM scenarios. On the other hand divertor solutions with detachment control and advanced magnetic configurations are studied. The type-I ELM energy flux is found to never exceed a value proportional to the pedestal top pressure times the minor radius and is never lower than 1/3 of this maximum value. Interestingly the actively controlled type-I ELMs with resonant magnetic perturbations on MAST and AUG also fit well into this operational band and so to reduce heat loads well below the ITER material limits a change of ELM regime is likely required. Such RMP aided small ELM regimes have been found at low and high density, although the low density, low collisionality regimes are often accompanied by an unacceptable drop in confinement. For the high-density regimes, also accessed without RMPs, evidence is mounting that a high scrape-off layer (SOL) density is a key parameter. The RMP ELM control has been found to be transferable to He plasmas. New data from TCV on ELM control with vertical kicks and edge ECRH will also be presented. Partial detachment of the divertor and its control is also a part of the integrated solution. The X-point radiation in N seeded discharges at high P/R has now been identified as a suitable control parameter. The studies of detachment have been extended to advanced divertor configurations experimentally on TCV and theoretically on MAST. Here target heat loads are reduced by geometrical means as well as by volumetric processes. Predicted impurity trapping between the two X-points of a snowflake configuration could further aid the heat flux reduction. SOL flows and filamentary transport become important in understanding the power distribution between the different strike zones. With RMPs, lobe structures form that locally increase the heat load and may influence the divertor radiation. The access to a wide parameter space, new concepts and integration within EU-MST is instrumental for progress in this area.
Speaker: Dr Hendrik Meyer (CCFE)
• 14:00
Overview of Recent COMPASS Activities 4h 45m
The COMPASS tokamak is one of the present devices operating with an ITER-like plasma shape. Its flexibility combined to an extensive set of diagnostics and NBI heating allow to address a broad range of key areas in support of the worldwide fusion programme such as H-mode, MHD, RAE, disruptions, PWI. The recent results obtained in COMPASS addressing these key issues are reviewed here. The control and characterization of the L-H transition and the pedestal physics represent a large part of the COMPASS scientific programme. Recycling and actuators such as the X-point height play a significant role in accessing H-mode. GAMs oscillating at frequencies 25-40 kHz are observed in L-mode discharges, increasing with the ion mass and with a decreasing amplitude from D to H plasmas. COMPASS also contributes to multi-machine databases with pedestal and SOL width scalings studies. Using perturbation coils, the influence of 3D fields on the strike-points splitting, ELM control and transport is reported. The MHD modes studies mainly concern the plasma interaction with RMPs, the characterization of Alfven-like modes and disruption/mitigation experiments. High frequency quasi-coherent oscillations (ranging from 200 kHz to above 1 MHz) that follow Alfvenic frequency scaling are observed in ohmic discharges. An extensive experimental study of MHD effects in losses of runaway electrons has been performed. In the field of disruptions, an inter-machine empirical scaling of critical magnetic disruption precursors has been developed, as well as the study of the disruptions toroidal asymmetry. The exhaust and plasma-material interaction studies in COMPASS contributed to the ITER divertor monoblocks design as part of the ITPA. Power deposition on leading edge was investigated both experimentally (inner-wall limiters with gaps and leading edges viewed by a high-resolution IR camera) and numerically (PIC simulations), with the latter reproducing well this experiment and the recent lamella melting experiment on JET. The ITER monoblocks shaping was also investigated in the frame of an ITER contract. Comparison with the deposited power from ion orbit calculations are consistent and confirm results presented at the previous IAEA FEC. However, the role of the E-field and the contribution from the electrons on the total power flux accounted in PIC calculations predict marginally lower power.
Speaker: Dr Renaud Dejarnac (CzIPP)
• 14:00
Overview of Recent Experimental Results from Aditya Tokamak 4h 45m
Several experiments, related to controlled thermonuclear fusion research and highly relevant for large size tokamaks including ITER, have been carried out in ADITYA, an ohmically heated circular limiter tokamak. Repeatable plasma discharges of maximum plasma current of ~ 160 kA and discharge duration beyond ~ 250 ms with plasma current flattop duration of ~ 140 ms has been obtained for the first time in ADITYA. The discharge reproducibility has been improved considerably with Lithium wall conditioning and improved plasma discharges are obtained by precisely controlling the plasma position. In these discharges, chord-averaged electron density ~ 3.0 – 4.0 x 10^19 m^-3 using multiple hydrogen gas puffs, electron temperature of the order of ~ 500 - 700 eV have been achieved. Novel experiments related to disruption control are carried out and disruptions, induced by hydrogen gas puffing are successfully mitigated using biased electrode and ICR pulse techniques. Runaway electrons are successfully mitigated by applying a short local vertical field (LVF) pulse. A thorough disruption database has been generated by identifying the different categories of disruption. Detailed analysis of several hundred disrupted discharges showed that the current quench time is inversely proportional to q_edge. Formation of current filaments are observed during most of the disruptions, which helps in identifying the cause of disruption. Apart from this, for volt-sec recovery during the plasma formation phase, low loop voltage start-up and current ramp-up experiments have been carried out using ECRH and ICRH. Successful recovery of volt-sec leads to achievement of longer plasma discharge durations. In order to achieve better coupling of lower hybrid waves to the plasma, multipl e gas puffs are injected prior to the launch of lower hybrid waves. The experiments showed considerable reduction in the reflection co-efficient indicating better absorption of LH waves in plasma. In addition to that Neon gas puff assisted radiative improved confinement mode has also been achieved in ADITYA. Further, the electrode biasing experiments have shown that during transition to better confinement mode, the Drift-Alfven fluctuations are suppressed and the current profile gets modified near the edge plasma region. In this paper, all the above mentioned experiments will be discussed.
Speaker: Mr Rakesh Tanna (Institute For Plasma Research)
• 14:00
Overview of Recent Experiments on HL-2A Tokamak 4h 45m
Recent experiments on the HL-2A tokamak have been aimed at the major challenges relevant to ITER operation and fusion energy development. Significant progress has been achieved in many areas, including the first demonstration of high coupling efficiency of LHCD passive-active multi-junction (PAM) antenna in H-mode discharges, pedestal instability and dynamics, ITB formation mechanism, energetic particle physics, ELM and disruption mitigation, real-time control of tearing modes with ECRH, etc.. A new PAM antenna as an LHCD launcher was designed and installed on the HL-2A tokamak. A high coupling efficiency was demonstrated under NBI heated ELMy H-mode plasmas. This was the first time that PAM antenna was applied in H-mode. The effects of LHCD on ELM mitigation and control of heat load on divertor plate were also observed. It was found that impurity accumulation and relaxation in the edge could trigger a series of I-H-I transitions through the excitation of a broadband (50-150 kHz) electromagnetic (EM) turbulence. EM turbulence could also be excited by impurity injection via laser blow-off. An improved confinement with complete suppression of ELMs was achieved by this technique. These findings reveal the underlying physics of how impurity affects the pedestal evolution, and suggest an important method to actively control pedestal via impurity-excited EM turbulence. An inward particle flux induced by a quasi-coherent mode at frequency 40-60 kHz was found to be responsible for the dramatic changes of the gradients in pedestal and the triggering of ELMs. Dependence of the correlation of resistive ballooning modes and trapped electron modes on electron temperature increase was observed experimentally. Formation of the ion ITB was found to be closely related to the Te/Ti ratio. A new nonlocal transport phenomenon triggered by the fishbone was observed and demonstrated to be caused by electromagnetic fluctuations. High-frequency RSAE and resonant kinetic ballooning mode were confirmed in experiments, and found to cause energetic ions losses. Low-n Alfvenic ITG were observed and identified in Ohmic and NBI plasmas. For the first time non-resonant internal kink modes destabilized by energetic electrons with ECRH+ECCD were found in current ramp-up phases.
Speaker: Prof. Xuru Duan (Southwestern Institute of Physics)
• 14:00
Overview of recent physics results from MAST 4h 45m
New results from MAST will be presented that focus on validating models in order to extrapolate to future devices. Particular attention will be given to the areas of scenario development, fast particle physics and plasma exhaust. Understanding filamentary transport across the scrape off layer is a key issue for the design and operation of future devices as it is crucial in determining the power loadings to the divertor and first wall of the machine. A detailed characterisation of the MAST Scrape Off Layer has been performed including results from new diagnostics giving plasma potential and ion temperature measurements. Detailed studies have revealed how filament characteristic are responsible for the broadening of the midplane density profile. These measurements have been compared to extensive modelling, including 3D effects on filaments dynamics with the BOUT++ code, and benchmarking the SOLPS code. Impurity transport studies have shown how the balance between neoclassical and anomalous transport leads to carbon and nitrogen being screened from the core plasma compared to helium which is peaked at the centre. These results, combined with SOLPS modelling, suggest that a stable detachment region can be produced if the impurity puffing is localised. Measurements from a Doppler Backscattering system combined with GS2 simulations have shown that both micro-tearing modes (MTMs) and electron temperature gradient (ETG) modes can be unstable at the top of the pedestal, along with kinetic ballooning modes at the bottom of the steep gradient region. The experimental observations of the relative amplitudes and wavelengths of the density and magnetic field fluctuations at the top of the pedestal are more similar to the linear characteristics of the ETG than the MTM. Comprehensive measurements from a suite of diagnostics on MAST have shown the effect that core MHD modes and resonant magnetic perturbations (RMPs) have on the confinement and redistribution of fast ions arising from neutral beam injection (NBI). Subsequent experiments on MAST demonstrated that by vertically displacing the plasma to achieve off-axis NBI fast ion injection or by changing plasma density or NBI power to vary the fast ion pressure gradient the redistribution could be mitigated.
Speaker: Dr Andrew Kirk (Culham Centre for Fusion Energy)
• 14:00
Overview of simulation results using computation resources in the framework of IFERC-CSC 4h 45m
Following the successful operation of a European High Performance Computer For Fusion applications (HPC-FF) in Jülich, Germany, from 2009 to 2013, a new supercomputer dedicated to magnetic fusion research was procured within the Broader Approach agreement between Europe and Japan. The new platform, “Helios”, was installed in the International Fusion Energy Research Centre - Computational Simulation Centre (IFERC-CSC) in Rokkasho, Japan and it started operations in January 2012 and is expected to serve until the end of 2016. The use of the Helios computer has been rather successful with a large scientific output expressed in the number of peer-reviewed publications of around 1 per project per year. In the 5th Cycle of operation of Helios, over 120 projects have been selected, corresponding to over 300 users. In this paper, the main scientific and technical results obtained in the Helios numerical simulations projects are described with emphasis on the impact in developing fusion science and related technologies. At the end of its life cycle the use of Helios will be replaced in Europe with a new EUROfusion supercomputer, allowing further development of fusion technologies based on computer modelling and simulations. Collaborations between Japan and Europe will continue with new opportunities for joint projects like Helios.
Speaker: Dr Duarte Borba (EUROfusion Programme Management Unit, Culham Science Center, Abingdon, UK)
• 14:00
Overview of Spherical Tokamak Research in Japan 4h 45m
Nationally coordinated research on spherical tokamak (ST) is being conducted in Japan, to strengthen the scientific basis and to broaden future options of ST applications. The research themes to concentrate on are (1) the physics of very high beta plasmas, (2) development of start-up, current drive, and control techniques without the use of the central solenoid (CS), and (3) demonstration of very long pulse operation and the study of steady-state issues. Research elements are developed on several devices optimized for each objective. The basic mechanism of tokamak plasma formation by ECW/EBW was investigated on LATE. The tokamak configuration with closed flux surfaces is formed spontaneously when the equilibrium current changes from the vertical charge separation current to the toroidal return current. Highly over-dense plasmas have been produced, indicating mode conversion to EBW. A maximum plasma current of 66 kA was achieved using 28 GHz on QUEST. Plasma current start-up by LHW is being investigated on TST-2. The most efficient ramp-up was achieved by the capacitively-coupled combline antenna, which excites a traveling LHW with a sharp wavenumber spectrum and high directionality. Experiments with a top-launch antenna, expected to improve single-pass absorption and increase current drive efficiency, have started. The formation of closed flux surfaces by transient coaxial helicity injection (CHI) was verified by internal magnetic probe measurements on HIST. A stable closed flux formation was achieved by high bias flux operation, and the validity of helicity balance was confirmed. CHI electrodes were installed on QUEST under US-Japan collaboration. An RF-driven long pulse discharge of up to 810 s has been achieved on QUEST. Operation with hot metal wall has started, with the aim to control particle recycling by active wall temperature control. Compact toroid injection is being developed as an advanced fueling method. High-power reconnection heating of ST plasmas using axial merging of two ST plasmas was demonstrated in TS-3, TS-4 and UTST. Collaboration on the MAST device has demonstrated that reconnection heating can be extended successfully to larger scale and higher magnetic field. ST plasma stability improvement was accomplished by applying a helical field in TOKASTAR-2, an ST-helical hybrid device equipped with helical field coils.
Speaker: Prof. Yuichi Takase (University of Tokyo)
• 14:00
Overview of SST-1 Up-gradation & Recent Experiments in SST-1 4h 45m
Steady State Superconducting Tokamak (SST-1) is a operational’ experimental superconducting device since late 2013. Since last IAEA-FEC; SST-1 has been upgraded with Plasma Facing Components being installed and integrated in the vacuum vessel and is getting prepared towards long pulse operations in both circular and elongated configurations. The PFC integration has been completed in August 2015 and initial experiments have begun in SST-1 with circular plasma configurations. SST-1 offers a unique possibility of investigating long pulse discharges with large aspect ratio (> 5.5) compared to contemporary devices. Presently, SST-1 standard ohmic discharges are in excess of 100 KA with typical core density ~ 2 × 1019 m-3 and core electron temperatures ~ 500 eV having duration in excess of 300 ms. A 42 GHz ECR pre-ionization source at ~ 150 KW in 1.5 T central field breaks down the gas, the current starts up at ~ 1.3 MA/s in 60-80 ms in an induced field of ~ 0.3 V/m. These standard discharges demonstrate copious saw teething and MHD activities as the pulse progresses including NTM, mode locking and MHD characteristics. PFC equipped SST-1 has completed these basic experimental studies confirmed with simulations. These includes eddy currents influencing the NULL dynamics, field errors, equilibrium index evolutions, wall influencing plasma characteristics, plasma positions, plasma rotational and Tearing Mode characteristics including the island width and island growths etc. Presently, SST-1 is attempting at multi second long high aspect ratio plasma discharges by coupling the Lower Hybrid with the Ohmic plasma as well as with robust real time position and density controls. SST-1 device has been upgraded with a pair of internal coil aimed at effective fast plasma control and a pair of segmented coil aimed at controlling some of the rotational aspects of plasma including the RMPs and ELMs. Supersonic Molecular Beam Injection (SMBI) from both high field and low field sides and Pellets Injection Systems have also been added with several edge plasma diagnostics aimed at both density control and edge plasma turbulence studies. The up-gradation details including the planned ones, salient early plasma characteristics in large aspect ratio PFC equipped SST-1 plasma and future experimental plans towards long pulse operations in SST-1 will be elaborated in this paper.
Speaker: Dr Subrata Pradhan (Institute for Plasma Research)
• 14:00
Overview of the FTU results 4h 45m
Experiments of Electron Cyclotron (EC) assisted breakdown have shown the presence of runaway electrons (RE) also below the Dreicer electric field threshold, indicating that the RF power acts as seeding for fast electrons, and a large database of post-disruption generated RE beams has been analysed in order to identify linear dynamical models for new position and current RE beam controllers. A linear micro-stability analysis of Neon doped pulses has been carried out to investigate the mechanisms leading to the observed density peaking. A study of the ExB drift effect on the MARFE instability has been performed and the peaking of density profile has been well reproduced using a particle pinch term of the form D_T n d(lnT)/dr. The 2/1 tearing mode (TM) observed in high density plasmas has shown a final phase characterized by limit cycles on the amplitude/frequency plane. The analysis of the linear stability has highlighted a destabilization with increasing peaking of the current profile during the density ramp-up, and experiments of real-time control of such a TM, by means of EC heating of the magnetic islands, have shown a considerable stabilizing effect. A Cooled Lithium Limiter with thermal load capability up to 10 MW/m^2 has been tested. The pulse duration has been extended up to 4.5 s and elongated configurations have been obtained for 3.5 s, with the X-point just outside the plasma chamber. W/Fe samples have been exposed in the SOL in order to study the sputtering of Fe and the W enrichment of the surface layer. Dust has been collected and analyzed, showing that the metallic population exhibits a high fraction of magnetic grains. A new diagnostic for in-flight RE studies has allowed to provide simultaneously the image and the visible/infrared spectrum of the forward and backward radiation. A fast infrared camera for thermo-graphic analysis has provided the pattern of the toroidal limiter heating by disruption heat loads, and a triple-GEM detector has been mounted on one equatorial port for soft-X rays diagnostic. The Collective Thomson Scattering diagnostic has been upgraded and used for investigations on Parametric Decay Instability excitation by EC beams correlated with magnetic islands, and new capabilities of the Cherenkov probe have been explored in the presence of Beta-induced Alfvén Eigenmodes associated to high-amplitude magnetic islands.
Speaker: Dr Gianluca Pucella (ENEA)
• 14:00
Overview of the IFMIF/EVEDA Project 4h 45m
IFMIF, the International Fusion Materials Irradiation Facility, presently in its Engineering Validation and Engineering Design Activities (EVEDA) phase under the Broader Approach Agreement, will allow accelerated testing of structural materials with fusion relevant neutrons at >20dpa/year in 500cm3. IFMIF consists of two 125 mA and 40 MeV D+ linear accelerators operating in CW mode. The concurrent beam lines impact on a liquid lithium target with a 200mm  50mm beam cross section. The target consists of a 25mm 1mm thick liquid lithium screen flowing at 15m/s and 250°C channelled by a R250mm concave RAFM backplate. The suitable neutron flux generated in the forward direction will irradiate 12 test capsules housing around 1000 small specimens independently cooled with helium gas. The Engineering Design Activity (EDA) phase of IFMIF was successfully accomplished within the allocated time. The Engineering Validation Activity (EVA) phase has focused on validating the Accelerator Facility, the Target Facility and the Test Facility with the construction of various prototypes. The ELTL has successfully demonstrated the long term stability of a lithium flow under IFMIF nominal operational conditions with 25 days continuous operation in Oarai (JAEA) at 250°C and 15m/s within 1mm free surface fluctuations. A full-scale prototype of the High Flux Test Module has been successfully tested in the HELOKA loop (KIT Karlsruhe) demonstrating the feasibility of the uniformity in the temperature selected for the specimen set irradiated in each capsule. LIPAc, presently under installation and commissioning, will validate the concept of IFMIF Accelerators with a D+ beam of 125mA and 9MeV. The commissioning of the H+/D+ beams in Rokkasho Fusion Institute at 100keV was concluded early 2016; the commissioning of the 5MeV beam is to follow till early 2017. The 9MeV D+ beam will be achieved with a superconducting cryomodule during 2018. The realisation of a fusion relevant neutron source is a necessary step for the successful development of fusion. The stable progress achieved in this final EVEDA phase has ruled out technical concerns and potential showstoppers raised in the past. In the light of costs, which are unquestionably marginal to those of a fusion plant, a situation has emerged where soon steps towards constructing a Li(d,xn) fusion relevant neutron source could be taken.
Speaker: Dr juan knaster (IFMIF/EVEDA)
• 14:00
Overview of the JET results in support to ITER 4h 45m
The JET contributors Europe has elaborated a Roadmap to the realisation of fusion energy in which ‘ITER is the key facility and its success is the most important overarching objective of the programme’. We review the contribution of the recent JET experiments with the ITER first wall materials mix, and, the underlying physics understanding to mitigate the scientific risks identified in the ITER research plan. Indeed, together with the ITER scenario development, a strong focus on JET is pursued for addressing ITER needs and developing a sound physics basis for the extrapolation through first principle and integrated modelling: plasma wall interaction, disruption mitigation (installation of a third mitigation valve), H mode access, W-control with higher electron heating (ICRH ITER-like antenna re-instated), pellet ELMs pacing with the optimised vertical high field side track. The JET ITER-Like Wall experiment provides an insight in the coupling between tokamak-plasma operation and plasma-surface interaction in the unique Be/W material environment and acts as test-bed to verify models and modelling tools for ITER. Disruptions are considered as the highest programmatic risk in the ITER Research Plan and experimental and modelling effort in Europe and JET are reviewed. High spatial resolution Doppler backscattering measurements have revealed novel insights into the development of the edge transport barrier. The operational constraints of a metal wall can prevent reaching plasma energy confinement required for QD-T=10 on ITER. Progress on JET to mitigate this risk is reported aiming at maximizing the core and pedestal performance in stationary condition with the W divertor constrain. The measured D-D neutron fluence and gamma dose rates have been successfully compared with simulations performed with the codes used for ITER nuclear safety analyses. Finally, the benefit to further use JET beyond 2020 to train the international ITER team with an upgrade tungsten divertor and with the ITER control tools will be discussed. This work has been carried out within the framework of the EUROfusion Consortium and has received funding from the Euratom research and training programme 2014-2018 under grant agreement No 633053. The views and opinions expressed herein do not necessarily reflect those of the European Commission
Speaker: Dr xavier Litaudon (EUROfusion)
• 14:00
Overview of the KSTAR Research in Support of ITER and DEMO 4h 45m
The KSTAR device has been operated since the first plasma in 2008 with the mission of exploring the physics and technologies of high performance steady-state operation that are essential for ITER and fusion reactor. KSTAR has been focusing on maximizing performance and extending pulse length targeting H-mode dischage up to 300s at higher plasma current up to 2 MA, and at higher normalized beta (βN) up to ~5. In the 2015 campaign, various long pulse H-mode discharges have been operated after the improved plasma shape control, and a longest H-mode discharge was achieved up to 55 s at 0.6 MA in plasma current and 2.9 T in toroidal field utilizing 4.2 MW neutral beam and 0.65MW ECCD systems. This will be the longest H-mode discharge in tokamak devices. Fully non-inductive discharges have been carried out at the reduced plasma current of 0.4 MA to access the steady-state operation condition. The first fully non-inductive operation was achieved with relatively high plasma performances (βN ~ 2.1 and βP ~ 3.0). However the shot was terminated at about 16s due to the excessive heat in poloidal limiter. KSTAR device could be an ideal device to investigate the basic of the stability limits and confinement improvement utilizing unique features of KSTAR such as extremely low error field, versatile in-vessel control coils (IVCC), and advanced 2D/3D imaging diagnostics. In this paper, the progress of the KSTAR research to support ITER and DEMO will be reported
Speaker: Dr Yeong-Kook Oh (National Fusion Research Institute)
• 14:00
Overview of the Present Progresses and Activities on the Chinese Fusion Engineering Test Reactor 4h 45m
The Chinese Fusion Engineering Testing Reactor (CFETR) is the next device for the Chinese magnetic confinement fusion (MCF) program which aims to bridge the gaps between the fusion experiment ITER and the demonstration reactor DEMO. CFETR will be operated in two phases: Steady-state operation and tritium self-sustainment will be the two key issues for the first phase with a modest fusion power up to 200 MW. The second phase aims for DEMO validation with a fusion power over 1 GW. Advanced H-mode physics, high magnetic fields up to 7T, high frequency electron cyclotron resonance heating (230 GHz) & lower hybrid current drive (7.5GHz) together with off-axis negative-ion neutral beam injection will be used for achieving steady-state advanced operation. The detailed design, research and development activities including integrated modeling, R&D on high field magnet, material, T plant, remote handling, physical validation on EAST tokamak to demonstrate feasibility of high performance steady state operation, and future MCF road map will be introduced in this paper.
Speaker: Prof. Yuanxi WAN (Institute of Plasama Physics, Chinese Academy of Sciences)
• 14:00
Overview of the RFX-mod fusion science activity 4h 45m
Thanks to its flexibility and unique control capability, the RFX-mod device has been operated in the last two years to investigate a wide range of experimental conditions. Reversed-Field Pinch (RFP), Tokamak and the full range of magnetic configurations in between the two, the ultra-low q, have been produced to contribute to physics common topics highlighting similarities and/or peculiarities. The experiments have been inspired and complemented by an intense theoretical modeling activity, based on 3D nonlinear visco-resistive MHD, advanced non-local transport simulation, Hamiltonian guiding center and non-linear gyrokinetic codes. The RFX-mod scientific program thus provides contribution to magnetically confined plasma physics on various fundamental aspects: 3D effects, transport barriers and MHD control. The effect of spontaneous or externally induced 3D (helical) equilibria on high current RFP plasmas has been deeply investigated, with particular emphasis on the role of the isotopic effect. An enhancement of confinement in deuterium plasmas has been observed and reproduced by simulations. The role of 3D effects on transport and small scale turbulence in the presence of magnetic islands has been studied, as well as the role of a helical boundary in the formation of 3D core equilibria, relevant for the dynamo effect in hybrid regimes in Tokamaks. Physics issues associated to the density limit phenomenon have been addressed in all magnetic configurations. The analysis of the locking-unlocking threshold for the spontaneous rotation of the tearing in the RFP has shown the absence of hysteresis in the presence of feedback control. The results are well reproduced by a code, reliable for Tokamak plasmas as well in the investigation of the feedback control of the (2,1) mode, successfully experimented in RFX-mod. The application of 3D perturbations has demonstrated to be effective in deconfining runaway electrons in Tokamak plasmas. Long lasting H-modes have been attained in both circular and shaped (single null) plasmas thanks to the exploitation of an edge polarized electrode. Indications of the first L-H transitions in q(a)<2 circular plasmas have been obtained. In order to enhance the confinement properties in RFP and to extend the operational scenarios both in RFP and Tokamak, a series of modifications for the RFX-mod device has been proposed.
Speaker: Dr Matteo Zuin (Consorzio RFX, Padova, Italy)
• 14:00
Overview of the TCV Tokamak Program: Scientific Progress and Facility Upgrades 4h 45m
A broad upgrade program is underway at the TCV tokamak. A historic first step is the present commissioning of the first neutral beam injector (NBI), delivering 1 MW of power at energies in the 15-30 keV range. Four gyrotrons are also being added in 2016-2018 to bring the total ECRH power to 6 MW. A second, counter-injected, 1-MW neutral beam is also planned, in addition to the introduction of variable-configuration divertor baffles, to expand the role of TCV in preparing the grounds for both ITER and DEMO. TCV is now operating partly as a European Medium-Size Tokamak (MST) facility within the EUROfusion consortium. In the realm of edge and exhaust physics, access to divertor detachment has been investigated through density ramps and nitrogen puffing, first revisiting the conventional single-null divertor, then proceeding through the variety of alternative geometrical divertor configurations that TCV can sustain, including the X-divertor, X-point target, and the snowflake minus and plus. Studies of scrape-off layer (SOL) transport are focusing especially on the enhanced convection that leads to profile broadening at high density and is generally attributed to intermittent filamentary structures; through a large scan in parallel connection length, no evidence for far SOL profile broadening is found. The double heat-flux scale length measured in limited L-modes in the TCV SOL, a possible concern for reactor start-up, has been reproduced by the 3D turbulent-transport GBS code. A wall cleaning solution based on ECRH-sustained, current-less helium discharges was recently tested on TCV for JT-60SA. In the area of control, a generalized plasma shape and position controller, based on real-time, sub-ms equilibrium reconstruction was recently tested successfully. Considerable attention has been given to disruptions. In addition to exploring techniques for disruption mitigation or avoidance (by massive gas injection or ECCD, and with assistance from real-time modeling), the related problems of runaway electron generation, mitigation, and control, are also being tackled. Investigations of the density disruption limit are ongoing, in particular to explore its dependence on gas puffing and plasma shape. The possible “seedless” excitation of NTMs, mediated by neoclassical toroidal viscosity, has been successfully studied by modifying the rotation profile with ECRH.
Speaker: Dr Stefano Coda (CRPP-EPFL)
• 14:00
Progress in ITER Construction, Manufacturing and R&D 4h 45m
The ITER project is a critical step in the development of fusion energy: its role is to confirm the feasibility of exploiting magnetic confinement fusion for the production of energy for peaceful purposes by providing an integrated demonstration of the physics and technology required for a fusion power plant. Rapid progress is being made in the design, manufacturing, construction and R&D activities, and the facility is now taking shape at St-Paul-lez-Durance. Supported by impressive achievements in fusion technology R&D, manufacturing of ITER components is in full swing. The international collaboration formed around the production of superconducting magnets for ITER has produced over 600 t of Nb3Sn and almost 250 t of NbTi superconducting strand. 80% of the superconductors required for the ITER magnets are complete, and coil fabrication activities are underway in 6 of the 7 partners’ factories. Fabrication of the vacuum vessel is moving forward, with structures being manufactured under the responsibility of four contributing Domestic Agencies, manufacturing of the thermal shield is also in progress, and the first elements of the cryostat (~29 m diameter × ~29 m height) have been delivered to the ITER site. Substantial progress has also been achieved in prototyping and R&D activities in areas such as plasma facing components, in-vessel coils, H&CD systems, remote handling and power supplies in preparation for manufacturing. A wide-ranging physics R&D programme, closely integrated with the ITPA and the major fusion facilities in the ITER Members, is also addressing key issues impacting on finalization of the ITER design and preparations for operation. These R&D activities encompass studies of disruption mitigation, analysis of ELM control by magnetic perturbations, characterization of heat loads in stationary and transient plasma phases, plasma-wall interactions with all-metal PFCs, and studies of plasma scenarios for non-active and nuclear phases of the ITER experimental programme. The paper will review the progress made in developing the advanced technologies required for ITER and in the manufacturing activities for major tokamak components, discuss advances made in experimental and modelling studies of key physics issues, detail measures taken to establish a more effective project organization and present the status of construction of the ITER facility.
Speaker: Mr Bernard Bigot
• 14:00
Progress of the Recent Experimental Research on the J-TEXT Tokamak 4h 45m
The progress of experimental research over last two years on the J-TEXT tokamak is reviewed, the most significant results including: the investigation of the effect of resonant magnetic perturbations (RMPs) on the J-TEXT operation region, impurity transport and confinement, and runaway electrons suppression; study of the threshold for runaway current generation; and identification of the quasi-coherent characteristics in spectra of density fluctuations. The effect of RMPs on the J-TEXT Ohmically heated operation region is studied on J-TEXT by applying RMPs in high density limit and low-q limit discharges. It is found that moderate amplitude of applied RMPs either increases the density limit from less than 0.7nGW to 0.85nGW or lowers edge safety factor qa from 2.15 to nearly 2.0. As a result, the disruption precursor is suppressed and the disruption is delayed by about 30-150 ms. The influence of RMPs on impurity behavior is also studied by applied RMP with m/n = 2/1 dominant component. It is found that the CV decay time after methane injection decreases as the RMP amplitude increases. When the RMP penetration occurs, the emission of CIII (464.7nm) from the edge region develops a gradually increasing asymmetry. Stronger emission occurs at the high-field-side (HFS) edge. The potential suppression of runaway electrons by RMP is also investigated on J-TEXT. The experimental result indicates that the magnetic perturbation enhanced the runaway loss rate by the formation magnetic islands rather than by the magnetic perturbation itself. Both the amplitude and the length of runaway current can be reduced by applying the RMP before the disruption. Regarding the threshold for runaway current generation, it is found that the key parameter affecting runaway generation is the edge safety factor qa, not BT, on J-TEXT. The threshold of qa decreases with increasing BT. The electrostatic turbulence exhibited quasi-coherent characteristics in the spectra of density fluctuations observed in the J-TEXT ohmic confinement regime. These quasi-coherent modes (QCMs) are detectable in a large plasma region (r/a ~ 0.3 − 0.8) with frequency of 30 – 140 KHz. The mode rotates in the electron diamagnetic direction. The combined experimental results indicate that the QCMs survive in the linear Ohmic confinement regime of the plasma, where the TEM is predicted to be unstable.
Speaker: Mr Ge Zhuang (Huazhong University of Science and Technology)
• 14:00
Recent Progress of JT-60SA Project 4h 45m
The JT-60SA project has been promoted since June 2007 under the framework of the Broader Approach (BA) agreement and Japanese national fusion programme for an early realization of fusion energy by conducting supportive and complementary work for the ITER project and directing DEMO design activity. With the powerful and varied deposition profile of heating and current drive system, flexible plasma shaping capability and various kinds of in-vessel coils to supress MHD instabilities, JT-60SA is sure to play an essential role to address essential issues to achieve long sustainment of high beta_N burning plasmas expected in DEMO. Components and systems of JT-60SA are procured by the implementing agencies (IAs): Fusion for Energy in EU and JAEA in Japan. Their design, fabrication, installation and commissioning have been actively directed and supervised by the IAs. As of the end of 2015, twenty-seven procurement arrangement (PAs) have been concluded covering 95% of the values of in-kind contribution for JT-60SA. In spite of the size, components of JT-60SA have been manufactured well within the tolerance of 1 mm order. EU procures TF coils, most of the power supply systems, cryogenic system, cryostat and so on. The cold test of the first TF coil with a nominal current of 25.7 KA at 4.5-7.0 K was successfully carried out. JA procures EF coils, Central Solenoids, Vacuum Vessel, thermal shields, heating system, diagnostics system and so on. Vacuum Vessel sectors were welded on the cryostat base forming a 340° torus. The heating systems (P-NBI, N-NBI and ECRF) has been conditioned to be operated at their full power (41 MW in total) for 100s. The first plasma of JT-60SA is scheduled in 2019. Wide range of operational region of JT-60SA kept in mind, the JT-60SA research plan (SARP) has been regularly updated on the basis of intensive discussion among European and Japanese researchers. The latest SARP (version 3.3) open to the public in March 2016 shows that wide operational region of JT-60SA covers that of recent European and Japanese DEMO designs. DEMO oriented researches such as study of ECRF assisted startup, investigation of non-inductive current overdrive scenario using TOPICS code were added. This paper summarize the recent progress of JT-60SA Project pushed forward by close collaboration of EU and Japan.
Speaker: Dr Hiroshi SHIRAI (Japan Atomic Energy Agency)
• 14:00
Review of Recent Experiments on the T-10 Tokamak with All Metal Wall 4h 45m
Review of the recent experimental results obtained on the T-10 tokamak is presented. To decrease the level of light impurities in 2015 both the rail and circular limiters were replaced with ones made of tungsten. The used tungsten type «POLEMA» as well as the technology of its soldering to the bronze substrate are similar to those applied for the production of the ITER divertor tiles. In the same time a movable lithium limiter was installed in the upper port. This limiter based on capillary-porous structure was made by JSC “Red Star”. With the tungsten limiter a considerable increase of the core radiation losses was obtained. Results on prevention of tungsten penetration in the core plasma by central ECRH and by insertion of the lithium limiter are presented in the paper. The efficiency of removal of heavier iron impurity depending on the discharge parameters and the power of the central ECRH was investigated. The maximal decrease of the heavier impurities concentrations is 5. Performed using the canonical profiles model analysis of the experiments on the density profile dynamic upon variation of the ECRH power showed that the density profile stiffness rises linearly with the heating power, while the peaking of the pressure profile in the core plasma asymptotically approaches to the canonical value. Using of bispectral analysis applied to the fluctuations of potential, density and poloidal magnetic field measured with heavy ion beam probe diagnostic showed an existence of three-wave interaction between GAM and broad-band turbulence. Also shown that the GAM amplitude declines with the mean density growth. The investigations of density fluctuation characteristics with correlation reflectometry confirmed a considerable decrease of the fluctuation amplitude together with disappearance of the quasicoherent modes on the inner side of the torus. Modeling showed that this effect can be, to a great extent, explained by nonlocality of reflectometry. Experiments with tangential X-ray detector indicated that abrupt restructuring of the low-m MHD modes and inward plasma shift during an energy quench are accompanied by bursts of the fast-scale (~0.5MHz) magnetic fields oscillations. Plasma discharge recovery after an energy quench is demonstrated in the T-10 high density plasma using ECRH auxiliary heating and controllable operation of the plasma current.
Speaker: Mr Dmitrii Sarychev (NRC "Kurchatov Institute")
• 14:00
The Quest For Laboratory Inertial Fusion Ignition in the US 4h 45m
M. J. Edwards1, T. C. Sangster2, D. B. Sinars3 1 Lawrence Livermore National Laboratory, Livermore, CA 94551, USA 2 Laboratory for Laser Energetics, 250 E. River Rd, Rochester, NY 14623-1299, USA 3 Sandia National Laboratories, Albuquerque, NM 87185, USA Ignition and significant fusion yield from Inertial Confinement Fusion (ICF) remains a grand scientific challenge. The ICF community in the US, together with international collaborators is executing a coordinated effort exploring 3 approaches to ignition each with different risks and advantages: laser driven x-ray drive, laser direct drive, and magnetic direct drive. This talk presents the status and future focus of these approaches in the US. X-ray drive is pursued at the National Ignition Facility (NIF). In this approach ~ 1.8 MJ of laser light illuminates a cylindrical gold hohlraum to produce a highly uniform x-ray field to implode a spherical capsule containing DT fuel. The original ignition target design gave fusion yields (~2kJ or ~5x1014 neutrons) far from ignition because of the challenging hydrodynamics associated with the high (~35X) convergence ratio (CR) compounded by laser plasma instabilities (LPI) in the hohlraum introducing strong time dependent drive asymmetry. A more stable, lower CR variation of that design resulted in yields approaching 1016 neutrons (~26kJ) and for the first time demonstrated significant alpha self-heating that roughly doubled the fusion yield. It has become clear that further improvements in performance will require better control of the implosion shape by reducing the LPI that currently prevents the precision drive symmetry needed for ignition as well as improved capsule mounting schemes that perturb the implosion less. In laser direct drive (LDD) the capsule is directly irradiated spherically with laser light. This couples more energy to the fuel than in x-ray drive reducing the capsule convergence ratio needed for ignition to ~ 20 at NIF’s energy. However, the proximity of the laser to the capsule places stringent demands on the laser target coupling uniformity and the allowable levels of LPI. The laser coupling and hydrodynamics of LDD are being refined at the LLE’s Omega laser in hydro-scaled targets with laser imprinting and LPI mitigation being studied in collaboration with NRL.
Speaker: Dr Michael John Edwards (Lawrence Livermore National Laboratory)
• 16:10 16:40
Coffee Break 30m
• 16:40 18:45
ITER Technology: FIP/1
Convener: Mr Jerome Pamela (CEA Cadarache)
• 16:40
Recent Progress of ITER Package in ASIPP 20m
ASIPP has taken the responsibility of most CN ITER package. All packages follow current ITER schedule. The superconducting conductor package consists of 106 conductors with 6 kinds includes 7.5% TF conductor,totalPF conductor from PF2 to PF5, total CC conductor, and MB and CB conductor of feeder. Now total CN TF conductor package has been completed in production,and delivered to IO; completed the production and acceptance test of CC and feeder conductor package; completedthe production and acceptance testof 35 PF conductors and 4 dummy conductors. The ITER feeder system consists of 31 units. They convey power and coolant to magnets, and hold the numerous instrumentation channelswith the functioning of magnets system operationand monitoring. Now ASIPP has completed all qualification work and started manufacturing after PF4 CFT MRA meeting. The ITER Correction Coils (CC) consists of three sets of six coils each. Each pair of coils located on opposite sides with respect to the plasma is series connected with polarity such to produce asymmetric fields. TheCC PA was signed between IO and CN DAin 2010, now ASIPP has developed the manufacturing process including winding process, VPI technology, laser beam welding, helium inlet/outlet welding technology, and production qualification process is still on going. CN power supply package consists of PF AC/DC converter,reactive power compensationandharmonic filter (RPC&HF), and pulsed power electrical network materials (PPEN). ASIPP takes responsibility of key technology R&D, all kind test, integration and technical support. ASIPP has completed AC/DC converter and RPC&HF prototype test and integration test, and started its manufacturing since 2015. Now first PF AC/DC converter unit manufacturing has been completedandqualified by IO, 15 kinds of PPEN equipment has been delivered to IO site. ASIPP has two ITER diagnostic procurements, #12 horizontal port plug and radial X-ray camera(RXC).Port integration group has organized a system integration review meeting in July 2015 with mostly general issues be resolved or analyzed, most model clashes between tenants have been resolved. RXC's structure design is optimized and installation process is studied considering the simplification and easiness of maintenance. Remote handling skills and tools are designed for the system maintenance after being activated.
Speaker: Prof. Peng FU (Institute of Plasma Physics, Chinese Academy of Sciences)
• 17:00
ITER Central Solenoid Module Fabrication 20m
The fabrication of the modules for the ITER Central Solenoid (CS) has started in a dedicated production facility located in Poway, California, USA. The necessary tools have been designed, built, installed and tested in the facility to enable the start of production. The current schedule has first module fabrication completed in 2017, followed by testing and subsequent shipment to ITER. The Central Solenoid is a key component of the ITER tokamak providing the inductive voltage to initiate and sustain the plasma current and to position and shape the plasma. The design of the CS has been a collaborative effort between the US ITER Project Office (US ITER), the international ITER Organization (IO) and General Atomics (GA). GA’s responsibility includes completing the fabrication design, developing and qualifying the fabrication processes and tools, and then completing the fabrication of the seven 110 tonne CS modules. The modules will be shipped separately to the ITER site, stacked and aligned in the Assembly Hall prior to insertion in the core of the ITER tokamak. A dedicated facility in Poway, California, USA has been established by GA to complete the fabrication of the seven modules. Infrastructure improvements included thick reinforced concrete floors, a diesel generator for backup power along with cranes for moving the tooling within the facility. The fabrication process for a single module requires approximately 22 months followed by 5 months of testing, which includes preliminary electrical testing followed by high current (48.5kA) tests at 4.7K. The production of the 7 modules is completed in a parallel fashion through ten process stations. The process stations have been designed and built with most stations having completed testing and qualification for carrying out the required fabrication processes. The final qualification step for each process station is achieved by the successful production of a prototype coil. Fabrication of the first ITER module is in progress. The seven modules will be individually shipped to Cadarache upon their completion. This paper describes the processes and status of the fabrication of the CS Modules for ITER.
Speaker: Mr John Smith (General Atomics)
• 17:20
Long-pulse acceleration of 1MeV negative ion beams toward ITER and JT-60SA neutral beam injectors & Towards powerful negative ion beams at the test facility ELISE for the ITER and DEMO NBI system 20m
A. In order to realize the negative-ion-based neutral beam (NB) systems for ITER and JT-60SA, development of the Multi-Aperture and Multi-Grid (MAMuG) electrostatic accelerator is one of common critical issues. For these NB injectors, 5- and 3-stage MAMuG accelerators are being developed to achieve the acceleration of negative ion beams up to 1 MeV, 40 A (200 A/m^2) for 3600 s and 0.5 MeV, 22 A (130 A/m^2) for 100 s, respectively. However, there were no experiments of long-pulse MeV-class beam acceleration. Though JAEA achieved the rated beam energy of 1 MeV, the pulse duration was limited to be less than 1 s [1] due to a low voltage holding capability and high grid power loads. After the last FEC conference, following issues were investigated such as multi-grid effect on the voltage holding capability and reduction of the grid power loads. New accelerators have been designed to realize stable voltage holding by taking into account the multi-grid effect on voltage holding capability, which satisfies the requirement of beam energy for ITER and JT-60SA with 5-stage and 3-stage, respectively. The grid power load has been suppressed less than a half of the design values of the accelerators by modifying the geometry of the extractor and the acceleration grids to suppress generation of secondary electrons. By applying the developed techniques based on the R&D results, the hydrogen negative ion beams of 0.97 MeV, 190 A/m^2 have been successfully accelerated up to 60 s from the ITER prototype accelerator. The pulse duration of such high power density negative ion beams (~184 MW/m^2) has been extended from 0.4 to 60 s, which is the longest pulse length in the world. There is no limitation to extend the pulse duration, since no degradation of the voltage holding has been observed during the long-pulse operations neither by cesium accumulation nor by thermal damage of the acceleration grids. This achievement is one of breakthroughs toward the realization of the high-energy NB systems. [1] A. Kojima, et al., Nucl. Fusion 55 (2015) 063006. B. The negative ion source test facility ELISE represents an important step in the European R&D roadmap towards the neutral beam injection (NBI) systems at ITER. ELISE provides early experience with operation of large RF-driven negative hydrogen ion sources. Its source area is 1x0.9 m2 and the net extraction area of 0.1 m2, formed by 640 apertures, corresponds to a half-size ITER source. The test facility aims at demonstrating large-scale extraction and acceleration of negative hydrogen ions (H‾, D‾) for pulses of up to 1 h with half the current required on ITER. Additionally, the ratio of co-extracted electrons to ions must be kept below one, which is quite demanding in particular for deuterium operation. Starting with first plasma pulses in March 2013, ELISE has meanwhile demonstrated stable 1 h plasma discharges in hydrogen with repetitive 10 s extraction every 3 min with 9.3 A extracted current and an electron-to-ion ratio of 0.4 at the pressure required by ITER of 0.3 Pa but using only one quarter of the available RF power. At half of the available RF power a stable 400 s plasma discharge was achieved with 18.3 A beam pulses at an electron-to-ion ratio of 0.7. Linear scaling towards full RF power predicts that the target value of the negative ion current can be achieved or even exceeded. Issues in long pulse operation are the caesium dynamics and the stability of the co-extracted electron current. Newly developed magnetic filter field configurations allowed achieving for the first time 1 h pulses in deuterium with an electron-to-ion ratio below one, however only at a quarter of the available RF power. Advanced beam diagnostics such as beam emission spectroscopy and a sophisticated diagnostic calorimeter reveal that the requirement on the uniformity of these large beams (deviations < 10%) can be met. For a DEMO fusion reactor, the requirements of a heating and current drive system will strongly depend on the DEMO scenario and are presently assessed within EUROfusion WPHCD. As NBI systems based on negative ions are regarded as one candidate, ELISE could serve in a later stage as a test bed for concepts concerning RF efficiency, operation without caesium or with largely reduced caesium consumption, and neutralization by a laser neutralizer in order to improve efficiency and reliability. IPP’s present small scale experiments show promising results.
Speaker: Dr JUNICHI HIRATSUKA (National Institutes for Quantum and Radiological Science and Technology)
• 17:40
Progress of Experimental Study on Negative Hydrogen Ion Production and Extraction 20m
Development of the high performance negative hydrogen ion source is a fundamental demand in realizing fusion reactor. In order to clarify the extraction mechanism of H^-, temporal and spatial variations of the negative ions and electrons in the extraction region are intensively surveyed at NIFS. In addition, the beam acceleration experiments have been performed by changing the accelerator configuration in order to improve the voltage holding capability and to study the negative ion beam optics. In a cesiated hydrogen plasma, it was observed that the negative hydrogen ion density (n_H-) becomes one order higher in magnitude than the electron density (n_e) in the vicinity of the plasma grid (PG). The response of negative-ion rich plasmas to the extraction field was investigated by measuring the plasma potential (V_p) profiles in the axis perpendicular to the PG before and during beam extraction. The V_p increases with applying the extraction field, and the influence of the extraction field on the V_p was observed at 30 mm from the PG. As for the n_H-, it was also observed that the extraction field affects on the n_H- at 30 mm from the PG, where the n_H- decreases simultaneously with the beam extraction. These observations indicate that the extraction field affects the particle dynamics in the wide region extending over 30 mm from the PG. This feature is completely different from that of electron-ion plasmas. We also found that the negative ion production efficiency becomes twice higher by changing the shape of grounded grid holes without any modification on the plasma chamber. We assumed that the behavior of the back-streaming ion was affected by changing the GG. The back-streaming ion trajectory was analyzed with the beam trajectory simulation, and it was found that the back-streaming ion distributes in the larger area on the back plate with the slot GG. This implies that some part of condensed Cs on the back plate was more effectively evaporated by the back-streaming ion with the slot GG and that Cs flux onto the PG surface increased, which resulted in the enhancement of the negative ion production. This result suggests that the accelerator configuration is one of the key factors to determine the negative ion production efficiency and that the Cs consumption can be reduced by the Cs recycling from the wall of the ion source chamber.
Speaker: Dr Masashi Kisaki (JpNIFS)
• 18:00
Progress in High Power Test of R&D Source for ITER ICRF system 20m
The IC H&CD system is one of the major tools for achieving the plasma performances foreseen in ITER's operation scenario. This system is designed to provide 20 MW into the plasma, at frequencies included in the band 40 MHz to 55 MHz. For ensuring 20 MW power availability for plasma operation, 24 MW is required at the output of the RF sources. India is responsible to deliver nine numbers of RF Sources to ITER system. Each source shall have the power handling capability of 2.5 MW/CW at VSWR 2:1 in the frequency range 35 – 65 MHz or 3.0 MW/CW at VSWR 1.5:1 in the frequency range 40 – 55 MHz, along with other stringent requirement. An urgent need for pre-qualification of final stage tube and few critical components is established through R&D program to bridge the significant gap between demonstrated capability of RF source system at various worldwide fusion facilities vs. ITER need, in terms of power level, pulse duration & bandwidth (BW) requirement. In 2012, ITER-India has signed contract with Thales Electron Devices (France) for establishing the technology in very high power RF amplifiers, using Diacrode tube. The contract is to design & develop driver and final stage amplifiers. Tubes and cavities are integrated in full amplifier chain developed by ITER-India. To test the performance of the amplifier chains at matched and mis-matched load condition, high power Test Rig (3MW/CW capability) is developed at Indian test facility. After successful assembly & integration of RF amplifier at Indian test facility, high power RF tests initiated. The objective for the tests is to confirm 1.5 MW output at 35 - 65 MHz for 2000 seconds with 1MHz BW (at 1dB point) over central frequency and to check the reliability of both the tube and the amplifier with a mismatched load (up to VSWR 2:1) which simulates power transmission to an antenna coupled to the plasma. This paper reports successful commissioning of RF amplifier and the achievement of 1.5 MW of RF power for more than 2000s, confirming other extremely challenging specifications and describes the operating scenarios, dissipation limit, safety system and various infrastructures developed at Indian test facility to support such operation.
Speaker: Mrs Aparajita Mukherjee (Institute for Plasma Research, India)
• 18:20
New Results of Development of Gyrotrons for Plasma Fusion Installations & Development of Multi-Frequency Mega-Watt Gyrotrons for Fusion Devices in JAEA & Development of Over MW Gyrotrons for Fusion at Frequencis from 14 GHz to Sub-terahertz 20m
A. 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 seconds. To provide operation with indicated parameters the gyrotrons have very large transverse cavity sizes, output barrier windows made of CVD diamond discs, effective collectors with particle energy recovery. In ITER installation there will be 24 gyrotron systems with 1 MW power each. Russian contribution consists of 8 gyrotron systems. ITER requirements are: frequency 170 GHz, 1 MW power, 1000 s pulse duration, high efficiency of the gyrotrons over 50%, possibility of power modulation with frequency up to 5 kHz, compatibility of the gyrotron complex with ITER control system. In May, 2015 a Prototype of ITER Gyrotron System was completed and its operation was demonstrated. The system consists of gyrotron oscillator, liquid-free superconducting magnet, supplementary magnets, several electric power supplies, cooling systems control and protection systems, and other auxiliary units. The tests were performed in presents of ITER IO and ITER RF DA representatives. In October, 2015 Final Design Procedure for the gyrotron system was successfully passed. High-level parameters were also achieved with long-pulse 140 GHz gyrotrons developed for EAST and KSTAR installations. Significant results were shown on the way to 1.5-2MW, CW gyrotrons. The development of higher frequency (230-700 GHz) gyrotrons for future plasma installations and for plasma diagnostics began. Novel ideas were proposed to enhance gyrotron operation. B. Mega-watt gyrotrons with frequency tuning have become essential devices in fusion science to perform effective EC H&CD. JAEA is developing two types of multi-frequency gyrotrons equipped with a triode magnetron injection gun for ITER and JT-60SA. A TE31,11 mode, which is a candidate mode for 170 GHz oscillation, has sufficient margin for cavity heat-load in 1 MW operation, and it has a great advantage for multi-frequency oscillation. In the JT-60SA project, EC H&CD by second harmonic EC waves are planned using nine sets of 110 GHz/138 GHz dual-frequency gyrotrons to broaden the experimental research area. In FEC2014, demonstrations of 1 MW oscillations for 2 s at 170 GHz/137 GHz/104 GHz with the ITER gyrotron and achievement of 1 MW oscillations for 100 s at 110 GHz /138 GHz in the JT-60SA gyrotron were reported as world records. After FEC2014, oscillation methods to improve the efficiency at 170 GHz for ITER requirements and higher frequency oscillation for the demo-class reactor were investigated. For the JT-60SA gyrotron, the operation area was expanded to surpass maximum performance (1.5 MW/4 s) of the previous JT-60 110 GHz gyrotron. TE31,11 mode oscillations were often prevented by adjacent counter-rotating (ctr-) modes such as TE29,12, and TE28,12 modes. By introducing active anode-voltage control and beam-radius control to suppress adjacent counter-rotating modes, start-up of TE31,11 mode becomes stable and the overall efficiencies achieved ~ 50 % up to 1.1 MW. In looking ahead to a future gyrotron for the demo-class reactor, 203 GHz oscillation of higher-order volume mode (TE37,13) was performed for the first time by taking advantage of the multi-frequency gyrotron feature. In preliminary testing at 203 GHz, 0.9 MW for 0.3 ms and 0.42 MW for 5 s were demonstrated. ITER gyrotron having mega-watt-class power at four frequencies in wide range over 100 GHz was developed. High power gyrotron development toward 1.5 – 2 MW oscillation for several seconds has been carried for further extension of the experiment regime of high performance plasma in JT-60SA. In a test conducted in 2015, achievements of 1.8 MW/1.2 s at 110 GHz (TE22,8 mode) in non-coaxial type gyrotron and high-power oscillation of 1.3 MW/1.3 sat 138 GHz (TE27,10 mode)and 1 MW/1 s of 82 GHz (TE17,6 mode) have been demonstrated as a new world record. C. Megawatt (MW) gyrotrons with a wide frequency range from 14 to 300 GHz are being developed for the collaborative Electron Cyclotron Heating (ECH) study of advanced fusion devices and DEMO reactor. (1) In the first experiment of 300 GHz gyrotron, an output power of over 0.5 MW with TE32,18 single-mode was achieved with a pulse width of 2 ms. This is the first report of MW level oscillation with the DEMO-relevant ECH gyrotron mode. It was also found that the reflection at the output window affects the oscillation mode determination. (2) A new record of the 28 GHz gyrotron output of 1.38 MW was obtained. The fabrication of a newly designed tube aimed at a dual-frequency output power of 2 MW at 28 GHz (0.4 MW CW) and 1 MW at 35 GHz has begun, with all components ready for assembly. Before installing a double-disk window in the dual-frequency gyrotron, we confirmed the dependence of reflective power on the coolant thickness including the reflective power less than 2 % by the cold test using a Gunn diode power of 1 W and the hot test using the gyrotron output power of 600 kW. (3) Based on the successful results of 77 and 154 GHz LHD tubes, the new design of a 154/116 GHz dual-frequency gyrotron with output of over 1.5 MW has been presented.
Speaker: Prof. Grigory Denisov (Institute of Applied Physics Ruissian Academy of Sciences)
• 19:00 21:00
Welcome Reception
• Tuesday, 18 October
• 08:30 10:15
Overview 3: Inertial & Magnetic Fusion: OV/3
Convener: Richard Kamendje (IAEA)
• 08:30
First plasma operation of Wendelstein 7-X 25m
The main objective of the optimized stellarator Wendelstein 7-X (W7-X) is the demonstration of steady-state plasma operation at fusion relevant plasma parameters thereby verifying that the stellarator is a viable fusion power plant concept. The design of W7-X is based on an elaborate optimization procedure to overcome the shortcomings of the concept. After completing the main construction phase of W7-X and successfully commissioning the device, first plasma operation started in December 2015. Plasma operation of W7-X follows a staged approach according to the successive completion of the in-vessel components. During the first operational phase five inboard limiters defined the last closed flux surface. Subsequently, W7-X will be equipped with a test divertor unit and eventually with a steady-state capable high heat flux divertor including active water cooling of all in-vessel components. Integral commissioning of plasma start-up and operation using an electron cyclotron resonance heating (ECRH) and an extensive set of plasma diagnostics, and initial physics studies during the first operational campaign have been successfully completed. Both in helium and hydrogen, plasma break-down was easily achieved. Gaining experience with plasma vessel conditioning, discharge lengths could be continuously extended. Eventually, discharges lasted up to 6 sec, reaching an injected energy of 4 MJ which is twice the limit originally agreed for the limiter configuration. At higher powers of 4 MW and central electron densities of 4.5x10^{19} m^{-3}, central temperatures reached values of 7 keV for the electrons and just above 2 keV for the ions. Important physics studies during this first operational phase include the assessment of the heat load distribution over the inboard limiters changing the toroidal phase and amplitude of deliberately applied error fields, impurity injection and confinement experiments including the effect of the rotational transform, and ECRH power deposition and heat pulse propagation experiments. Also a first assessment of the central electron root confinement, 2nd harmonic O-mode ECRH using multi-pass absorption, and the investigation of confinement and stability of discharges with co- and counter current drive (ECCD) have been achieved. This paper will give an overview of the results of the first experimental campaign of W7-X.
Speaker: Prof. Robert Wolf (Max-Planck-Institute for Plasma Physics)
• 08:55
The Quest For Laboratory Inertial Fusion Ignition in the US 25m
Ignition and significant fusion yield from Inertial Confinement Fusion (ICF) remains a grand scientific challenge. The ICF community in the US, together with international collaborators is executing a coordinated effort exploring 3 approaches to ignition each with different risks and advantages: laser driven x-ray drive, laser direct drive, and magnetic direct drive. This talk presents the status and future focus of these approaches in the US. X-ray drive is pursued at the National Ignition Facility (NIF). In this approach ~ 1.8 MJ of laser light illuminates a cylindrical gold hohlraum to produce a highly uniform x-ray field to implode a spherical capsule containing DT fuel. The original ignition target design gave fusion yields (~2kJ or ~5x1014 neutrons) far from ignition because of the challenging hydrodynamics associated with the high (~35X) convergence ratio (CR) compounded by laser plasma instabilities (LPI) in the hohlraum introducing strong time dependent drive asymmetry. A more stable, lower CR variation of that design resulted in yields approaching 1016 neutrons (~26kJ) and for the first time demonstrated significant alpha self-heating that roughly doubled the fusion yield. It has become clear that further improvements in performance will require better control of the implosion shape by reducing the LPI that currently prevents the precision drive symmetry needed for ignition as well as improved capsule mounting schemes that perturb the implosion less. In laser direct drive (LDD) the capsule is directly irradiated spherically with laser light. This couples more energy to the fuel than in x-ray drive reducing the capsule convergence ratio needed for ignition to ~ 20 at NIF’s energy. However, the proximity of the laser to the capsule places stringent demands on the laser target coupling uniformity and the allowable levels of LPI. The laser coupling and hydrodynamics of LDD are being refined at the LLE’s Omega laser in hydro-scaled targets with laser imprinting and LPI mitigation being studied in collaboration with NRL.
Speaker: Dr Michael John Edwards (Lawrence Livermore National Laboratory)
• 09:20
Recent Progress of JT-60SA Project 25m
The JT-60SA project has been promoted since June 2007 under the framework of the Broader Approach (BA) agreement and Japanese national fusion programme for an early realization of fusion energy by conducting supportive and complementary work for the ITER project and directing DEMO design activity. With the powerful and varied deposition profile of heating and current drive system, flexible plasma shaping capability and various kinds of in-vessel coils to supress MHD instabilities, JT-60SA is sure to play an essential role to address essential issues to achieve long sustainment of high beta_N burning plasmas expected in DEMO. Components and systems of JT-60SA are procured by the implementing agencies (IAs): Fusion for Energy in EU and JAEA in Japan. Their design, fabrication, installation and commissioning have been actively directed and supervised by the IAs. As of the end of 2015, twenty-seven procurement arrangement (PAs) have been concluded covering 95% of the values of in-kind contribution for JT-60SA. In spite of the size, components of JT-60SA have been manufactured well within the tolerance of 1 mm order. EU procures TF coils, most of the power supply systems, cryogenic system, cryostat and so on. The cold test of the first TF coil with a nominal current of 25.7 KA at 4.5-7.0 K was successfully carried out. JA procures EF coils, Central Solenoids, Vacuum Vessel, thermal shields, heating system, diagnostics system and so on. Vacuum Vessel sectors were welded on the cryostat base forming a 340° torus. The heating systems (P-NBI, N-NBI and ECRF) has been conditioned to be operated at their full power (41 MW in total) for 100s. The first plasma of JT-60SA is scheduled in 2019. Wide range of operational region of JT-60SA kept in mind, the JT-60SA research plan (SARP) has been regularly updated on the basis of intensive discussion among European and Japanese researchers. The latest SARP (version 3.3) open to the public in March 2016 shows that wide operational region of JT-60SA covers that of recent European and Japanese DEMO designs. DEMO oriented researches such as study of ECRF assisted startup, investigation of non-inductive current overdrive scenario using TOPICS code were added. This paper summarize the recent progress of JT-60SA Project pushed forward by close collaboration of EU and Japan.
Speaker: Dr Hiroshi SHIRAI (Japan Atomic Energy Agency)
• 09:45
Overview of the Present Progresses and Activities on the Chinese Fusion Engineering Test Reactor 25m
The Chinese Fusion Engineering Testing Reactor (CFETR) is the next device for the Chinese magnetic confinement fusion (MCF) program which aims to bridge the gaps between the fusion experiment ITER and the demonstration reactor DEMO. CFETR will be operated in two phases: Steady-state operation and tritium self-sustainment will be the two key issues for the first phase with a modest fusion power up to 200 MW. The second phase aims for DEMO validation with a fusion power over 1 GW. Advanced H-mode physics, high magnetic fields up to 7T, high frequency electron cyclotron resonance heating (230 GHz) & lower hybrid current drive (7.5GHz) together with off-axis negative-ion neutral beam injection will be used for achieving steady-state advanced operation. The detailed design, research and development activities including integrated modeling, R&D on high field magnet, material, T plant, remote handling, physical validation on EAST tokamak to demonstrate feasibility of high performance steady state operation, and future MCF road map will be introduced in this paper.
Speaker: Prof. Yuanxi WAN (Institute of Plasama Physics, Chinese Academy of Sciences)
• 08:30 12:30
Poster 1: P1
• 08:30
Active control/stabilization of locked mode in tokamaks at high magnetic Reynolds number 4h
We report a numerical study of a mode locking in tokamaks, which reveals an active stabilization effect of the control field against the locking event. We developed the resistive MHD simulation code “AEOLUS-IT”, which can simulate mode locking, where the magnetic island interacts with error/control field, under JT-60SA class high magnetic Reynolds number condition. The developed code successfully simulates the stabilization effect of the control field against the error field, which reveals a frequency dependence of the control field for suppressing the island evolution. The obtained dependencies have different natures between high and low magnetic Reynolds number (large scale and medium size tokamaks), which agrees well with the theoretical prediction. Taking into account of the successful calculation of the interaction between magnetic island and the error/control field under the high magnetic Reynolds number condition, as well as the adoption of the flux coordinate system, the developed code will enable us not only to check the agreement between our numerical studies and future JT-60SA experiments, but also to predict the error-field threshold in ITER.
Speaker: Dr Shizuo Inoue (Japan Atomic Energy Agency)
• 08:30
Advances in Numerical Modelling of MGI Mitigated Disruptions in ITER 4h
Disruption mitigation with use of the massive injection of noble gases (MGI) is widely used and experimentally validated on contemporary tokamaks. The disruption mitigation system (DMS) in ITER is aiming to subsequently or simultaneously achieve a solution for 3 main goals including mitigation of the heat loads on the plasma facing components during thermal quenches (TQs), keeping tolerable electro-mechanical loads on the conducting structures surrounding the plasma, and preventing the appearance of or suppressing the relativistic electron (RE) beams at the current quench (CQ) stage of the disruption. To assess the feasibility and operation domain of the ITER DMS, extended simulations are needed. The present report describes recent developments of the physical models for accurate and effective simulations of mitigated disruptions in ITER. The integrating core of these simulations is the Disruption Simulator based on the DINA code (DINA-DS). For specific conditions of impurity-dominated CQ plasmas, a special transport solver has been developed. Ionization of injected impurities due to interactions with REs is taken into account in the presented advanced transport model. MGI mitigated CQs are accompanied by fast vertical movement of the plasma column. A precise evaluation of the eddy currents induced in blanket modules and in the vacuum vessel has to take these dynamics into account. Representation of the ITER double wall vacuum vessel structure as 2 sets of 50 thin rings with rectangular cross-sections and relevant resistances for the inner and outer walls provides the necessary accuracy of the calculations. Recently, DINA has been updated to include parallel heat fluxes in the halo region for a complete energy balance. This provides the basis for a better estimate of the halo temperature and, therefore, plasma resistivity, affecting the resulting halo current amplitudes and CQ dynamics. The evolution of the RE distribution function in DINA-DS is simulated with the use of a recently developed analytical model. Knowledge of the RE distribution function instead of just RE current is of principal importance in assessing the total kinetic energy deposited to the first wall due to the loss of REs. Representative scenarios of mitigated disruptions in ITER simulated with updated models are presented. The operation domain for the ITER DMS based on MGI is discussed.
Speaker: Dr Victor Lukash (National Research Centre «Kurchatov Institute», Pl. Kurchatova 1, Moscow 123182, Russia)
• 08:30
An analytic scaling relation for the maximum tokamak elongation against n=0 MHD resistive wall modes 4h
In this study, the maximum achievable elongation in a tokamak against the n=0 MHD resistive wall mode is investigated theoretically and compared with experimental observations. A highly elongated plasma is desirable to increase plasma pressure and confinement for high fusion power output. However, there is a limit on the maximum achievable elongation which is set by vertical instabilities driven by the n=0 MHD mode. This limit can be increased by optimizing several parameters characterizing the plasma and the wall. The purpose of our study is to explore how and to which extent this can be done. Specifically, we extend many earlier calculations of the n=0 mode to determine maximum elongation as a function of dimensionless parameters describing (1) the plasma profile (beta_p and l_i), (2) the plasma shape (eps and delta), (3) the wall radius (b/a) and (4) most importantly the feedback system capability parameter gamma*tau. We make use of a new formulation of n=0 MHD theory developed in our recent study [Freidberg et. al. 2015; Lee et. al. 2015] that reduces the 2-D stability problem into a 1-D problem. This method includes all the physics of ideal MHD axisymmetric instability but it reduces the computation time significantly so that many parameters can be explored during the optimization process. We have explored a wide range of parameter space, and compared our results with data from tokamak experiments. Perhaps the most useful final result is a simple analytic fit to the simulations which gives the maximum elongation and corresponding optimized triangularity as functions kappa(eps,beta_p,l_i,b/a,gamma*tau) and delta(eps,beta_p,l_i,b/a,gamma*tau). These theoretically obtained scaling relations should be useful for determining optimum plasma shape in current experiments and future tokamak designs.
Speaker: Dr Jungpyo Lee (MIT Plasma Science and Fusion Center, USA)
• 08:30
Collisional generation of runaway electron seed distributions leading to sub-criticality, avalanche, or fast transfer 4h
Well before ITER operations begin, we must have a comprehensive understanding of the potential for runaway electron generation, as well as methods for their control and mitigation, as the destructive potential to the plasma facing components is severely intolerable. This makes for a unique situation in requiring an assessment based on plasma theory and computation well before validation experiments can be performed. Among the most important questions given a thermal collapse event is that of how many seed electrons are available for runaway acceleration and the avalanche process. Seed electrons remain with a kinetic energy above the critical energy for runaway after a thermal quench, either natural or induced. The expected seed generation is a critical question that needs to be addressed, and new methods are now available to do so. The most important source of seed electrons is the high-energy tail of the pre-thermal-quench Maxwellian. This high energy tail can be lost in two ways: (1) collisional drag on cold electrons or (2) loss to the walls if all the magnetic surfaces within the plasma are destroyed. In this study, we use the kinetic equation for electrons and ions to investigate how different cooling scenarios lead to different seed distributions. Given any initial distribution, we study their subsequent avalanche and acceleration to runaway with Adjoint and test particle methods [Chang Liu, Dylan P. Brennan, Amitava Bhattacharjee and Allen H. Boozer, Phys. Plasmas 23, 010702 (2016)]. This method gives an accurate calculation of the runaway threshold by including the collisional drag of background electrons (assuming they are Maxwellian), pitch angle scattering, and synchrotron and Bremsstrahlung radiation. A resulting probability to runaway is determined in phase space, which has a sharp transition, such that electrons with energy above this transition become highly likely to runaway. Summing the electrons above this threshold determines the number of seed electrons N_s. When N_s exceeds the number of relativistic electrons needed to produce the entire equilibrium current, fast transfer to runaway current is possible. Alternatively, N_s can be small enough that the runaway process is too slow to cause any significant runaway population on the experimental timescale. Between these limits, the avalanche process determines the runaway population.
Speaker: Dylan Brennan (Princeton Plasma Physics Laboratory)
• 08:30
Current profile shape effects on the formation and termination of runaway beams in tokamak disruptions and implications for ITER 4h
Runaway electrons (REs) generated during disruptions are usually found to deposit their energy in very short pulses and on localized areas of the plasma facing components (PFCs). In ITER, there is serious concern about the potential that large amounts of MeV REs generated during the disruption current quench (CQ) have for erosion / melting of the PFCs. Although zero-dimensional (0-D) modeling has shown to provide a rather complete physics picture of the CQ and termination phases of the disruption, there is evidence indicating that current profile shape effects could be important. In this work, a one dimensional model (1-D) beyond the 0-D model is used to evaluate effects associated with the evolution of the plasma and RE current profiles during the disruption. The model predictions are found to be in agreement with measurements of the plasma internal inductance for 2 MA JET disruptions with RE current plateau formation. The resulting runaway plasma is more peaked in the plasma center than the pre-disruption plasma current. The peaking decreases when the RE current increases and is also found to be dependent on the runaway seed profile shape, increasing with the internal inductance of the seed current. These results can have important implications for ITER as: (1) due to the increase in the plasma internal inductance, for the same RE current magnitude, the magnetic energy of the RE plasma would be substantially larger; (2) the post-CQ plasma current profile might be MHD unstable as plasmas with peaked current profiles can be prone to the tearing-mode instability. Moreover, the magnetic energy does not scale linearly with the square of the RE current. In order to investigate these effects, an integrated 1-D analysis of the runaway beam formation and termination during disruptions in ITER has been carried out, and including the essentials of the involved physical processes such as the main RE generation mechanisms expected in ITER as well as corrections to the RE dynamics to account for the collisions of the RE electrons with the partially stripped impurity ions. This work was carried out within the framework of the EUROfusion Consortium and has received funding from the Euratom research and training programme 2014-2018 under grant agreement No 633053. The views and opinions expressed herein do not necessarily reflect those of the European Commission.
Speaker: Dr Jose Ramon Martin-Solis (Universidad Carlos III de Madrid)
• 08:30
Development of Multi-Frequency Mega-Watt Gyrotrons for Fusion Devices in JAEA 4h
Mega-watt gyrotrons with frequency tuning have become essential devices in fusion science to perform effective EC H&CD. JAEA is developing two types of multi-frequency gyrotrons equipped with a triode magnetron injection gun for ITER and JT-60SA. A TE31,11 mode, which is a candidate mode for 170 GHz oscillation, has sufficient margin for cavity heat-load in 1 MW operation, and it has a great advantage for multi-frequency oscillation. In the JT-60SA project, EC H&CD by second harmonic EC waves are planned using nine sets of 110 GHz/138 GHz dual-frequency gyrotrons to broaden the experimental research area. In FEC2014, demonstrations of 1 MW oscillations for 2 s at 170 GHz/137 GHz/104 GHz with the ITER gyrotron and achievement of 1 MW oscillations for 100 s at 110 GHz /138 GHz in the JT-60SA gyrotron were reported as world records. After FEC2014, oscillation methods to improve the efficiency at 170 GHz for ITER requirements and higher frequency oscillation for the demo-class reactor were investigated. For the JT-60SA gyrotron, the operation area was expanded to surpass maximum performance (1.5 MW/4 s) of the previous JT-60 110 GHz gyrotron. TE31,11 mode oscillations were often prevented by adjacent counter-rotating (ctr-) modes such as TE29,12, and TE28,12 modes. By introducing active anode-voltage control and beam-radius control to suppress adjacent counter-rotating modes, start-up of TE31,11 mode becomes stable and the overall efficiencies achieved ~ 50 % up to 1.1 MW. In looking ahead to a future gyrotron for the demo-class reactor, 203 GHz oscillation of higher-order volume mode (TE37,13) was performed for the first time by taking advantage of the multi-frequency gyrotron feature. In preliminary testing at 203 GHz, 0.9 MW for 0.3 ms and 0.42 MW for 5 s were demonstrated. ITER gyrotron having mega-watt-class power at four frequencies in wide range over 100 GHz was developed. High power gyrotron development toward 1.5 – 2 MW oscillation for several seconds has been carried for further extension of the experiment regime of high performance plasma in JT-60SA. In a test conducted in 2015, achievements of 1.8 MW/1.2 s at 110 GHz (TE22,8 mode) in non-coaxial type gyrotron and high-power oscillation of 1.3 MW/1.3 sat 138 GHz (TE27,10 mode)and 1 MW/1 s of 82 GHz (TE17,6 mode) have been demonstrated as a new world record.
Speaker: Dr Ryosuke Ikeda (Japan Atomic Energy Agency)
• 08:30
Development of Over MW Gyrotrons for Fusion at Frequencis from 14 GHz to Sub-terahertz 4h
Megawatt (MW) gyrotrons with a wide frequency range from 14 to 300 GHz are being developed for the collaborative Electron Cyclotron Heating (ECH) study of advanced fusion devices and DEMO reactor. (1) In the first experiment of 300 GHz gyrotron, an output power of over 0.5 MW with TE32,18 single-mode was achieved with a pulse width of 2 ms. This is the first report of MW level oscillation with the DEMO-relevant ECH gyrotron mode. It was also found that the reflection at the output window affects the oscillation mode determination. (2) A new record of the 28 GHz gyrotron output of 1.38 MW was obtained. The fabrication of a newly designed tube aimed at a dual-frequency output power of 2 MW at 28 GHz (0.4 MW CW) and 1 MW at 35 GHz has begun, with all components ready for assembly. Before installing a double-disk window in the dual-frequency gyrotron, we confirmed the dependence of reflective power on the coolant thickness including the reflective power less than 2 % by the cold test using a Gunn diode power of 1 W and the hot test using the gyrotron output power of 600 kW. (3) Based on the successful results of 77 and 154 GHz LHD tubes, the new design of a 154/116 GHz dual-frequency gyrotron with output of over 1.5 MW has been presented.
Speaker: Dr Tsuyoshi Kariya (Plasma Research Center, University of Tsukuba)
• 08:30
Drift-Alfven Instabilities and Turbulence of Magnetic Field Aligned Shear Flows 4h
Numerous experimental observations from number of tokamaks and stellarators have found large nearly sonic magnetic field aligned (parallel) shear flows that are peaked at the last closed flux surfaces and extend for a few centimeters into the plasma and into the far SOL. The important consequence is that these plasma regions are unstable in the presence of shear flows. The shear flows along the magnetic field are the additional sources of free energy for the modification of the instabilities, existing in shearless plasma flows, as well as for the development of specific shear flow driven (SFD) instabilities, which are absent in the shearless plasma. It was reported recently, that in plasmas with ion temperature equal to or even higher than the electron temperature, that is the case relevant to tokamak plasma, the Kelvin-Helmholtz (or D'Angelo) driving mechanism of the excitation the hydrodynamic instabilities changes onto the combined effect of the velocity shear and ion Landau damping. It results in the development of a new set of the ion-kinetic SFD instabilities, which distinguish by strong interaction of waves with thermal ions. This is a striking difference between the instabilities of the parallel shear flows and the shearless plasma, where the ion Landau damping is, as a rule, a process that suppresses the development of the drift instabilities. It was found, that in the parallel shear flow of plasmas with comparable ion and electron temperatures two distinct drift--Alfven instabilities(DAI) may be developed: the shear flow modified DAI, which develops due to the inverse electron Landau damping and exists in the shearless plasma as well, and the SFD DAI, which develops due to the combined effect of the velocity shear and ion Landau damping and is absent in the shearless plasma flows. In this report, we present the results of the investigation of the nonlinear saturation of both these instabilities and the processes of the anomalous heating and transport of ions. The results of the analytical and numerical investigations of the SFD DAI and corresponding turbulence of the shear flow with inhomogeneous ion temperature, which develops due to the coupled reinforcing action of parallel flow shear, ion temperature gradient and ion Landau damping, will be given. This work was supported by NRF of Korea (Grant No. NRF-2014M1A7A1A03029878) and BK21PLUS.
Speaker: Prof. Volodymyr Mykhaylenko (Pusan National University)
• 08:30
Equilibrium solutions of MHD equations for GAMs in the edge tokamak plasma 4h
Numerical calculations of nonlinear MHD equations in frames of reduced two-fluid Braginskij equations for geodesic acoustic modes (GAM) with n = 0, m = 0, +1, -1 in high collisional edge tokamak plasma were performed. N = 0, m = 0, +1, -1. It was shown that with account of parallel dissipation (finite conductivity sigma//) allows us to obtain the steady state equilibrium solutions for GAMs. The obtained 2D equilibrium includes the velocity of poloidal rotation and the equilibrium electric potential, which value is close to well-known Pfirsh-Schluter potential. It was shown that the main role in formation of the equilibrium poloidal rotation plays two forces: the Stringer-Winsor force and the neoclassical force, linked with the parallel viscosity. Maximum values of GAM are located near the maximum of pressure gradient. Calculated radial profile of electric field E looks like the parabolic negative well (E < 0).
Speaker: Dr Remir Shurygin (NRC ‘Kurchatov Institute’, Moscow, Russia)
• 08:30
Excitation of frequency jump by barely Passing Electrons 4h
An e-fishbone frequency jump has been observed on Tore Supra , which is important for the redistribution of energetic electrons and energetic particle losses. E-fishbone periodic frequency jump phenomena are also observed on HL-2A . Soft X-ray tomography shows that the poloidal and toroidal mode numbers are 1/1 and 2/2 with the frequency jump. In this paper we present a theoretical base of the frequency jump in the e-fishbone experiments. It is identified that barely passing electrons are the drive of the e-fishbone, rather than the trapped electrons. The frequency jumps in HL-2A E-fishbone experiments are numerically reproduced. E-fishbone frequency increases with the hot electron energy which is consistent with the experiments. The growth rate of the mode (m=2, n=2) is greater than the one of the mode (m=1, n=1) in contrast to the pure MHD prediction. The calculated temporal evolutions of the hot electron energy and the kink mode amplitude are periodic which in good agreement with the observed e-fishbone jump cycle. The theory provides an insight on HL-2A and Tore Supra experiments.
Speaker: Prof. Zhongtian Wang (School of Science, Xihua University;Southwestern Institute of Physics)
• 08:30
Extension of numerical matching method to weakly nonlinear regime -- beyond the Rutherford theory of magnetic island evolution 4h
The theory of matched asymptotic expansion for resistive MHD is well established for linear modes [1] and for weakly nonlinear evolution [2]. Since then many applications of the Rutherford equation [2] have made much progress in fusion research, especially in the neoclassical tearing mode (NTM) studies [3]. However, the theoretical framework is still based on the Rutherford equation essentially. We have recently developed a new framework for linear stability analysis of resistive MHD, the numerical matching method [4]. This method utilizes a finite-width inner region around a resonant surface, instead of an infinitely thin inner layer. We devised the boundary condition for the direct matching at interfaces between the inner and outer regions. Then we succeeded to remove difficulties that remain in the numerical applications of the traditional matched asymptotic expansion even though sophisticated theories were developed [5]. We developed both an eigenvalue and an initial-value approaches. In this paper, we extend the initial-value approach of the numerical matching method to weakly nonlinear cases. In the presentation, we will explain the theory, and will show numerical results using reduced MHD in cylindrical plasmas that successfully reproduced Rutherford regime of magnetic island evolution. The computational cost is reduced, making inclusion of detailed physical effects easier because such a model requires more efficient solution method. Our new method will certainly aid understanding physics and will substantially contribute to the prediction of MHD activities such as NTMs in fusion plasmas. This work was supported by KAKENHI Grant No. 23760805 and No. 15K06647. [1] H. P. Furth, J. Killeen and M. N. Rosenbluth, Phys. Fluids 6, 459 (1963). [2] P. H. Rutherford, Phys. Fluids 16, 1903 (1973). [3] A. I. Smolyakov, Plasma Phys. Control. Fusion 35, 657 (1993); O. Sauter et al., Phys. Plasmas 4, 1654 (1997); R. J. La Haye, Phys. Plasmas 13, 055501 (2006). [4] M. Furukawa, S.Tokuda and L.-J. Zheng, Phys. Plasmas 17, 052502 (2010); M. Furukawa and S. Tokuda, Phys. Plasmas 18, 062502 (2011); Id., Phys. Plasmas 19, 102511 (2012). [5] A. Pletzer and R. L. Dewar, J. Plasma Phys. 45, 427 (1991); A. Pletzer, A. Bondeson and R. L. Dewar, J. Comput. Phys. 115, 530 (1994); S. Tokuda and T. Watanabe, Phys. Plasmas 6, 3012 (1999); S. Tokuda, Nucl. Fusion 41, 1037 (2001).
Speaker: Prof. Masaru FURUKAWA (Graduate School of Engineering, Tottori University)
• 08:30
First Principle Fluid Modelling of Neoclassical Tearing Modes and of their Control 4h
The confinement degradation of Tokamak plasma by magnetic islands motivates numerous approaches to better understand their dynamics and possible suppression. We report here on nonlinear simulations using a consistent two-fluid implementation of neoclassical friction forces in the framework of the toroidal Magneto-Hydro-Dynamic model of the XTOR code, recently improved with the implementation of parallel heat fluxes that allows recovering a self-generated bootstrap current that compares well with analytical formulae. The island saturation width increases as expected with the bootstrap current fraction, but with a much weaker dependence than predicted by a Rutherford-type equation. We evidence the strong influence of diamagnetic rotations and to a lesser extent of neoclassical friction on the saturation size and shape of the island, by comparing with a pure resistive MHD simulation. In metastable plasmas, a seed is required for triggering a NTM. We find in a case taken from Asdex-Upgrade that the shape of the seed has an influence on the island dynamics that follows. The control of these magnetic islands by the coupling of Radio-Frequency waves is modelled by a source term in Ohm's law accounting for the propagation of accelerated electrons along field lines. This implementation is validated against analytical theory regarding the stabilization efficiency, and the 3D spatial localization of the RF source is shown to impact the island dynamics when the plasma is nearly static: the flip instability is recovered and described, and the possibility to create an island from the RF current filament is evidenced. For metastable modes, a control method allowing the island to go below the critical size is appropriate. A promising stabilization technique based on the radial sweeping of the ECCD source has been successfully demonstrated in TCV and Asdex-Upgrade. Our numerical investigation shows that the effective stabilization efficiency of the sweep is better when remaining close to the resonance or when being decentred on the outside. These developments on a fluid implementation of neoclassical physics and RF sources in a global MHD code allow further investigations on NTM triggering and saturation, on ECCD stabilization process, as well as on their possible characterizations thanks to density and temperature fluctuation signals.
Speaker: Mr PATRICK MAGET (CEA)
• 08:30
Impact of Kinetic Effects of Energetic Particles on Resistive Wall Mode Stability in Rotating High-beta Plasmas 4h
We found that inclusion of self-consistent rotation effect in the energetic particles’ dynamics has significant impact on resistive wall mode (RWM) stability in tokamaks. For the first time, we apply the extended kinetic-magnetohydrodynamic (MHD) theory for rotating plasmas to energetic particles. The theory invokes an extended energy exchange term between the MHD mode and energetic particles’ motion. In this study, the extended theory has been applied to RWM stability analysis in high-β JT-60SA plasmas. By using a new model equilibrium distribution function of energetic particles, it is found that extended energy exchange terms enhance the stabilization effect.
Speaker: Dr Junya Shiraishi (Japan Atomic Energy Agency)
• 08:30
Long-pulse acceleration of 1MeV negative ion beams toward ITER and JT-60SA neutral beam injectors 4h
In order to realize the negative-ion-based neutral beam (NB) systems for ITER and JT-60SA, development of the Multi-Aperture and Multi-Grid (MAMuG) electrostatic accelerator is one of common critical issues. For these NB injectors, 5- and 3-stage MAMuG accelerators are being developed to achieve the acceleration of negative ion beams up to 1 MeV, 40 A (200 A/m^2) for 3600 s and 0.5 MeV, 22 A (130 A/m^2) for 100 s, respectively. However, there were no experiments of long-pulse MeV-class beam acceleration. Though JAEA achieved the rated beam energy of 1 MeV, the pulse duration was limited to be less than 1 s [1] due to a low voltage holding capability and high grid power loads. After the last FEC conference, following issues were investigated such as multi-grid effect on the voltage holding capability and reduction of the grid power loads. New accelerators have been designed to realize stable voltage holding by taking into account the multi-grid effect on voltage holding capability, which satisfies the requirement of beam energy for ITER and JT-60SA with 5-stage and 3-stage, respectively. The grid power load has been suppressed less than a half of the design values of the accelerators by modifying the geometry of the extractor and the acceleration grids to suppress generation of secondary electrons. By applying the developed techniques based on the R&D results, the hydrogen negative ion beams of 0.97 MeV, 190 A/m^2 have been successfully accelerated up to 60 s from the ITER prototype accelerator. The pulse duration of such high power density negative ion beams (~184 MW/m^2) has been extended from 0.4 to 60 s, which is the longest pulse length in the world. There is no limitation to extend the pulse duration, since no degradation of the voltage holding has been observed during the long-pulse operations neither by cesium accumulation nor by thermal damage of the acceleration grids. This achievement is one of breakthroughs toward the realization of the high-energy NB systems. [1] A. Kojima, et al., Nucl. Fusion 55 (2015) 063006.
Speaker: Dr JUNICHI HIRATSUKA (Japan Atomic Energy Agency)
• 08:30
Magnetic Island Behavior under Non-axisymmetric Halo Current at Vertical Displacement Event 4h
The non-axisymmetric halo current arising due to loss of plasma vertical equilibrium, the so-called vertical displacement event (VDE), during plasma disruption in vertically elongated tokamak can be one of possible sources of helical magnetic perturbation. This perturbation penetrate into plasma producing magnetic islands in the vicinity of resonant magnetic surface with the same helicity. Results of simulation and analysis of magnetic island production by helical magnetic perturbation generated under non-axisymmetric halo current are presented. Some predictions for ITER-like tokamak are presented with view of the disruption risk analysis. Calculations are carried out with the TEAR code based on the two-fluid MHD approximation. The radial distribution of the magnetic flux perturbation is calculated with account of the external helical field produced by halo current. The equations for the magnetic flux perturbation describe the dynamics of the tearing mode depending on plasma rotation. In sequence, this rotation is affected by electromagnetic forces depending on the tearing mode magnetic field and external magnetic perturbation. Numerically, the diffusion-type equations for the helical flux function and for the plasma rotation velocity are treated in a similar way. The magnetic island behavior is analyzed for different plasma parameters and possible mode numbers. The width of the produced magnetic islands extends to a significant part of plasma minor radius. These magnetic islands can affect plasma stability, equilibrium and confinement, in particular the confinement of runaway electrons, thus affecting the development of the disruption and its impact on tokamak components.
Speaker: Dr Nikolay Ivanov (Kurchatov Institute)
• 08:30
Magneto-thermal Reconnection Processes, Related Angular Momentum Transport issues and Formation of High Energy Particle Populations 4h
In the context of a two-fluid theory of magnetic reconnection [1], when the longitudinal electron thermal conductivity is relatively large, the perturbed electron temperature tends to become singular [2] in the presence of a reconnected field component and an electron temperature gradient. A transverse thermal diffusivity is introduced in order to remove this singularity while a finite inductivity’’ can remove the singularity of the corresponding transverse plasma displacement [1]. Then i) a new magneto-thermal reconnection’’ producing mode, driven by the electron temperature gradient, and involving a considerable range of scale distances is found [3]; ii) the characteristic widths of the layers in which magnetic reconnections takes place remain significant even when the macroscopic distances involved in the process are very large; iii) the phase velocities of the modes that are found can be both in the direction of the electron diamagnetic velocity as well as those in the opposite (ion) direction. A numerical solution of the complete set of equations has been carried out and followed by a simplified analytical reformulation of the problem. The mode growth rate is related to the effects of a finite viscous diffusion coefficient or to those of a small electrical resistivity. The features that can lead to a possible explanation of the fact that high energy particle populations are produced during reconnection events involve mode-particle resonances producing the transfer of energy to super-thermal particle populations [4] and the spatial near-singularity of the electron temperature that can enhance the thermal energy of particles in one region while depleting that of particles in a contiguous region [3]. The low collisionality modes that produce magnetic reconnection can extract angular momentum from the plasma column and thereby sustain a spontaneous rotation’’ [5] of it. This process is to be considered in addition to that associated with electrostatic modes excited at the edge of the plasma column [5]. Supported by the U.S. DOE, award DE-FG02-03ER54700. [1]B. Coppi, Phys. Fluids 8, 2273 (1965). [2]B. Coppi, B. Basu, P. Montag, et al. Nucl. Fus., 55, 093018 (2015). [3]B. Coppi, MIT (LNS) Report HEP 15/06 (2015). In print for Fizika Plazmy. [4]B. Coppi, L. Sugiyama, J. Mark and G. Bertin Ann. Phys. 119, 370 (1979). [5]B. Coppi, Nucl. Fus. 42, (2002).
Speaker: Prof. Bruno Coppi (M.I.T.)
• 08:30
MHD stability of ITER H-mode confinement with pedestal bootstrap current and diamagnetic effects taken into account 4h
MHD stability of ITER H-mode confinement is investigated with bootstrap current included for equilibrium, together with diamagnetic drift and rotation effects for stability. The ITER pedestal has high temperature, so the bootstrap current is large and diamagnetic effects are important. We construct numerically ITER equilibria with bootstrap current taken into account.Especially, we have considered a more realistic scenario in which density and temperature profiles can be different. The direct consequence of bootstrap current effects on equilibrium is the modification of local safety factor profile at pedestal, so that the magnetic shear can be reduced or reversed locally. This local q value is referred to as $q_s$. This q profile change results in a dramatic change of MHD mode behavior. The stability of ITER numerical equilibria is investigated with AEGIS code. Both low-n and peeling-ballooning modes are investigated. Note that pressure gradient at pedestal is steep. High resolution computation is needed. Since AEGIS code is an adaptive code, it can well handle this problem. Also, the analytical continuation technique based on the Cauchy-Riemann condition of dispersion relation is applied, so that the marginal stability conditions can be determined. It is found that the pedestal stability depends not only on the edge current ($J_{ped}$) and pressure gradient ($p'_{ped}$), but also on the $q_s$ value. This shows that the pedestal stability can be affected by the global current profile. The diamagnetic drift and rotation effects are also investigated. Both numerical scheme and results will be presented. The physical interpretation will be explained.
Speaker: Linjin Zheng (University of Texas at Austin)
• 08:30
Modeling and Simulation of Pedestal Control Techniques for NSTX-U 4h
In this paper we present high level simulations and modeling of pedestal control for NSTX-U. Real-time pedestal control is a crucial topic for future fusion reactors and ITER where pedestal has to be kept Edge-Localized-Modes (or ELMs) free. We developed and tested many different control schemes to adjust and regulate the pedestal at DIII-D and we plan to test them on NSTX-U. But to do this it is important to understand the physics bases for how the control actuators affect the pedestal. It has been observed many times that a control scheme that work for a specific machine or a regime might not be applicable to other machines and regimes. This is especially the case for future reactors such as ITER. We thus do high-level numerical simulations with the M3D-C1 code. M3D-C1 has been developed to study the plasma response when several actuators are triggered (gas puffing, 3D magnetic perturbations and LGI). The aim is to combine all these methods to get an adaptive and automatic pedestal control in tokamaks. In this paper, we focus on the effect of each actuators on the ELM frequency and amplitude. First modeling results of ELM-triggering by LGI have been obtained with M3D-C1. Mesh adaptation techniques and high order 3D finite elements allow simulation of sub-mm granules, without constraints on the granule toroidal width. This unique capability of M3D-C1 allows the simulation of realistic pellet sizes. For this study, two models for LGI are implemented in M3D-C1. The first one is a Neutral Gas Shielding Model (NGS) calibrated on DIII-D experimental measurements of the Lithium granule ablation rates. The second one is valid for small size granules (sub-mm) where the contribution of plasma ions to the ablation of the granule is not negligible. In the simulation, it takes about 100 microseconds for the pellet to totally being ablated. NSTX-U L-mode and H-mode simulations have been done and will be compared to available experimental data. Moreover, stability calculations from ELITE and M3D-C1 during the penetration process will be presented, as well as comparison with the EPED code. Among the granule parameters, it is found that the most important are the type of element in the pellet, its size and the angle of attack.
Speaker: Dr Alexandre Fil (Princeton University)
• 08:30
New Results of Development of Gyrotrons for Plasma Fusion Installations 4h
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 seconds. To provide operation with indicated parameters the gyrotrons have very large transverse cavity sizes, output barrier windows made of CVD diamond discs, effective collectors with particle energy recovery. In ITER installation there will be 24 gyrotron systems with 1 MW power each. Russian contribution consists of 8 gyrotron systems. ITER requirements are: frequency 170 GHz, 1 MW power, 1000 s pulse duration, high efficiency of the gyrotrons over 50%, possibility of power modulation with frequency up to 5 kHz, compatibility of the gyrotron complex with ITER control system. In May, 2015 a Prototype of ITER Gyrotron System was completed and its operation was demonstrated. The system consists of gyrotron oscillator, liquid-free superconducting magnet, supplementary magnets, several electric power supplies, cooling systems control and protection systems, and other auxiliary units. The tests were performed in presents of ITER IO and ITER RF DA representatives. In October, 2015 Final Design Procedure for the gyrotron system was successfully passed. High-level parameters were also achieved with long-pulse 140 GHz gyrotrons developed for EAST and KSTAR installations. Significant results were shown on the way to 1.5-2MW, CW gyrotrons. The development of higher frequency (230-700 GHz) gyrotrons for future plasma installations and for plasma diagnostics began. Novel ideas were proposed to enhance gyrotron operation
Speaker: Prof. Grigory Denisov (Institute of Applied Physics Russian Academy of Sciences)
• 08:30
Non-linear MHD modelling of Edge Localized Modes dynamics. 4h
The non-linear MHD modelling of full ELM crash dynamics was performed using JOREK code for KSTAR pulse parameters and compared to the ECEI diagnostic observations. Some experimentally observed trends were reproduced in modelling. In particular the localization of the peeling-ballooning modes in the pedestal region inside the separatrix, the most unstable modes toroidal numbers and structures, poloidal velocity and the direction of the modes rotation are similar to the experimental observations on KSTAR. The rotation of the modes in electron diamagnetic direction is more common observation in many tokamaks due to the typically large negative radial electric field well in the pedestal region. However it was shown in JOREK modelling that at relatively large toroidal plasma rotation, which was the case for the KSTAR pulse modeled in the paper, the modes can rotate in the ion diamagnetic direction before ELM crash similar to KSTAR ECEI observations. Multi-modes (n=1-8) modelling demonstrated the acceleration of growth of the peeling-ballooning modes and even destabilization of previously linearly stable modes approaching the ELM crash due to the strongly increasing non-linear coupling at this stage of the instability. Moreover, a strongly sheared mean poloidal flow occurs on the non-linear phase of an ELM leading to the filaments detachment from the main plasma in the form of “blobs” which propagate in the SOL mainly in the ion diamagnetic direction. In/out divertor heat flux asymmetry (~2:1) due to ELM crash was obtained with two fluid diamagnetic drifts included in the modelling.
Speaker: Dr Marina Becoulet (IRFM/CEA)
• 08:30
Non-linear MHD Simulations of Pellet Triggered ELMs 4h
ITER operation relies on the achievement of the H-mode confinement regime, which is expected to lead to the quasi-periodic triggering of ELMs (Edge Localized Modes). The energy fluxes associated with natural ELMs will produce excessive erosion and/or damage on the plasma facing component. Controlled triggering of ELMs by the injection of small pellets at frequencies exceeding those of natural ELMs is one of the foreseen schemes to control ELMs in ITER. Although the technique has been demonstrated to decrease ELM size successfully in ASDEX Upgrade [1], JET [2], and DIII-D [3], uncertainties still remain regarding the physics understanding as well as of the consequence of its application, such as localised power loads associated with this technique [4]. Modelling of ELM triggering by pellet injection for ASDEX Upgrade, JET discharges, and the ITER 15MA Q=10 scenario has been carried out with the non-linear MHD code JOREK [5, 6]. The JOREK code allows the simulation of a full pellet triggered ELM cycle, i.e. to study the non-linear consequences of a pellet triggered instability and determine the ELM energy and particle losses. The dependence of the pellet injection geometry has been studied and it is found that pellet injection from High Field Side eases the pellet ELM triggering, consistent with the findings of DIII-D [5]. The dependence of the power deposition asymmetry on the injection geometry and the consequences for ITER with the JOREK simulation of JET which confirms the result will be presented. Detailed investigation of the particle and the energy loss during the full ELM cycle of pellet triggered ELM is will be presented. Acknowledgement This work has been carried out within the framework of the EUROfusion Consortium and has received funding from the Euratom research and training programme 2014-2018 under grant agreement No 633053. The views and opinions expressed herein do not necessarily reflect either those of the European Commission or those of the ITER Organization. [1] P. Lang et al., Nucl. Fusion 44 665 (2004). [2] P. Lang et al., Nucl. Fusion 53 (2013) 073010. [3] L. Baylor et al., Phys. Rev. Lett. 110 (2013) 245001. [4] R. Wenninger et al., Plasma Phys. Control. Fusion 53 (2011) 105002. [5] S. Futatani et al., Nucl. Fusion 54, 073008 (2014). [6] G.T.A. Huysmans and O. Czarny, Nucl. Fusion 47, 659 (2007).
Speaker: Dr Shimpei Futatani (Barcelona Supercomputing Center)
• 08:30
Non-linear modeling of the Edge Localized Mode control by Resonant Magnetic Perturbations in ASDEX Upgrade 4h
One of the foreseen methods to control the Edge Localized Modes (ELMs) in ITER is the application of Resonant Magnetic Perturbations (RMPs), proved capable to mitigate or suppress ELMs in existing tokamaks. However the significant uncertainties that remain regarding the way plasma flows and ELMs interact with RMPs must be overcome to give reliable predictions for ITER. This work aims at assessing the impact of the different plasma responses (including resonant and kink components) on the ELM mitigation, in order to move towards more quantitative understanding of current experiments and better predictive capabilities for future experiments. Non-linear resistive MHD simulations were performed with the JOREK code, using input equilibrium profiles and n=2 RMP spectrum closely matching the experimental data of ASDEX Upgrade shots at low collisionality. In a first step, the interaction between n=2 RMPs and plasma flows is considered without ELMs. In experiments, a given RMP coil configuration was identified to lead to a stronger ELM mitigation: this is found to be correlated with the largest excitation of the kink response in the vicinity of the X-point observed in our modeling with JOREK (in good agreement with other modeling performed with MARS-F and VMEC). On the resonant surfaces q=m/n located at the edge, the coupling between this excited kink component (poloidal mode m+2) and the resonant component m induces the amplification of the resonant component, resulting in an enhanced ergodicity at the edge. The ergodicity and the large displacement of temperature and density near the X-point therefore generate an increased radial transport. In a second step, RMP effects on ELMs are considered in multi-harmonic n simulations. First results on the ELM mitigation induced by non-linear coupling of unstable modes with n=2 RMPs depending on the plasma response are presented and compared to experiments.
Speaker: Dr Francois Orain (Max Planck Institute for Plasma Physics, Garching, Germany)
• 08:30
Nonlinear 3D M3D-C1 Simulations of Tokamak Plasmas Crossing a MHD Linear Stability Boundary 4h
The goal of the present work is to better understand and develop a predictive capability for when approaching and crossing a MHD linear instability boundary leads to a thermal quench and subsequent disruption (hard limit), and when it just leads to increased transport or small amplitude oscillations (soft limit). Understanding the difference between hard and soft limits is crucial for effective disruption prediction and avoidance. We present several examples of both hard and soft beta limits. Recent advances in implicit numerical algorithms for solving the 3D extended magneto-hydrodynamic equations in strongly magnetized plasmas have enabled massively parallel simulations of the internal global dynamics of tokamaks that can use very large time steps which allow one to span the timescales of ideal MHD stability, magnetic reconnection, and particle, energy, and momentum transport. It is now possible and feasible to run these high-resolution time-dependent initial value simulations for 10^6 or more Alfvén times so as to span all relevant timescales in a single simulation. In addition, a new multi-region and adaptive meshing capability allows simulation of the self-consistent interaction of the plasma with a resistive wall. In the examples presented here, we begin the simulation with the plasma stable to all modes. During the simulation the plasma crosses a stability boundary due to evolving profiles, loss of control, or injection of mass, energy, and or flux. This can lead to saturation or disruption
Speaker: Dr Stephen jardin (UsPPPL)
• 08:30
Nonlinear extended-MHD modeling by the NIMROD code of broadband-MHD turbulence during DIII-D QH-mode discharges 4h
It is desirable to have an ITER H-mode regime that is quiescent to edge-localized modes (ELMs). ELMs deposit large, localized and impulsive heat loads that can damage the divertor. A quiescent regime with edge harmonic oscillations (EHO) or broadband MHD activity is observed in some DIII-D, JET, JT-60U, and ASDEX-U discharge scenarios [Garofalo et al, PoP (2015); Burrell et al., PoP (2012); Garofalo et al, NF (2011) and refs. within]. These ELM-free discharges have the pedestal-plasma confinement necessary for burning-plasma operation on ITER. The mode activity associated with the EHO or broadband MHD is characterized by small toroidal-mode numbers (n≈1-5) and is thus suitable for simulation with global MHD codes. The particle transport is enhanced during QH-mode, leading to essentially steady-state profiles in the pedestal region. Relative to QH-mode operation with EHO, operation with broadband MHD tends to occur at higher densities and lower rotation and thus may be more relevant to ITER. Nonlinear NIMROD simulations initialized from a reconstruction of a DIII-D QH-mode discharge with broadband MHD saturate into a turbulent state. Results from a nonlinear NIMROD simulation of DIII-D QH-mode shot 145098 at 4250ms with broadband MHD are presented. The measured toroidal and poloidal rotation profiles are included in the simulation as experimental observations indicate that the QH-mode operational regime is dependent on the rotation profile. The simulation develops into a saturated turbulent state and the n=1 and 2 modes become dominant through an inverse cascade. Each toroidal mode in the range of n=1-5 is dominant at a different time. The perturbations are advected and sheared apart in the counter-clockwise direction consistent with the direction of the poloidal flow inside the LCFS. Work towards validation through comparison to ECE, BES and Doppler reflectometry measurements is presented. Consistent with experimental observations during QH-mode, the simulated state leads to large particle transport relative to the thermal transport. A discussion of the transport assumptions built into our MHD modeling concludes that future QH-mode simulation studying the induced transport should run as a turbulence calculation where profiles are fixed and needs to include first-order FLR drift effects that stabilize high-n modes.
Speaker: Jacob King (Tech-X Corporation)
• 08:30
Nonlinear MHD simulations of Quiescent H-mode pedestal in DIII-D and implications for ITER 4h
Non-linear MHD simulations of DIII-D QH-mode plasmas have been performed with the non-linear MHD code JOREK as a first step towards determining whether the physics mechanisms leading to the QH-mode behaviour would be at work in ITER plasmas and thus whether this confinement regime can be considered as an alternative to the controlled Type I ELMy H-mode for ITER high Q operation. In the nonlinear MHD simulations it is found that low n kink-peeling modes (KPM) are unstable and grow to a saturated level, consistent with the physics picture put forward in linear study. The features of the dominant MHD modes found in the simulations of the KPM mode, which are due to its toroidal localization caused by the coupling of harmonics, are in good agreement with the observations of the EHO typically present in DIII-D QH-mode experiments. The influence of a realistic resistive wall in these DIII-D simulations shows that the inclusion of a resistive wall and plasma rotation has an effect on the non-linear KPM evolution. In this work, the non-linear evolution of MHD modes with toroidal mode numbers n from 0 to 20, including both kink-peeling modes and ballooning modes, will be investigated through MHD simulations starting from initial conditions either close to the ballooning or the kink-peeling mode limit in the edge stability diagram, both for DIII-D and ITER plasmas. The identification of the physics mechanisms that lead to the saturation of the KPM and to the appearance of the EHO in DIII-D QH-modes will allow us to evaluate whether this regime is an option for high fusion performance operation at the specific characteristics of ITER plasmas.
Speaker: Dr FENG LIU (University of Nice)
• 08:30
Nonlinear simulation of ELM dynamics in the presence of RMPs and pellet injection 4h
We report on nonlinear simulation studies on the dynamical behaviour of ELMs under the influence of RMPs and/or the presence of pellet injection using a two-fluid initial value electromagnetic nonlinear global code (CUTIE). The full set of model fluid equations are solved for the so-called mesoscale, an intermediate scale between the device size and the ion gyroradius, incorporating approximations for the underlying classical and neoclassical transport effects. To simulate ELMs we introduce a particle source in the confinement region and a particle sink in the edge region. The code also incorporates turbulent transport effects and allows the development of profile-turbulence interactions thereby enabling a self-consistent description of the evolution of the mode. To study ELM control using RMPs we have applied an n=2 static external magnetic perturbation at the edge and made numerous runs under varying conditions for the machine and plasma parameters typical of COMPASS-D. Our results show that ELM mitigation is possible for RMP powers beyond a specific threshold consistent with experimental observations of several tokamaks. The results also provide valuable insights into the RMP induced modifications of the complex nonlinear dynamics of the ELMs, in particular on the redistribution of mode energy and the cascading of energy to shorter scale lengths. We also observe a hysteresis in states as we increase the amplitude of RMPs and then decrease it to the same value. Preliminary simulations with pellet injection also show encouraging ELM mitigation results with corresponding changes in the ambient electromagnetic turbulence. Based on these results we have also used CUTIE in a predictive manner to map out parametric regions for safe operation of SST-1 by in terms of RMP thresholds and pellet pacing frequency for ELM control/mitigation in SST-1 H mode scenarios.
Speaker: Prof. Abhijit Sen (Institute for Plasma Research)
• 08:30
Numerical calculations of plasma response to external magnetic perturbations 4h
We investigate the effect of resistivity, mainly on pitch resonant responses induced by plasma rotation. As a confirmation of the newly developed code, we report that the detailed physics may not be important since the pitch resonant response is relatively weak at high resistivity and the penetration is strongly dependent on plasma rotation at low resistivity. At low resistivity, ion collisionality can affect the penetration of RMPs through poloidal flow. The preliminary quasilinear results with $n = 0$ parallel flow and radial electric field show that the torque induced by RMP may modify parallel flow significantly $t\gt 10^4 t_A \sim 1ms$ after RMP application. The detailed quasilinear responses will be presented with the possible implication on ELM suppression.
Speaker: Dr Juhyung Kim (National Fusion Research Institute)
• 08:30
Nyquist analysis of kinetic effects on the plasma response in NSTX and DIII-D experiments 4h
Externally applied, nonaxisymmetric magnetic perturbations can strongly modify tokamak plasmas, leading to the plasma response. Plasma response, often closely related to the resonant field amplification and to the ELM control using magnetic coils, has been systematically observed in tokamak experiments. In particular, the importance of drift kinetic effects on modifying the plasma response has been demonstrated via quantitative modeling of NSTX and DIII-D high beta experiments [1, 2]. In this work, Nyquist analysis, as a very powerful tool in stability theory, is applied to analyze the plasma response with the intrinsically stable plasmas, where the technique, combined with Padé approximation, provides the deep physics understanding of the plasma behavior. Based on the idea that the plasma response to externally applied 3D fields is often due to the linear combination of certain stable eigenmodes’ response, Nyquist analysis clearly shows how the kinetic effects change the damping rate of these stable eigenmodes in the plasma, without resorting to direct stability computations. The capability of Nyquist analysis to infer the plasma stability for potential ELM mitigation or suppression, direct observation of so-called multi-mode response and identification of amplification associated with the preferred eigenmode is also presented in this work, where the multi-model response is a phenomenon currently under extensive discussions [3]. The results suggest the application of Nyquist technique in 3D plasma response experiment since the plasma transfer function extracted from the experiments directly can be very useful to design the MHD control system and to better predict plasma behavior in future experiments. *This research was supported by U.S. DOE contracts #DE-AC02-09CH11466. [1] Z.R. Wang, M.J. Lanctot et al, 56th APS Conference, DPP.TI1.3 (2014). [2] Z.R. Wang, M.J. Lanctot et al, Phys, Rev. Lett. 114, 145005 (2015). [3] C. Paz-Soldan, R. Nazikian et al, Phys, Rev. Lett. 114, 105001(2015).
Speaker: Dr Zhirui Wang (Princeton Plasma Physics Laboratory)
• 08:30
Pfirsch-Tasso versus standard approaches in the plasma stability theory 4h
The paper is devoted to theoretical description of plasma stability in toroidal fusion systems with a resistive wall. Its aim is elimination of contradictions between different approaches and between theory and experiment. The study is related to two predictions stated as theorems, see [H. Tasso and G. N. Throumoulopoulos, Phys. Plasmas 18, 070702 (2011)] and references therein. One is that an MHD-unstable configuration with a dissipationless plasma surrounded by vacuum and possibly superconducting walls cannot be stabilized by introducing walls of finite electrical conductivity. The other is that in the absence of dissipation in the plasma such as viscosity, it is expected that the flow cannot stabilize the system. Both predictions forbid the experimentally demonstrated long-lasting wall stabilization of the tokamak plasmas. In particular, they do not allow the rotational stabilization and the regimes with edge harmonic oscillations (EHOs) observed on the DIII-D tokamak. Besides, they cannot be reconciled with a number of theoretical studies on the plasma rotation effect on the stability. Situations when the results are incompatible with those theorems are not rare, but still remain unresolved compromising the conclusions of the both sides. The most known first theorem was published in 1971 (Nuclear Fusion, p. 259), but since then it has never been analyzed, confirmed or corrected by independent researchers. This task is addressed here. A missing chain of derivations is restored and earlier unknown limitations that restrict the applicability of the Pfirsch-Tasso theorems are established. Thereby, the disagreements with the models of the rotational stabilization are explained and shown to be amendable. Replacement of the Pfirsch-Tasso energy principle is proposed. The new result is free from the constraints implicitly imposed in the Pfirsch-Tasso proofs. It eliminates the contradictions and can be used with any plasma model (not necessarily ideal) and for arbitrary perturbations. The proposed extensions allow applications for the cases of practical interest such as feedback stabilization of RWMs, analysis of the rotational stabilization and optimization of the ITER scenarios. Examples are presented and consequences are discussed.
Speaker: Dr Vladimir Pustovitov (National Research Centre “Kurchatov Institute”)
• 08:30
Phase Locking, Phase Slips and Turbulence: A New Approach to Mechanisms for Quiescent H-Mode 4h
We demonstrate E × B shear governs the dynamics of the cross phase of the peeling-ballooning-(PB)mode-driven heat flux, and so determines the evolution from the edge-localized (ELMy) H mode to the quiescent (Q) H mode. A physics-based scaling of the E × B shearing rate for accessing the QH mode is predicted. The ELMy H mode to the QH-mode evolution is shown to follow from the conversion from a phase locked state to a phase slip state. In the phase locked state, PB modes are pumped continuously, so bursts occur. In the slip state, the PB activity is a coherent oscillation. Strong E × B shearing implies a higher phase slip frequency. PB turbulence can degrade slip coherency. This model predicts a new state of cross phase dynamics and gives a new understanding of the mechanism for ELMy to QH-mode evolution. This material is based upon work supported by the U.S. Department of Energy, Office of Science, Office of Fusion Energy Sciences, under Award Numbers DE-FG02-04ER54738 and DE-SC0008378.
Speaker: Dr ZHIBIN GUO (University of California, San Diego)
• 08:30
Physics of flux closure during plasmoid-mediated reconnection in Coaxial Helicity Injection 4h
In a low-aspect-ratio Spherical Torus (ST), and in particular in an ST-based fusion reactor, due to the restricted space for a central solenoid, elimination of the central solenoid, and thus non-inductive current-drive techniques, is necessary. Transient Coaxial Helicity Injection (CHI) is a leading candidate for plasma start-up and current formation in NSTX-U. In NSTX, transient CHI has generated over 200 kA of toroidal current on closed flux surfaces without the use of the conventional central solenoid. To correctly model current generation and to better understand the physics of CHI start-up, comprehensive resistive MHD simulations have been conducted for the NSTX and NSTX-U geometries. It has been shown that magnetic reconnection has a fundamental role in the plasma start up and current formation in NSTX/NSTX-U. Here, we report two major findings from these CHI simulations: 1) formation of an elongated Sweet-Parker (S-P) current sheet and a transition to plasmoid instability has for the first time been demonstrated by simulations of CHI experiments and 2) a large-volume flux closure, and large fraction conversion of injected open flux to closed flux in the NSTX-U geometry have also now been demonstrated for the first time. This material is based upon work supported by the U.S. Department of Energy, Office of Science, Office of Fusion Energy Sciences under Award Numbers DE-SC0010565 and DE-FG02-99ER54519.
Speaker: Dr Fatima Ebrahimi (Princeton University/PPPL)
• 08:30
Plasma Disruption and VDE modeling in support of ITER 4h
Accurate modeling of major disruption (MD) and vertical displacement events (VDEs) in ITER is necessary to determine the halo current amplitude during these events and hence the electromagnetic loads on the machine components. The modeling of these events were originally done by DINA code and the results were later validated by TSC simulations and they both agree remarkably well when similar code assumptions are made. However, in these simulations, the halo current amplitude depends critically on the choice of halo parameters, namely the temperature and width of the halo region, Due to lack of credible experimental data of these two parameters and also no any sound physics based model so far, these parameters are chosen rather ad-hoc. For validation simulations with existing experiments, these parameters, including their temporal profiles, are chosen carefully for each experimental discharge so as to give a good match between the experiments and simulations. But for predictive simulations for ITER, this creates a problem as to what parameters to be chosen. To resolve this issue, a concerted effort to validate the TSC model against a wider set of experiments in different machines are presently underway. We have selected a set of four shots each in DIII-D and CMOD which are simulated in TSC. The halo parameters are set carefully only for one experiment in each machine and for the rest of the shots, they are kept unchanged. Thus the difference between the experimental and simulated halo current amplitude in these discharges would give an indication of the possible error in predictive modeling. We have already modeled three DIII-D discharges and we can reproduce the halo currents within about 10% of their experimental value. More discharges are being simulated at present both in DIII-D and CMOD. Details of these simulations and their results will be presented in this paper. We shall also explore any possible scaling laws of the halo current amplitude on these two parameters in these discharges.
Speaker: Dr Indranil Bandyopadhyay (ITER-India, Institute for Plasma Research)
• 08:30
Plasma Effects in Full-Field MHD-Equilibrium Calculations for W7-X 4h
Wendelstein 7-X aims at quasi-steady-state operation to demonstrate the reactor-viability of stellarators optimized with respect to MHD-equilibrium and -stability, low neoclassical transport, small bootstrap current and good fast-particle confinement. To reach this goal an island divertor is foreseen for particle and energy exhaust, which utilizes the naturally occurring boundary islands connected with the appearance of low-order rational values of the rotational transform at the plasma boundary. The island separatrix thus bounds the plasma, and the strike lines of the island fans determine the heat load distribution on the divertor structures. Although the configuration of W7-X has been optimized to display a small impact of plasma currents on the configuration, these effects still persist and change the plasma shape and the boundary islands’ width and location. From previous studies it is known, for example, that with growing plasma-beta the island width increases, and the X- and O-point locations move poloidally, consistent with the effect of the Shafranov-shift. A net toroidal current is known to shift the island-generating resonance radially which, depending on the amount of plasma current, can lead to undesired deviations from proper island divertor operation, e.g. the shifting of the island structures away from the divertor plates resulting in a limiter magnetic configuration or in heat loads misdirected to critical components. The contribution presents and discusses an approach for full-field calculations based on the VMEC-EXTENDER code combination. The effect of plasma-beta and of net-toroidal currents on the width and location of the islands is investigated in configurations in which the bootstrap current is expected to be small enough (according to transport simulations) to allow high-performance, quasi-steady-state operation compatible with the island divertor. The calculated fields will be compared to calculations using the 3D MHD-equilibrium code HINT, whose numerical scheme does not rely on the existence of flux surfaces and allows the self-consistent treatment of islands and stochastic regions. The differences resulting from the two approaches will be discussed.
Speaker: Dr Joachim Geiger (Max-Planck-Institute for Plasma Physics, Greifswald, Germany)
• 08:30
Pressure Driven Currents Near Magnetic Islands in 3D MHD Equilibria: Effects of Pressure Variation Within Flux Surfaces and of Symmetry 4h
In toroidal MHD equilibria, pressure can generally be regarded as constant on the flux surfaces. The regions near small magnetic islands, and those near the X-lines of larger islands, are exceptions. We show that the variation of the pressure within the flux surfaces in those regions has significant consequences for the pressure driven current. We further show that the consequences are strongly affected by the symmetry of the magnetic field if the field is invariant under combined reflection in the poloidal and toroidal angles. (“stellarator symmetry”.) In non-stellarator-symmetric equilibria, the pressure-driven currents have logarithmic singularities at the X-lines. In stellarator-symmetric MHD equilibria, the singular components of the pressure-driven currents vanish. In contrast, in equilibria having p constant on the flux surfaces the singular components of the pressure-driven currents vanish regardless of the symmetry. In 3D MHD equilibria having simply nested flux surfaces, the pressure-driven current goes like 1/x near a rational surface, where x is the distance from the rational surface. To calculate the pressure-driven current near a magnetic island, we work with a closed subset of the MHD equilibrium equations that involves only perpendicular force balance, and is decoupled from parallel force balance. Two approaches are pursued to solve our equations for the pressure driven currents. First, the equilibrium equations are applied to an analytically tractable magnetic field with an island, obtaining explicit expressions for the rotational transform and magnetic coordinates, and for the pressure-driven current and its limiting behavior near the X-line. The second approach utilizes an expansion about the X-line to provide a more general calculation of the pressure-driven current near an X-line and of the rotational transform near a separatrix. The calculations described here are motivated, in part, by tokamak experiments where significant differences are observed between the behavior of stellarator-symmetric and non-stellarator-symmetric configurations with regard to stabilization of edge localized modes (ELMs) by resonant magnetic perturbations (RMPs). Implications for the coupling between neoclassical tearing modes (NTMs), and for magnetic island stability calculations, are also discussed. Supported by DOE contract DE-AC02-09CH11466.
Speaker: Allan Reiman (Princeton Plasma Physics Lab)
• 08:30
Role of explosive instabilities in high-beta disruptions in tokamaks 4h
Explosive growth of a ballooning finger is demonstrated in nonlinear magnetohydrodynamic calculations of high-beta disruptions in tokamaks. The explosive finger is formed by an ideally unstable n=1 mode, dominated by an m/n=2/1 component. The quadrupole geometry of the 2/1 perturbed pressure field provides a generic mechanism for the formation of the initial ballooning finger and its subsequent transition from exponential to explosive growth, without relying on secondary processes. The explosive ejection of the hot plasma from the core and stochastization of the magnetic field occur in Alfvenic time scales, accounting for the extremely fast growth of the precursor oscillations and the rapidity of the thermal quench in some high-beta disruptions.
Speaker: Dr Ahmet Aydemir (National Fusion Research Institute, Daejeon, Korea)
• 08:30
Securing high beta_N JT-60SA operational space by MHD stability and active control modelling 4h
A careful numerical evaluation of MHD stability and of active control strategies is of paramount importance to reach one of the main goals of JT-60SA (Super Advanced) device, namely the development and qualification of high beta_N, steady-state regimes for future reactors like DEMO. Thanks to its powerful and flexible additional systems for heating and current drive, to its shaping capabilities and to several actuators for different kinds of real-time plasma control, JT-60SA aims at studying plasmas exceeding both the threshold for neoclassical tearing mode (NTM) destabilization and the so called Troyon no-wall beta limit for external kink instabilities. This work reports on the latest results on key issues in MHD stability and control of JT-60SA advanced tokamak plasmas, with particular reference to Neoclassical Tearing modes (NTM) and Resistive Wall Mode (RWM) physics. The amplitude evolution of NTM instabilities in the reference high beta_N scenarios is investigated by numerical tools developed in the framework of the European Integrated Tokamak Modelling effort. By solving the Generalized Rutherford Equation the role of different effects (such as bootstrap, curvature and polarization) is evaluated, including mode frequency evolution. Active NTM stabilization techniques are explored by modeling the action of the dual frequency (110 GHz and 138 GHz) electron cyclotron system. JT-60SA steady state scenarios present also new challenges for RWM stability studies given their targets in terms of beta_N (~4) and bootstrap current fraction (~70%). A further issue is given by the presence of a population of fast particles generated by high-power, high energy (10 MW at 500 keV) negative neutral beam injection system. The 2D stability code MARS-F/K is used to study plasma flow and drift kinetics stabilizing effects. Wall stabilization effects are estimated by the CarMa code that couples the 2D plasma stability to a 3D description of the passive boundaries surrounding the plasma. Feedback control of RWMs as provided by a set of 18 active coils is studied by the self-consistent inclusion in the model of a representation of the control system producing an overall dynamic model cast in the state variable space.
Speaker: Tommaso BOLZONELLA (ItRFX)
• 08:30
Self-consistent optimization of neoclassical toroidal torque with anisotropic perturbed equilibrium in tokamaks 4h
Control of toroidal rotation is an important issue for tokamaks and ITER since the rotation and its shear can significantly modify plasma stability from microscopic to macroscopic scales. A potentially promising actuator for the rotation control is the non-axisymmetric (3D) magnetic perturbation, as it can substantially alter toroidal rotation by neoclassical toroidal viscosity (NTV). The optimization of the 3D field distribution for the NTV and rotation control is however a highly complicated task, since NTV is mostly non-linear to the magnitude of the applied field with a complex dependency on the 3D field distribution [1]. In this paper we present a new method that entirely redefines the optimizing process, using the new general perturbed equilibrium code (GPEC). GPEC solves a non-self-adjoint force operator and balance with the first-order change in pressure anisotropy by non-axisymmetry, and integrates its second-order change for NTV under the force balance. This self-consistent calculation uniquely yields the torque response matrix function, and enables the NTV profile optimization by a single code run based on the full eigenmode structure of the matrix function. The code applications to non-axisymmetric control coil (NCC) design in NSTX-U demonstrated the efficiency and accuracy of the new method, and in addition showed the importance of the backward helicity modes and self-shielding by torque [2] in local NTV control. The access to the optimized field distribution is limited in practice by available coils, but it is also straightforward to couple the coils to torque matrix and optimize the current distributions in the coils, as has been actively studied in KSTAR, NSTX-U, and ITER. A number of other GPEC applications will also be discussed, including the verification and validation of high-β 3D plasma response [3] and kinetic stabilization with the self-consistent eigenfunctions. [1] S. Lazerson, J.-K. Park et al., Plasma Phys. Control. Fusion 57, 104001 (2015) [2] A. H. Boozer, Phys. Rev. Lett. 86, 5059 (2001) [3] Z. R. Wang, M. Lanctot, Y. Liu, J.-K. Park et al., Phys. Rev. Lett. 114, 145005 (2015) *This research was supported by U.S. DOE contracts #DE-AC02-09CH11466.
Speaker: Dr Jong-Kyu Park (Princeton Plasma Physics Laboratory)
• 08:30
Simulation study of interaction between runaway electron generation and resistive MHD modes over avalanche timescale 4h
Runaway electron (RE) generation after major disruptions is simulated over a full current quench (CQ) timescale, which covers both fast MHD events and slow RE avalanche amplification. A novel 3D RE analysis code EXTREM allows us to study (1) fast, global transport of REs with macroscopic MHD modes and (2) the RE generation triggered by electric fields induced owing to fast MHD dynamics (i.e., ‘mode-induced REs’). In the EXTREM code, slow CQ process (n = 0; n: the toroidal mode number) is described using the current diffusion model with a time varying resistivity, whereas RE transport and fast MHD dynamics (n > 0) are treated on the basis of reduced MHD models. The long-timescale simulations with the EXTREM code demonstrate their advantage for the analysis of net RE generation in a self-consistent manner with the anomalous transport and generation mechanisms due to resistive MHD modes. Effects of the resistive mode on the spatial profiles of generated REs can also be compared with a 1D diffusion model. These extensions of the model is a valuable step towards self-consistent treatments of the RE generation with thermal collapse, which is a highly complex task but is an important challenge for predictive simulations of the RE generation during mitigated disruptions in ITER. In this paper, a particular attention has been paid to m = 1 resistive modes triggered with the current peaking resulting from massive RE generation (m: the poloidal mode number). It is shown that sawtooth-like events are triggered when the central safety factor q(0) drops below unity, and the burst of Dreicer electrons is induced as a return current that compensates the expelled poloidal magnetic flux. In view of the global energy balance, the return current plays a role in converting the potential energy of the MHD instability to RE kinetic energies. It is still a small perturbation on a fast MHD timescale but can be amplified over the avalanche timescale. The scope of the paper will also be extended to the impact of tearing modes.
Speaker: Dr Akinobu Matsuyama (Japan Atomic Energy Agency)
• 08:30
Simulations of Runaway Electron Generation including Hot-Tail Effect 4h
The suppression and mitigation of runaway electron (RE) is an urgent issue of large scale tokamak operation. The contribution of hot-tail effect, which arises from the fast thermal quench, is studied using Fokker-Planck simulation. It is found that if the thermal quench is fast enough to invoke the hot-tail effect, it may produce seed REs and enhance total RE current even in a high electron density plasma.
Speaker: Dr Hideo Nuga (Kyoto University)
• 08:30
Three-dimensional numerical analysis of interaction between plasma rotation and interchange modes 4h
Effects of the poloidal shear rotation on the magnetohydrodynamic (MHD) stability of interchange modes in a Large Helical Device (LHD) configuration are numerically studied. This simulation is the first three-dimensional (3D) full-MHD nonlinear analysis for heliotron plasmas including the flow. In LHD, the highest average beta value of 5.1% is successfully obtained in the configuration where the plasma is predicted to be unstable with respect to the Mercier criterion. Thus, some stabilizing effects work on the plasma and it is crucial to identify the key physics of this stabilization for not only the understanding of the LHD plasmas but also the accurate design of the helical DEMO. Recently, it is observed in the experiments that the magnetic perturbation grows rapidly just after the mode rotation stops and causes a partial collapse of the electron temperature. This phenomenon indicates that the plasma rotation may suppress the growth of the mode. Thus, we numerically study the rotation effects on the MHD stability against the interchange modes in the LHD plasmas. As the numerical procedure, we employ a static equilibrium and incorporate a model shear flow as the rotation in the initial perturbation of the stability calculation. The 3D numerical codes of HINT and MIPS are utilized for the equilibrium and the stability calculations, respectively. We apply this method to an LHD equilibrium that is unstable for the interchange mode. In the no flow case, the pressure profile collapses and the magnetic field lines becomes stochastic in the nonlinear saturation phase. When a flow of which the kinetic energy is much larger than the saturation level in the no flow case is applied initially, such pressure collapse and field line stochasticity are not seen. Hence, this simulation result shows that a large initially-applied poloidal flow can suppress the interchange modes. In this simulation, the flow needed for the suppression is much larger than the flow observed in LHD experiments. However, we expect that we can reduce the needed flow when we take it into account that the nonlinear relaxation of the mode continues in the beta ramp-up phase. Therefore, the plasma rotation is considered as one of the candidates of the stabilization mechanism in LHD.
Speaker: Prof. Katsuji ICHIGUCHI (National Institute for Fusion Science)
• 08:30
Toroidal gyrokinetic studies of the tearing mode in tokamak plasmas 4h
Our present understanding of the physics of the tearing mode (TM) still does not allow a quantitative prediction of TM evolution in fusion reactors. The early phase of a TM, in particular, is determined by a complex interplay of different processes. We investigate the physics of the TM via gyrokinetic (GK) simulations in toroidal geometry using the code GKW. Two routes are followed, namely simulating the response of the plasma to a prescribed magnetic island and addressing the complete problem of the growth of the TM in the presence of GK turbulence. In simulations with prescribed magnetic perturbation, it is found that the density profile inside an island whose width does not exceed significantly the ion orbit width follows an adiabatic law (flattened for islands rotating at the ion diamagnetic frequency, unperturbed at the electron diamagnetic frequency) also in the presence of ITG turbulence, although the physics becomes more complex. The bootstrap current flowing in an island of this size is hence a function of the rotation frequency. Simulations performed to isolate the role of the island electric field show that for frequencies of the order of the parallel ion streaming the perpendicular fluxes exhibit a linear scaling with the island frequency omega and a complex radial pattern, confirming drift-kinetic results. A polarization signal (quadratic with omega) can be identified only in a narrow range at higher frequencies. In linear, self-consistent simulations, the TM rotates at the electron diamagnetic frequency at low collisionalities, but reverses direction at higher collisionality. The growth rate scales with 1/7-power of the resistivity in the semi-collisional regime. The growth of a TM embedded in GK electromagnetic turbulence shows, as in other (fluid) studies, that the turbulent fluctuations provide a seed for the magnetic island. They drive its growth at a rate significantly faster than the linear tearing growth rate. Depending on the value of the plasma beta, the subsequent evolution exhibits a Rutherford behaviour largely independent of the turbulence, or a disruption of the Rutherford phase, with the TM growing at its linear growth rate even if the island width exceeds the singular layer width. The island rotation is also modified by the presence of the turbulence. Generally, the mode rotation slows as the island grows.
Speaker: Dr Emanuele Poli (Max-Planck-Institute for Plasma Physics, Germany)
• 08:30
Towards powerful negative ion beams at the test facility ELISE for the ITER and DEMO NBI system 4h
The negative ion source test facility ELISE represents an important step in the European R&D roadmap towards the neutral beam injection (NBI) systems at ITER. ELISE provides early experience with operation of large RF-driven negative hydrogen ion sources. Its source area is 1x0.9 m2 and the net extraction area of 0.1 m2, formed by 640 apertures, corresponds to a half-size ITER source. The test facility aims at demonstrating large-scale extraction and acceleration of negative hydrogen ions (H‾, D‾) for pulses of up to 1 h with half the current required on ITER. Additionally, the ratio of co-extracted electrons to ions must be kept below one, which is quite demanding in particular for deuterium operation. Starting with first plasma pulses in March 2013, ELISE has meanwhile demonstrated stable 1 h plasma discharges in hydrogen with repetitive 10 s extraction every 3 min with 9.3 A extracted current and an electron-to-ion ratio of 0.4 at the pressure required by ITER of 0.3 Pa but using only one quarter of the available RF power. At half of the available RF power a stable 400 s plasma discharge was achieved with 18.3 A beam pulses at an electron-to-ion ratio of 0.7. Linear scaling towards full RF power predicts that the target value of the negative ion current can be achieved or even exceeded. Issues in long pulse operation are the caesium dynamics and the stability of the co-extracted electron current. Newly developed magnetic filter field configurations allowed achieving for the first time 1 h pulses in deuterium with an electron-to-ion ratio below one, however only at a quarter of the available RF power. Advanced beam diagnostics such as beam emission spectroscopy and a sophisticated diagnostic calorimeter reveal that the requirement on the uniformity of these large beams (deviations < 10%) can be met. For a DEMO fusion reactor, the requirements of a heating and current drive system will strongly depend on the DEMO scenario and are presently assessed within EUROfusion WPHCD. As NBI systems based on negative ions are regarded as one candidate, ELISE could serve in a later stage as a test bed for concepts concerning RF efficiency, operation without caesium or with largely reduced caesium consumption, and neutralization by a laser neutralizer in order to improve efficiency and reliability. IPP’s present small scale experiments show promising results.
Speaker: Prof. Ursel Fantz (Max-Planck-Institut fuer Plasmaphysik)
• 08:30
Two-fluid sub-grid-scale viscosity in nonlinear simulation of ballooning modes in a heliotron device 4h
Nonlinear growth of ballooning modes in a heliotron device is studied by means of two-fluid numerical simulations. A model to substitute an influence of the scales smaller than the grid size, Sub-Grid-Scale (SGS), on the scales larger than the Grid Scale (GS) is introduced. A simulation with the SGS model, a Large Eddy Simulation (LES), of two-fluid MHD model successfully shows growth of the ballooning modes with a diamagnetic flow and nonlinear saturation, showing usefulness of the LES approach. In order to enable two-fluid simulations of ballooning modes and clarify saturation mechanism of the modes in a heliotron device, we focus on influences of the SGS modes truncated because of the finite numerical resolution, instead of adopting an unphysically large viscosity. Since the truncation often contaminates nonlinear dynamics of the GS modes, compensating the influences of the SGS modes to the GS modes by a SGS model is essential. The SGS terms in two-fluid momentum equations can be modeled by the SGS viscosity and the resistivity which are composed of the rate of strain tensor, the fluctuation part of current density and two model constants. Firstly the two model constants are calibrated by comparing a direct numerical simulation and LESes of homogeneous magnetized Hall MHD turbulence. A full 3D simulation of the ballooning modes in Large Helical Device with a small viscosity shows that the the SGS viscosity can become locally considerably large. Secondly, it is shown in full-3D two-fluid MHD simulations that a diamagnetic flow generated by the two-fluid term is coupled with the ballooning modes and restricts the instability to relatively low modes. By the use of the SGS model, our LESes are carried out without numerical instability even though high modes grow in linear phase, and the computational cost is reduced to about 1/64 of a precise simulation with a large number of grid points. In summary, our two-fluid MHD LESes achieve a nonlinear saturation of ballooning modes with a small viscosity. The LES approach enables a drastic reduction of the computational cost and better representation of dynamics in two-fluid simulations. The SGS-modeling in the pressure equation is left for future. Further results together with related subjects such as slab Rayleigh-Taylor instability will also be reported.
Speaker: Dr Hideaki Miura (National Institute for Fusion Science)
• 08:30 12:30
Poster FIP/1
• 08:30
ITER Central Solenoid Module Fabrication 4h
The fabrication of the modules for the ITER Central Solenoid (CS) has started in a dedicated production facility located in Poway, California, USA. The necessary tools have been designed, built, installed and tested in the facility to enable the start of production. The current schedule has first module fabrication completed in 2017, followed by testing and subsequent shipment to ITER. The Central Solenoid is a key component of the ITER tokamak providing the inductive voltage to initiate and sustain the plasma current and to position and shape the plasma. The design of the CS has been a collaborative effort between the US ITER Project Office (US ITER), the international ITER Organization (IO) and General Atomics (GA). GA’s responsibility includes completing the fabrication design, developing and qualifying the fabrication processes and tools, and then completing the fabrication of the seven 110 tonne CS modules. The modules will be shipped separately to the ITER site, stacked and aligned in the Assembly Hall prior to insertion in the core of the ITER tokamak. A dedicated facility in Poway, California, USA has been established by GA to complete the fabrication of the seven modules. Infrastructure improvements included thick reinforced concrete floors, a diesel generator for backup power along with cranes for moving the tooling within the facility. The fabrication process for a single module requires approximately 22 months followed by 5 months of testing, which includes preliminary electrical testing followed by high current (48.5kA) tests at 4.7K. The production of the 7 modules is completed in a parallel fashion through ten process stations. The process stations have been designed and built with most stations having completed testing and qualification for carrying out the required fabrication processes. The final qualification step for each process station is achieved by the successful production of a prototype coil. Fabrication of the first ITER module is in progress. The seven modules will be individually shipped to Cadarache upon their completion. This paper describes the processes and status of the fabrication of the CS Modules for ITER.
Speaker: Mr John Smith (General Atomics)
• 08:30
Progress in High Power Test of R&D Source for ITER ICRF system 4h
The IC H&CD system is one of the major tools for achieving the plasma performances foreseen in ITER's operation scenario. This system is designed to provide 20 MW into the plasma, at frequencies included in the band 40 MHz to 55 MHz. For ensuring 20 MW power availability for plasma operation, 24 MW is required at the output of the RF sources. India is responsible to deliver nine numbers of RF Sources to ITER system. Each source shall have the power handling capability of 2.5 MW/CW at VSWR 2:1 in the frequency range 35 – 65 MHz or 3.0 MW/CW at VSWR 1.5:1 in the frequency range 40 – 55 MHz, along with other stringent requirement. An urgent need for pre-qualification of final stage tube and few critical components is established through R&D program to bridge the significant gap between demonstrated capability of RF source system at various worldwide fusion facilities vs. ITER need, in terms of power level, pulse duration & bandwidth (BW) requirement. In 2012, ITER-India has signed contract with Thales Electron Devices (France) for establishing the technology in very high power RF amplifiers, using Diacrode tube. The contract is to design & develop driver and final stage amplifiers. Tubes and cavities are integrated in full amplifier chain developed by ITER-India. To test the performance of the amplifier chains at matched and mis-matched load condition, high power Test Rig (3MW/CW capability) is developed at Indian test facility. After successful assembly & integration of RF amplifier at Indian test facility, high power RF tests initiated. The objective for the tests is to confirm 1.5 MW output at 35 - 65 MHz for 2000 seconds with 1MHz BW (at 1dB point) over central frequency and to check the reliability of both the tube and the amplifier with a mismatched load (up to VSWR 2:1) which simulates power transmission to an antenna coupled to the plasma. This paper reports successful commissioning of RF amplifier and the achievement of 1.5 MW of RF power for more than 2000s, confirming other extremely challenging specifications and describes the operating scenarios, dissipation limit, safety system and various infrastructures developed at Indian test facility to support such operation.
Speaker: Mrs Aparajita Mukherjee (Institute for Plasma Research)
• 08:30
Progress of Experimental Study on Negative Hydrogen Ion Production and Extraction 4h
Development of the high performance negative hydrogen ion source is a fundamental demand in realizing fusion reactor. In order to clarify the extraction mechanism of H^-, temporal and spatial variations of the negative ions and electrons in the extraction region are intensively surveyed at NIFS. In addition, the beam acceleration experiments have been performed by changing the accelerator configuration in order to improve the voltage holding capability and to study the negative ion beam optics. In a cesiated hydrogen plasma, it was observed that the negative hydrogen ion density (n_H-) becomes one order higher in magnitude than the electron density (n_e) in the vicinity of the plasma grid (PG). The response of negative-ion rich plasmas to the extraction field was investigated by measuring the plasma potential (V_p) profiles in the axis perpendicular to the PG before and during beam extraction. The V_p increases with applying the extraction field, and the influence of the extraction field on the V_p was observed at 30 mm from the PG. As for the n_H-, it was also observed that the extraction field affects on the n_H- at 30 mm from the PG, where the n_H- decreases simultaneously with the beam extraction. These observations indicate that the extraction field affects the particle dynamics in the wide region extending over 30 mm from the PG. This feature is completely different from that of electron-ion plasmas. We also found that the negative ion production efficiency becomes twice higher by changing the shape of grounded grid holes without any modification on the plasma chamber. We assumed that the behavior of the back-streaming ion was affected by changing the GG. The back-streaming ion trajectory was analyzed with the beam trajectory simulation, and it was found that the back-streaming ion distributes in the larger area on the back plate with the slot GG. This implies that some part of condensed Cs on the back plate was more effectively evaporated by the back-streaming ion with the slot GG and that Cs flux onto the PG surface increased, which resulted in the enhancement of the negative ion production. This result suggests that the accelerator configuration is one of the key factors to determine the negative ion production efficiency and that the Cs consumption can be reduced by the Cs recycling from the wall of the ion source chamber.
Speaker: Dr Masashi Kisaki (National Institute for Fusion Science)
• 08:30
Recent Progress of ITER Package in ASIPP 4h
ASIPP has taken the responsibility of most CN ITER package. All packages follow current ITER schedule. The superconducting conductor package consists of 106 conductors with 6 kinds includes 7.5% TF conductor,totalPF conductor from PF2 to PF5, total CC conductor, and MB and CB conductor of feeder. Now total CN TF conductor package has been completed in production,and delivered to IO; completed the production and acceptance test of CC and feeder conductor package; completedthe production and acceptance testof 35 PF conductors and 4 dummy conductors. The ITER feeder system consists of 31 units. They convey power and coolant to magnets, and hold the numerous instrumentation channelswith the functioning of magnets system operationand monitoring. Now ASIPP has completed all qualification work and started manufacturing after PF4 CFT MRA meeting. The ITER Correction Coils (CC) consists of three sets of six coils each. Each pair of coils located on opposite sides with respect to the plasma is series connected with polarity such to produce asymmetric fields. TheCC PA was signed between IO and CN DAin 2010, now ASIPP has developed the manufacturing process including winding process, VPI technology, laser beam welding, helium inlet/outlet welding technology, and production qualification process is still on going. CN power supply package consists of PF AC/DC converter,reactive power compensationandharmonic filter (RPC&HF), and pulsed power electrical network materials (PPEN). ASIPP takes responsibility of key technology R&D, all kind test, integration and technical support. ASIPP has completed AC/DC converter and RPC&HF prototype test and integration test, and started its manufacturing since 2015. Now first PF AC/DC converter unit manufacturing has been completedandqualified by IO, 15 kinds of PPEN equipment has been delivered to IO site. ASIPP has two ITER diagnostic procurements, #12 horizontal port plug and radial X-ray camera(RXC).Port integration group has organized a system integration review meeting in July 2015 with mostly general issues be resolved or analyzed, most model clashes between tenants have been resolved. RXC's structure design is optimized and installation process is studied considering the simplification and easiness of maintenance. Remote handling skills and tools are designed for the system maintenance after being activated.
Speaker: Prof. Peng FU (Institute of Plasma Physics, Chinese Academy of Sciences)
• 10:15 10:45
Coffee Break 30m
• 10:45 12:30
3D Physics: EX/1 & TH/1
Convener: Prof. Saskia Mordijck (The College of William and Mary)
• 10:45
Role of MHD dynamo in the formation of 3D equilibria in fusion plasmas 20m
This work investigates the formation of helical core equilibria in toroidal fusion plasmas, focusing on the role of dynamo, or magnetic flux pumping mechanisms in determining the equilibrium current profile. Dynamo effects determine the safety factor profile of the final 3D equilibrium, with important consequences on MHD stability and transport. We compare experimental results from multiple machines (RFX-mod, MST, AUG, DIII-D) and nonlinear MHD modelling. Two paradigmatic cases of helical state formation are considered and common physics is identified, by direct measurements of dynamo effects and MHD simulations: spontaneous formation in high-current reversed-field pinch (RFP) plasmas [1] and the hybrid scenario in high-beta tokamak plasmas [2]. Helical cores form in both cases, either spontaneously via saturation of MHD modes, or due to the marginally-stable ideal MHD response to external 3D fields. Direct measurements of the dynamo emf associated to 3D plasma distortions will be presented for a database of helical RFP plasmas from RFX-mod and MST, covering a wide range of plasma parameters. Similar measurements were also done in helical states forming in response to external 3D fields in Ohmic RFX-mod tokamak plasmas and in DIII-D high-beta hybrid plasmas. Experimental results qualitatively agree with nonlinear MHD modelling performed with the codes SpeCyl [3], PIXIE3D [4], and NIMROD [5]. They indicate that central current is redistributed by a dominantly electrostatic MHD dynamo. The underlying physics common to RFP and tokamak is thus revealed: a helical core displacement modulates parallel current density along flux tubes, which requires a helical electrostatic potential to build up, giving rise to a helical dynamo flow. Similar results were also recently obtained with the M3D-C1 code [6]. [1] R. Lorenzini et al., Nature Phys. 5, 570 (2009); [2] T.C. Luce et al., Nucl. Fusion 54, 013015 (2014); [3] D. Bonfiglio et al., Phys. Rev. Lett. 94, 145001 (2005); [4] D. Bonfiglio et al., Plasma Phys. Control. Fusion 57, 044001 (2015); [5] J.R. King, C.R. Sovinec, V.V. Mirnov, Phys. Plasmas 19, 055905 (2012); [6] S.C. Jardin et al., Phys. Rev. Lett. 115, 215001 (2015).
Speaker: Paolo Piovesan (Consorzio RFX)
• 11:05
Optimization of the Plasma Response for the Control of Edge-Localized Modes with 3D Fields 20m
Measurements and modeling of the plasma response to applied 3D magnetic perturbations – specifically its dependence on collisionality, beta, and rotation – yield new insight into the physics of edge-localized mode (ELM) control and better define the criteria needed to achieve ELM suppression in ITER. ELM control depends on the coupling of the applied field to a stable edge mode that drives resonant fields on edge rational surfaces and is directly observed on high-field side (HFS) magnetic sensors. The edge mode amplitude is inversely proportional to pedestal collisionality yet is insensitive to global beta, consistent with a current-driven mode as opposed to a pressure-driven kink, and reinforcing the importance of ITER-like collisionality to resonant field drive [1]. Advances in ideal MHD modeling have identified highly stable, beta-independent plasma response modes that nonetheless drive strong resonant fields – showing a path to ELM control with minimal impact on global stability [2,3]. Onset of ELM-suppression is consistent with a transport bifurcation driven by the penetration of resonant fields, evidenced by sudden changes in: boundary heat flux, visible helical striations, pedestal-top rotation and fluctuations, and HFS magnetic effects [4]. Systematic torque and beta scans reveal a loss of ELM suppression consistent with two-fluid modeling predictions of a reduction in the penetrated field [5]. These results further the development of quantitative models for the conditions necessary to achieve ELM suppression, emphasizing both edge mode coupling for resonant field drive, and low pedestal-top electron rotation for resonant field penetration. Optimization of the equilibrium conditions and discharge evolution, in addition to the applied field structure, will be required to successfully achieve ELM suppression in ITER. [1] C. Paz-Soldan et al, Nucl. Fusion 2016 (in press) [2] C. Paz-Soldan et al, Phys. Rev. Lett. 114, 105001 (2015) [3] N. Logan et al, Phys. Plasmas 2016 (in review) [4] R. Nazikian et al, Phys. Rev. Lett. 114, 105002 (2015) [5] R. Moyer et al, Nucl. Fusion 2016 (in preparation)
Speaker: Dr Carlos Paz-Soldan (Oak Ridge Institute for Science Education)
• 11:25
Penetration and amplification of resonant perturbations in 3D ideal-MHD equilibria 20m
The nature of ideal-MHD equilibria in three-dimensional geometry is profoundly affected by resonant surfaces, which beget a non-analytic dependence of the equilibrium on the boundary. Furthermore, non-physical currents arise in equilibria with continuously-nested magnetic surfaces and smooth pressure and rotational-transform profiles. We demonstrate that three-dimensional, ideal-MHD equilibria with nested surfaces and delta-function current-densities that produce a discontinuous rotational-transform are well defined and can be computed both perturbatively and using fully-nonlinear equilibrium calculations. The results are of direct practical importance: we predict that resonant magnetic perturbations penetrate past the rational surface (i.e. shielding'' is incomplete, even in purely ideal-MHD) and that the perturbation is amplified by plasma pressure, increasingly so as stability limits are approached.
Speaker: Dr Stuart Hudson (Princeton Plasma Physics Laboratory)
• 11:45
Enhanced understanding of non‐axisymmetric intrinsic and controlled field impacts in tokamaks 20m
An extensive study of intrinsic and controlled non-axisymmetric field impacts in KSTAR has enhanced the understanding about non-axisymmetric field physics and its implications, as well as demonstrating the importance of optimal 3-D configurations in resonant magnetic perturbation (RMP)-driven control on edge localized modes (ELMs) in tokamaks. The $n=1$ intrinsic non-axisymmetric field was measured to remain as low as ${\langle\delta B/B_0\rangle}_{m/n=2/1} \sim 4\times 10^{-5}$ at high-beta plasmas ($\beta_N\sim2$), which corresponds to approximately 20% below the targeted ITER tolerance level. A systematic survey of $n=1$ controlled resonant field has revealed that KSTAR has a lower power threshold for L-H transition (at least 10 %) than DIII-D (configured with $n=3$ RMP) with similar plasma densities of $n_e=(2 -2.6)\times 10^{19} m^{-3}$, possibly benefiting from a low level of intrinsic error field and toroidal field ripple. As for the RMP ELM control, a high-quality $n=1$ RMP ELM suppression (duration of $\sim40 \tau_E$) was achieved using an operationally ‘reproducible’ approach. Throughout this investigation, we diagnosed edge activities using 3‐D ECE imaging diagnostics (ECEI) on both high-field-side (HFS) and low-field-side (LFS) simultaneously for the first time. According to ECEIs, the RMP ELM suppression was full of lively edge activities, which appears quite challenging to a prevailing theory that ‘peeling‐ballooning’ stability boundary is crossed from unstable to stable regimes due to RMP. While exploring the most favorable 3-D configuration ($n=1$, +90 deg. phasing), we discovered that midplane IVCC coils played a major role in mitigating the ELMs, while two off-midplane IVCCs ($n=1$ odd-parity) appeared insignificant on ELMy behavior change. In contrast, when the off-midplane IVCCs are configured with n=1 even-parity, strong plasma response was observed, even triggering mode-locking at high RMP currents. Considering that the ITER RMP coils are composed of 3-rows, just like in KSTAR, further 3-D physics study in KSTAR is expected to help us minimize the uncertainties of the ITER RMP coils, as well as establish an optimal 3-D configuration for ITER and beyond.
Speaker: Dr Yongkyoon In (National Fusion Research Institute)
• 12:05
Enhancement of helium exhaust by resonant magnetic perturbation fields 20m
Exhaust of helium as a fusion born plasma impurity is a critical requirement for future burning plasmas. We demonstrate in this paper that resonant magnetic perturbation (RMP) fields can be used to actively improve helium exhaust features. We present results from the TEXTOR tokamak with a pumped limiter and from the LHD heliotron with the closed helical divertor. The results show an important additional functionality of the ITER RMP ELM control coils and dedicated experiments on present day devices like DIII-D, EAST or KSTAR which obtained full ELM suppression by RMP field application are motivated. In both devices RMP fields are applied to generate a magnetic island located in the very plasma edge and this magnetic island has a noticeable impact on the helium exhaust. At the TEXTOR tokamak, the effective helium confinement time τp,He* is reduced by up to 43% and the actual reduction depends on the coupling of the magnetic island to the pump device. The LHD heliotron device, in contrast, features intrinsically a 3-D boundary and the closed helical divertor was designed for optimal pumping in this geometry. Without RMP field applied, τp,He* is a factor of ~4 higher for LHD compared to TEXTOR discharges in a comparable plasma density range. Ion root transport - one out of several different impurity transport regimes at LHD - is the most likely inward transport driver causing the high τp,He*. When a magnetic island is seeded into the intrinsic edge stochastic layer, a decrease of τp,He* by up to 30% and hence values closer to the tokamak situation are established. This shows that RMP fields are a fine-tuning actuator for the exhaust of helium, which is an attractive additional functionality for the ITER ELM control coils. 3-D fluid plasma edge transport and kinetic neutral gas modeling with the EMC3-EIRENE code shows for LHD that the actual helium concentration in the plasma core is dominated by wall recycling of helium. This points out that the back-fueling of the plasma by helium emitted from the plasma and recycled at the wall elements needs to be controlled. The edge magnetic island induced is shown to be an effective actuator to retain the recycled helium in the plasma periphery where it can be pumped away.
Speaker: Dr Oliver Schmitz (University of Wisconsin - Madison, Department of Engineering Physics)
• 12:30 14:00
Lunch Break 1h 30m
• 14:00 16:10
Overview 4: Inertial & Magnetic Fusion: OV/4
Convener: Prof. Dennis Whyte (MIT Plasma Science Fusion Center)
• 14:00
Overview of the IFMIF/EVEDA Project 25m
IFMIF, the International Fusion Materials Irradiation Facility, presently in its Engineering Validation and Engineering Design Activities (EVEDA) phase under the Broader Approach Agreement, will allow accelerated testing of structural materials with fusion relevant neutrons at >20dpa/year in 500cm3. IFMIF consists of two 125 mA and 40 MeV D+ linear accelerators operating in CW mode. The concurrent beam lines impact on a liquid lithium target with a 200mm  50mm beam cross section. The target consists of a 25mm 1mm thick liquid lithium screen flowing at 15m/s and 250°C channelled by a R250mm concave RAFM backplate. The suitable neutron flux generated in the forward direction will irradiate 12 test capsules housing around 1000 small specimens independently cooled with helium gas. The Engineering Design Activity (EDA) phase of IFMIF was successfully accomplished within the allocated time. The Engineering Validation Activity (EVA) phase has focused on validating the Accelerator Facility, the Target Facility and the Test Facility with the construction of various prototypes. The ELTL has successfully demonstrated the long term stability of a lithium flow under IFMIF nominal operational conditions with 25 days continuous operation in Oarai (JAEA) at 250°C and 15m/s within 1mm free surface fluctuations. A full-scale prototype of the High Flux Test Module has been successfully tested in the HELOKA loop (KIT Karlsruhe) demonstrating the feasibility of the uniformity in the temperature selected for the specimen set irradiated in each capsule. LIPAc, presently under installation and commissioning, will validate the concept of IFMIF Accelerators with a D+ beam of 125mA and 9MeV. The commissioning of the H+/D+ beams in Rokkasho Fusion Institute at 100keV was concluded early 2016; the commissioning of the 5MeV beam is to follow till early 2017. The 9MeV D+ beam will be achieved with a superconducting cryomodule during 2018. The realisation of a fusion relevant neutron source is a necessary step for the successful development of fusion. The stable progress achieved in this final EVEDA phase has ruled out technical concerns and potential showstoppers raised in the past. In the light of costs, which are unquestionably marginal to those of a fusion plant, a situation has emerged where soon steps towards constructing a Li(d,xn) fusion relevant neutron source could be taken.
Speaker: Dr Juan Knaster (IFMIF/EVEDA (F4E))
• 14:25
A Pathway to Laser Fusion Energy: Fast Ignition Realization EXperiment (FIREX) 25m
Here we report recent progress of the fast ignition inertial confinement fusion demonstration. Fraction of low energy (< 1 MeV) component of the relativistic electron beam (REB), which efficiently heats the fuel core, increases by the factor of 4 by enhancing pulse contrast of heating laser and removing preformed plasma sources. Kilo-tesla magnetic field is studied to guide the diverging REB to the fuel core. The transport simulation of the REB accelerated by the heating laser in the externally applied and compressed magnetic field indicates that the REB can be guided efficiently to the fuel core. The integrated simulation shows >4% of the heating efficiency and > 4 keV of ion temperature are achievable by using GEKKO-XII and LFEX, properly designed cone-fuel and the external magnetic field.
Speaker: Prof. Hiroshi AZECHI (Institute of Laser Engineering, Osaka University)
• 14:50
Overview of SST-1 Up-gradation & Recent Experiments in SST-1 & Overview of Recent Experimental Results from Aditya Tokamak 25m
A. Steady State Superconducting Tokamak (SST-1) is a operational’ experimental superconducting device since late 2013. Since last IAEA-FEC; SST-1 has been upgraded with Plasma Facing Components being installed and integrated in the vacuum vessel and is getting prepared towards long pulse operations in both circular and elongated configurations. The PFC integration has been completed in August 2015 and initial experiments have begun in SST-1 with circular plasma configurations. SST-1 offers a unique possibility of investigating long pulse discharges with large aspect ratio (> 5.5) compared to contemporary devices. Presently, SST-1 standard ohmic discharges are in excess of 100 KA with typical core density ~ 2 × 1019 m-3 and core electron temperatures ~ 500 eV having duration in excess of 300 ms. A 42 GHz ECR pre-ionization source at ~ 150 KW in 1.5 T central field breaks down the gas, the current starts up at ~ 1.3 MA/s in 60-80 ms in an induced field of ~ 0.3 V/m. These standard discharges demonstrate copious saw teething and MHD activities as the pulse progresses including NTM, mode locking and MHD characteristics. PFC equipped SST-1 has completed these basic experimental studies confirmed with simulations. These includes eddy currents influencing the NULL dynamics, field errors, equilibrium index evolutions, wall influencing plasma characteristics, plasma positions, plasma rotational and Tearing Mode characteristics including the island width and island growths etc. Presently, SST-1 is attempting at multi second long high aspect ratio plasma discharges by coupling the Lower Hybrid with the Ohmic plasma as well as with robust real time position and density controls. SST-1 device has been upgraded with a pair of internal coil aimed at effective fast plasma control and a pair of segmented coil aimed at controlling some of the rotational aspects of plasma including the RMPs and ELMs. Supersonic Molecular Beam Injection (SMBI) from both high field and low field sides and Pellets Injection Systems have also been added with several edge plasma diagnostics aimed at both density control and edge plasma turbulence studies. The up-gradation details including the planned ones, salient early plasma characteristics in large aspect ratio PFC equipped SST-1 plasma and future experimental plans towards long pulse operations in SST-1 will be elaborated in this paper. B. Several experiments, related to controlled thermonuclear fusion research and highly relevant for large size tokamaks including ITER, have been carried out in ADITYA, an ohmically heated circular limiter tokamak. Repeatable plasma discharges of maximum plasma current of ~ 160 kA and discharge duration beyond ~ 250 ms with plasma current flattop duration of ~ 140 ms has been obtained for the first time in ADITYA. The discharge reproducibility has been improved considerably with Lithium wall conditioning and improved plasma discharges are obtained by precisely controlling the plasma position. In these discharges, chord-averaged electron density ~ 3.0 – 4.0 x 10^19 m^-3 using multiple hydrogen gas puffs, electron temperature of the order of ~ 500 - 700 eV have been achieved. Novel experiments related to disruption control are carried out and disruptions, induced by hydrogen gas puffing are successfully mitigated using biased electrode and ICR pulse techniques. Runaway electrons are successfully mitigated by applying a short local vertical field (LVF) pulse. A thorough disruption database has been generated by identifying the different categories of disruption. Detailed analysis of several hundred disrupted discharges showed that the current quench time is inversely proportional to q_edge. Formation of current filaments are observed during most of the disruptions, which helps in identifying the cause of disruption. Apart from this, for volt-sec recovery during the plasma formation phase, low loop voltage start-up and current ramp-up experiments have been carried out using ECRH and ICRH. Successful recovery of volt-sec leads to achievement of longer plasma discharge durations. In order to achieve better coupling of lower hybrid waves to the plasma, multipl e gas puffs are injected prior to the launch of lower hybrid waves. The experiments showed considerable reduction in the reflection co-efficient indicating better absorption of LH waves in plasma. In addition to that Neon gas puff assisted radiative improved confinement mode has also been achieved in ADITYA. Further, the electrode biasing experiments have shown that during transition to better confinement mode, the Drift-Alfven fluctuations are suppressed and the current profile gets modified near the edge plasma region. In this paper, all the above mentioned experiments will be discussed.
Speaker: Dr Subrata Pradhan (Institute for Plasma Research, Bhat, Gandhinagar 382428, Gujarat, India)
• 15:15
Overview of Recent Experiments on HL-2A Tokamak 25m
Recent experiments on the HL-2A tokamak have been aimed at the major challenges relevant to ITER operation and fusion energy development. Significant progress has been achieved in many areas, including the first demonstration of high coupling efficiency of LHCD passive-active multi-junction (PAM) antenna in H-mode discharges, pedestal instability and dynamics, ITB formation mechanism, energetic particle physics, ELM and disruption mitigation, real-time control of tearing modes with ECRH, etc.. A new PAM antenna as an LHCD launcher was designed and installed on the HL-2A tokamak. A high coupling efficiency was demonstrated under NBI heated ELMy H-mode plasmas. This was the first time that PAM antenna was applied in H-mode. The effects of LHCD on ELM mitigation and control of heat load on divertor plate were also observed. It was found that impurity accumulation and relaxation in the edge could trigger a series of I-H-I transitions through the excitation of a broadband (50-150 kHz) electromagnetic (EM) turbulence. EM turbulence could also be excited by impurity injection via laser blow-off. An improved confinement with complete suppression of ELMs was achieved by this technique. These findings reveal the underlying physics of how impurity affects the pedestal evolution, and suggest an important method to actively control pedestal via impurity-excited EM turbulence. An inward particle flux induced by a quasi-coherent mode at frequency 40-60 kHz was found to be responsible for the dramatic changes of the gradients in pedestal and the triggering of ELMs. Dependence of the correlation of resistive ballooning modes and trapped electron modes on electron temperature increase was observed experimentally. Formation of the ion ITB was found to be closely related to the Te/Ti ratio. A new nonlocal transport phenomenon triggered by the fishbone was observed and demonstrated to be caused by electromagnetic fluctuations. High-frequency RSAE and resonant kinetic ballooning mode were confirmed in experiments, and found to cause energetic ions losses. Low-n Alfvenic ITG were observed and identified in Ohmic and NBI plasmas. For the first time non-resonant internal kink modes destabilized by energetic electrons with ECRH+ECCD were found in current ramp-up phases.
Speaker: Prof. Xuru Duan (Southwestern Institute of Physics)
• 15:40
Review of Recent Experiments on the T-10 Tokamak with All Metal Wall 25m
Review of the recent experimental results obtained on the T-10 tokamak is presented. To decrease the level of light impurities in 2015 both the rail and circular limiters were replaced with ones made of tungsten. The used tungsten type «POLEMA» as well as the technology of its soldering to the bronze substrate are similar to those applied for the production of the ITER divertor tiles. In the same time a movable lithium limiter was installed in the upper port. This limiter based on capillary-porous structure was made by JSC “Red Star”. With the tungsten limiter a considerable increase of the core radiation losses was obtained. Results on prevention of tungsten penetration in the core plasma by central ECRH and by insertion of the lithium limiter are presented in the paper. The efficiency of removal of heavier iron impurity depending on the discharge parameters and the power of the central ECRH was investigated. The maximal decrease of the heavier impurities concentrations is 5. Performed using the canonical profiles model analysis of the experiments on the density profile dynamic upon variation of the ECRH power showed that the density profile stiffness rises linearly with the heating power, while the peaking of the pressure profile in the core plasma asymptotically approaches to the canonical value. Using of bispectral analysis applied to the fluctuations of potential, density and poloidal magnetic field measured with heavy ion beam probe diagnostic showed an existence of three-wave interaction between GAM and broad-band turbulence. Also shown that the GAM amplitude declines with the mean density growth. The investigations of density fluctuation characteristics with correlation reflectometry confirmed a considerable decrease of the fluctuation amplitude together with disappearance of the quasicoherent modes on the inner side of the torus. Modeling showed that this effect can be, to a great extent, explained by nonlocality of reflectometry. Experiments with tangential X-ray detector indicated that abrupt restructuring of the low-m MHD modes and inward plasma shift during an energy quench are accompanied by bursts of the fast-scale (~0.5MHz) magnetic fields oscillations. Plasma discharge recovery after an energy quench is demonstrated in the T-10 high density plasma using ECRH auxiliary heating and controllable operation of the plasma current.
Speaker: Mr Dmitrii Sarychev (NRC "Kurchatov Institute")
• 14:00 18:45
Poster 2: P2
• 14:00
A Model of The Saturation of Coupled Electron and Ion Scale Gyrokinetic Turbulence 4h 45m
Two important regimes, observed in non-linear gyrokinetic turbulence simulations, are not well modeled by the TGLF quasilinear model. The first is the Dimits shift regime characterized by a non-linear upshift in the effective critical ion temperature gradient above the linear threshold. The second is the electron temperature gradient (ETG) streamer regime characterized by high electron scale turbulence when the ion scale turbulence is weak or stable. The Dimits shift impacts the predicted temperature profile in the deep core. The streamer regime is important when the temperature gradient of the electrons exceed that of the ions. A new model of the saturated turbulence spectrum will be shown to be able to match the turbulence driven transport fluxes in both of these regimes when applied to the TGLF quasilinear model. Analysis of the spectrum of the saturated electric potential fluctuations from multi-scale (both ion end electron scales) gyrokinetic turbulence simulations in tokamak geometry reveals that fluctuating zonal (axisymmetric) ExB flows couple the ion and electron scales. The zonal flows are driven by the ion scale instabilities but strongly regulate the amplitude of the electron scale turbulence. When the linear growthrate of the ETG modes exceeds the zonal flow mixing rate due to advection of the ETG modes, the electron scale turbulence can grow to large amplitude (streamer regime). The standard paradigm that the turbulence is saturated when the zonal flow shearing rate competes with linear growth cannot explain the saturation of the electron scale turbulence. Instead, it is the mixing rate of the zonal ExB velocity spectrum that competes with linear growth at both electron and ion scales. A model of the zonal flow mixing is shown to be able to capture the suppression of electron-scale turbulence by ion-scale turbulence and the threshold for the increase in electron scale turbulence when the ion-scale turbulence is reduced. The Dimits shift results from the impact of the zonal flow mixing on the ion scale turbulence amplitude. This work was supported by the US Department of Energy under DE-FG02-95ER54309, DE-FC02-04ER54698 and DE-SC0006957.
Speaker: Dr Gary M. Staebler (General Atomics)
• 14:00
A New Understanding of the Bootstrap Current in Steep Edge Pedestal and its Effect on the Pedestal Stability 4h 45m
Based on the kinetic simulations with the new gyrokinetic neoclassical code XGCa in realistic magnetic separatrix geometry, we developed an improved bootstrap current formula [R. Hager, C.S. Chang, submitted to Phys. Plasmas (2015)] that is much more accurate in steep edge pedestal plasma than the widely used formula by Sauter et al. [O. Sauter et al., Phys. Plasmas 6, 2834 (1999)] while being equally easy to use. The standard deviation of the Sauter formula from the XGCa result is about 24.8% while that from the new formula is only 5.4%. The XGCa-based bootstrap current formula is then applied to the electromagnetic stability analyses in the hybrid gyrokinetic XGC1 code, which uses gyrokinetic ions and fluid electrons [S. Ku et al., Phys. Plasmas 16, 056108 (2009)], together with a magnetic equilibrium code that takes the bootstrap current into account. The improved formula is suitable for applications that require fast and accurate calculation of the bootstrap current, and it incorporates finite orbit-width effects and other non-local physics that are introduced by a magnetic separatrix and strong ExB shearing rate. The new formula was necessary because existing studies of the bootstrap current are often based on assumptions that are valid in the core plasma but easily violated in the plasma edge, and the accuracy of these conventional predictions become questionable. Two significant findings from this XGCa study of the bootstrap current are the significant contribution of trapped electrons to the total current and the finite orbit-width effects that generally decrease the bootstrap current compared to the prediction from the conventional neoclassical theories and simulations.
Speaker: Dr Robert Hager (Princeton Plasma Physics Laboratory)
• 14:00
Alpha heating and isotopic mass scaling in JET DT plasmas 4h 45m
Experiments to detect alpha heating were performed in TFTR (1994) [1] and in JET (DTE1 1997 [2]. The TFTR results were claimed to be consistent with alpha particle heating of electrons. The JET results were claimed to show that alpha particle heating had been unambiguously observed. Recent papers [3,4] reanalyzed the alpha heating and other discharges from the JET DTE1 using improved TRANSP analysis. One result [3] is that although alpha-electron heating most likely was occurring, thermal hydrogenic isotopic mass effects could explain most of the effects attributed to alpha heating, and thus alpha heating was not clearly demonstrated. There are plans for new experiments in JET to investigate alpha heating and effects. ITER plans to study these during the DT phase after 2034. To help prepare for these, further analysis including more discharges from the JET DTE1 campaign are studied. Examples are pairs studied in [5]. Correlations of sawtooth delay times and core temperatures with fast ion parameters are presented. JET contributors are listed in the appendix of [6]. This work has been carried out within the framework of the EUROfusion Consortium, and has received funding from the Euratom research and training programme 2014-2018 under grant agreement No 633053. The views and opinions expressed herein do not necessarily reflect those of the European Commission. This work was also supported in part by the US DoE contract No. DE-ACO2-76-CHO3073. [1] Taylor G., et al., PRL 76 (1996) 2722 [2] Thomas P.R. et al., PRL 80 (1998) 5548 [3] Budny R.V., Nucl Fusion 56 (2016) 036013 [4] Budny R.V., et al., to appear in Nucl Fusion 2016 [5] Cordey J.G, et al., Nucl Fusion 39 (1999) 301 [6] Romanelli F., et al., IAEA Saint Petersburg, Russia 2014
Speaker: Dr Robert Budny (Princeton University)
• 14:00
Analysis of weakly coherent mode in I-mode with the BOUT++ code 4h 45m
The weakly coherent mode (WCM) in I-mode has been studied by six-field two-fluid model based on the Braginskii equations under the BOUT++ framework. The calculations indicate that a tokamak pedestal exhibiting a WCM is unstable to drift Alfven wave (DAW) instabilities and resistive ballooning mode. The nonlinear simulation shows promising agreement with the experimental measurements of WCM. The spectrum of the largest toroidal number mode n=20 at the location of the reflectometry agrees with the experimental data. The mode propagating in electron diamagnetic direction is consistent with the results from the magnetic probes, a large ratio of particle to heat diffusivity is consistent with the distinctive experimental feature of I-mode, and the value of the electron thermal diffusivity from simulation is almost as same as the effective thermal diffusivity from the experiment. The prediction of the WCM shows that free energy is mainly provided by the electron pressure gradient, which gives a well guidance for pursuing future I-mode studies.
Speaker: Dr Zixi Liu (Princeton Plasma Physics Laboratory / Institute of Plasma Physics, Chinese Academy of Sciences)
• 14:00
Anomalous and Neoclassical Transport of Hydrogen Isotope and Impurity Ions in LHD Plasmas 4h 45m
Gyrokinetic and drift kinetic simulations are carried out to investigate anomalous and neoclassical transport of hydrogen isotope and impurity ions in Large Helical Device (LHD) plasmas. Turbulent transport in high electron temperature regime, where the trapped electron mode (TEM) is dominant, is a critical issue for future burning plasmas. To clarify an impact of hydrogen isotope species on the turbulent transport in LHD system, TEM turbulence simulations in hydrogen and deuterium LHD plasmas with real-mass kinetic electrons have been carried out for the first time by gyrokinetic simulations with multi-species collision operator. The strong isotope dependence on the growth rate of collisionless TEM branch appears through the stabilization effect due to a mass dependence in the normalized electron-ion collision frequencies. Nonlinear simulations clarify the significant dependence of the isotope species in the reduction of the electron and ion energy fluxes in deuterium plasma. Here, stronger TEM stabilization in the deuterium plasma leads to the enhancement in the ratio of ZF energy to total energy. Transport in high ion temperature (Ti) plasmas with extremely hollow impurity density profiles (impurity hole) is also a critical issue hole phenomena is also a critical issue for high-performance. In high-Ti LHD plasma, the simulation indicates the neoclassical particle fluxes of electron and bulk ion species are outward directed, although the flux of the impurity carbon is extremely small and its value and direction are sensitive to the radial electric field. On the other hand, the microinstability analyses by gyrokinetic simulations show that the anomalous contributions of quasi-linear particle fluxes of all species are inward-directed which is consistent with the fact that the positive neoclassical particle fluxes and the negative turbulent fluxes should be balanced in a steady state. The ratio of each particle flux is almost consistent with the neoclassical contributions.
Speaker: Dr Masanori Nunami (National Institute for Fusion Science)
• 14:00
Core-edge coupled predictive modeling of JT-60SA high-beta steady-state plasma with impurity accumulation 4h 45m
The integrated modeling code TOPICS has been extended to couple impurity transports in core and scrape-off-layer / divertor regions, and applied to predictive modeling of JT-60SA high-beta steady-state plasma with the accumulation of impurity seeded to reduce divertor heat load. Consistent evaluation of impurity transport from the edge to the core clarified the compatibility of impurity seeding with the core plasma with high-beta (beta_N > 3.5) and full current drive condition, i.e., when the Ar seeding reduces the divertor heat load below 10 MW/m^2, its accumulation in the core is so moderate that the core plasma performance can be recovered by additional heating within the machine capability to compensate the Ar radiation. Validating anomalous heat transport models with JT-60U experiments and judging the applicability of models to the conservative prediction, which considers a lower bound of plasma performance, improved the above prediction reliability.
Speaker: Dr Nobuhiko Hayashi (Japan Atomic Energy Agency)
• 14:00
Crucial role of zonal flows and electromagnetic effects in ITER turbulence simulations near threshold 4h 45m
Development of a validated integrated modeling framework is a fundamental research task within the US fusion energy program. A primary component of this framework is an accurate transport model describing the small-scale, gradient-driven plasma microturbulence and its associated cross-field transport. Design and calibration of accurate transport models requires a database of well-converged nonlinear gyrokinetic simulations, which is a very computationally challenging undertaking. As early as 2008, GYRO simulations of ITER operating scenarios were observed to produce levels of nonlinear zonal-flow (ZF) activity large enough to quench turbulence inside the plasma core. This observation implied that modeling estimates of fusion power in ITER may have been pessimistic because turbulent transport was overestimated. The existence of flow-dominated, low-transport states persisted even as more accurate and comprehensive predictions of ITER profiles were made using the TGLF transport model. This was in stark contrast to GYRO-TGLF comparisons for modern-day tokamaks, for which GYRO and TGLF are typically in close agreement and transport is well-above threshold. It was speculated that closeness to threshold, ZF activity, and electromagnetic effects, could all play a key role in this discrepancy. Importantly, it became clear that TGLF must be generalized to include ZF stabilization for more accurate ITER simulations. The ongoing recalibration process also uncovered additional new concerns related to the accuracy of both the GYRO simulations as well as the TGLF saturation rule. Plaguing the workflow for the ZF recalibration effort is the intermittency of the near-threshold GYRO simulations. A second surprising result was that magnetic compression is strongly destabilizing in these plasmas, another effect not reproduced by TGLF. Exhaustive simulation and recalibration efforts are summarized, which represent an enormous undertaking carried out in stages over a period of years.
Speaker: Dr Jeff Candy (General Atomics)
• 14:00
Development of ITER Non-Activation Phase Operation Scenarios 4h 45m
Non-activation phase H/He operations in ITER will be important for commissioning of tokamak systems, such as diagnostics, heating and current drive (HCD) systems, coils and plasma control systems, and for validation of techniques necessary for establishing feasible operations. The assessment of feasible HCD schemes at various toroidal fields (2.65-5.3T) has revealed that the previously applied assumptions need to be refined for the ITER non-active phase H/He operations. A study on the ranges of plasma density and profile shape using the JINTRAC suite of codes has indicated that the hydrogen pellet fueling system should be carefully utilized in He operation to optimize IC power absorption, neutral beam shine-through density limit and H-mode access. The EPED estimation of the edge pedestal parameters has been extended to various H operation conditions, and the combined EPED and SOLPS estimation has provided a good guidance for modelling the edge pedestal in H/He operations. The availability of ITER HCD schemes, ranges of achievable plasma density and profile shape, and estimation of the edge pedestal parameters for H/He plasmas have been combined with the previous modelling efforts on studying the H-mode access and flat-top duration within the coil system constraints. Feasible ITER non-activation phase H/He operation scenarios have been developed by performing integrated time-dependent tokamak discharge simulations using CORSICA.
Speaker: Dr Sun Hee KIM (ITER Organization)
• 14:00
Direct identification of Predator-Prey dynamics in Gyrokinetic Simulations 4h 45m
The interaction between spontaneously formed zonal flows and small-scale turbulence in nonlinear gyrokinetic simulations is explored in a shearless closed field line geometry. It is found that when clear limit cycle oscillations prevail, the observed turbulent dynamics can be quantitatively captured by a simple Lotka-Volterra type predator-prey model. Fitting the time traces of full gyrokinetic simulations by such a reduced model allows extraction of the model coefficients. Among other findings, it was observed that the effective growth rates of turbulence (i.e. the prey) remain roughly constant, in spite of the higher and varying level of primary mode linear growth rates. The effective growth rate that was extracted corresponds roughly to the zonal-flow-modified primary mode growth rate. The result also demonstrates that the effective damping of zonal flows (i.e. the predator) in the parameter range, where clear predator-prey dynamics is observed (i.e. near marginal stability), agrees with the collisional damping expected in these simulations. This implies when the tertiary instability plays a role the dynamics becomes more complex than a simple Lotka-Volterra predator prey model.
Speaker: Sumire Kobayashi (FrLPP, FrCNRS)
• 14:00
EUROfusion Integrated Modelling (EU-IM) capabilities and selected physics applications 4h 45m
Recent developments and achievements of the EUROfusion Code Development for Integrated Modelling project (WPCD, follow-up of EFDA-ITM-TF), which aims at providing a validated integrated modelling suite for the simulation and prediction of complete plasma discharges in any tokamak, are presented. WPCD develops generic complex integrated simulations, workflows, for physics applications, using the standardized EU Integrated Modelling (EU-IM) framework. The integration of codes in EU-IM workflows is besides accompanied by a thorough cross-verification and, recently, by the introduction of rigorous release procedures. Among the achievements, the European Transport Simulator (ETS), has now reached a capability equivalent to the state-of-the-art integrated modeling transport codes, including interchangeable physics modules for equilibrium (both fixed and free boundary), transport (interpretative analytical, neoclassical, anomalous), impurities (all ionization states), NTM, sawteeth, pellets, neutrals, Heating and Current Drive (HCD) sources including all the heating schemes (EC, NBI, IC, nuclear) and synergy effects. The core ETS has been released and deployed at JET, offering a leading tool for both interpretive transport analysis and predictive modelling of complex scenarios. Selected physics applications are presented, in particular ETS simulations of plasma density control in reactor-scale plasmas fueled with multiple pellets. A MHD stability chain was developed for the analysis of equilibria from any tokamak in the EU-IM platform; it includes a pool of interoperable high-resolution equilibrium and linear MHD stability codes. Having passed a benchmark on core and global ideal kink instabilities, the chain has been released and applied to the predictive analysis of DEMO and JT60-SA scenarios and can be straightforwardly used for interpretive runs on present devices as JET and ASDEX Upgrade. A predictive J-alpha MHD pedestal stability analysis workflow has also been developed. Routine application to sensitivity analysis of DEMO1 scenarios is performed. Furthermore, a workflow including a turbulence code and a synthetic probe was developed and applied to investigate the turbulent transport in the edge and Scrape-Off-Layer (SOL) of ASDEX Upgrade. Finally, a prototype edge workflow integrating the interaction with PFC was demonstrated.
Speaker: Dr Gloria Falchetto (CEA)
• 14:00
Evaluation of Predictive Capability for Hydrogenic and Impurity Density in L- and H-mode Tokamak Plasma using Multimode Transport Model 4h 45m
Predictive capability of hydrogenic density and impurity density in L and H-mode plasma is strongly desirable to fully understand behaviors of plasma in tokamak, which can exhibit many modes of transports depending on the conditions of plasma. Combining many modes from turbulent transports, the Multi-Mode Model version 1995 (MMM95) includes coefficients from the Weiland model for the ion temperature gradient (ITG) and trapped electron modes (TEM), the Guzdar–Drake model for drift-resistive ballooning (RB) modes, and modified kinetic ballooning (KB) modes. In this work, the hydrogenic and impurity density profiles in L- and H-mode plasma are investigated using self-consistent modeling of BALDUR integrated predictive modeling code in which theory-based models are used. In these simulations, a combination of NCLASS neoclassical transport and MMM95 anomalous transport model is used to compute a core transport. The boundary conditions for temperature and density are taken to be at the top of the pedestal, where the pedestal values are taken from experiments. The predictive capability is determined by comparing the predicted profiles with experimental data in 24 discharges from various tokamaks and plasma conditions. Statistical analysis such as the average relative root mean square (RMS) deviation and offsets are used to quantify the agreement. The multi-parameters optimization technique is used to derive suitable coefficients for the MMM95 transport model. The simulation results show that even when the electron density and temperature profiles, and the ion temperature profiles agree well with experiments, yielding low RMS and correct trends, the impurity density profiles do not often agree with experiments, yielding much higher RMS and even opposite trends. The effects of KB and RB contribution are comparable on the impurity profiles. In addition, it is clear from the RMS optimization that a universal model with the same set of coefficients for all discharges is unlikely, but a range of each coefficient from each transport mode can be estimated for a given plasma regime.
Speaker: Dr Sujin Suwanna (Department of Physics, Faculty of Science, Mahidol University, Bangkok, Thailand)
• 14:00
Extending the Validation of Multi-Mode Model for Anomalous Transport to High Poloidal Beta DIII-D Discharges 4h 45m
The Multi-Mode Model (MMM7.1) for anomalous transport [1] is tested in predictive modeling of temperature profiles of high beta poloidal DIII-D discharges. This new H-mode plasma regime, with high beta poloidal and high plasma currents, has been studied in DIII-D tokamak discharges [2]. The MMM7.1 anomalous transport model includes a combination of contributions based on different transport theories. It includes the Weiland module for ion temperature gradient modes, trapped electron modes and collision dominated MHD modes, the Rafiq module for drift-resistive-inertial ballooning modes (DRIBM) and the Horton module with the Jenko threshold for anomalous transport driven by Electron Temperature Gradient (ETG) modes. The role of different modes described by MMM7.1 is investigated. In particular, the temperature profiles for a number of high beta poloidal DIII-D discharges are predicted using only the Weiland and ETG components of the MMM7.1 model. The magnitudes of the predicted temperature profiles are found to be in reasonable agreement with experimental profiles. However, the experimental profiles have an internal transport barrier in temperature profiles, due to strong off-axis ECR heating, which is not reproduced in the predictive MMM7.1 simulations due to significant electron thermal transport from both the Weiland component and from the ETG component of MMM7.1. The effect of electron thermal transport due to the DRIBM model is also investigated. PTRANSP analysis of the DIII-D discharge 154406 shows that there is a significant transport predicted by DRIBM in the region from 0.3 to 0.7 of normalized minor radius. The electron temperature in this PTRANSP simulation is found below the experimental values. The DRIBM model includes contributions from other MHD modes in addition to the drift resistive ballooning modes that can be unstable in this region. This validation study suggests that the DRIBM predicts a significantly larger level of electron transport than expected. Possible effects that can contribute to stabilization of these modes, for example, effects associated with the large poloidal beta such as the Shafranov shift stabilization in the MMM7.1 model, are discussed. 1. T. Rafiq et al. Physics of Plasmas 20 (2013) 032506. 2. A.M. Garofalo et al. Proc. of 25th IAEA Fusion Energy Conference (St. Petersburg, Russian Federation 13-18 October 2014) 657.
Speaker: Dr Alexei Pankin (Tech-X Corporation)
• 14:00
Global 3D Braginskii simulations of the tokamak edge region 4h 45m
A study of L and H mode-like plasma turbulence in the edge of tokamaks is presented, with an emphasis on characterization of these plasmas in numerical simulations with a new Global Drift-Ballooning (GDB) model. This work employs drift-reduced Braginskii two-fluid equations for electromagnetic low-frequency turbulence and solves them in a global large-aspect ratio annulus centered on the last closed flux-surface (LCFS) as an approximation of small to medium-size tokamaks. The simulations include plasma sources at the inner edge of the pedestal region as well as a limiter region in the Scrape-Off-Layer (SOL) and evolves self-consistently the density, temperature, and ExB shear profiles on the transport time-scale. GDB is able to generate both L and H mode-like plasmas with realistic parameters. L-mode transport appears to be largely driven by resistive-ballooning structures, in the presence of a balance between ExB and the ion-diamagnetic drifts. Pressure profiles also appear to exhibit a near-SOL breakpoint that Mirror Langmuir Probes (MLP) detect in C-Mod, postulated to separate Drift Wave (DW) like and RB-like fluctuations. Separate simulations carried out with H-mode parameters develop improved confinement, $E_r$ wells at the LCFS and spontaneous generation of temperature pedestals with density pedestals remaining absent up to times in the order of 0.2 ms. Candidate first-principles explanations to the modification of the electric field profile are discussed. This material is based upon work supported by the U.S. Department of Energy, Office of Science, Office of Fusion Energy Sciences, under Award Number DE-SC0010508.
Speaker: Mr Manaure Francisquez (Dartmouth College)
• 14:00
Gyrokinetic analysis of the effects of electron-scale turbulence on ion-scale micro-instabilities 4h 45m
Most previous studies on plasma turbulence have assumed scale separation between electron-scale (~ electron gyro-radius) and ion-scale turbulence (~ ion gyro-radius). However, multi-scale turbulence studies by using the latest supercomputers indicated existence of cross-scale interactions and its significant impact on turbulent transport, and are necessary for explaining experimental transport levels. Our recent work revealed a part of the cross-scale interactions: suppression of electron temperature gradient modes (ETG) by flow shears in ion temperature gradient mode (ITG) turbulence, and enhancement of ion-scale transport due to damping of the zonal modes by electron-scale turbulence. Since effects of electron-scale turbulence on ion-scale transport have not yet been fully revealed, it is important to explore the physical process in detail and to extend the analysis to other micro-instabilities. Here, we report two analyses: (i) detailed investigation of the damping effects of ETG turbulence on zonal flows created by the ITG turbulence, and (ii) effects of the ETG turbulence on linear growth of the micro-tearing mode (MTM). First, we have analyzed the ITG/ETG turbulence simulation by using the gyrokinetic entropy transfer analysis. It is revealed that the zonal modes are mainly driven by the ion-scale modes, where relatively higher-wave-number modes are driven by the coupling with the twisted mode caused by kinetic electrons. Electron-scale turbulence effectively damps these higher-wave-number zonal modes. Second, we have investigated the effect of ETG turbulence on linear MTM growth. It is observed that the growth of MTM can be suppressed as ETG-driven turbulent fluctuations increase, which suggests that the ETG turbulence may interrupt the linear MTM growth. Our analyses shed a light on the effects of electron-scale turbulence on ion-scale micro-instabilities: (i) ETG turbulence damps relatively higher- wave-number zonal flows created by ITG turbulence with twisted modes, and (ii) ETG turbulence can distort the resonant mode structure of MTM and interrupt its linear growth. In both cases, kinetic electrons play important roles such as creation of the twisted mode and of the current sheet. This emphasizes the significance of intermediate-scale structures for the cross-scale interactions.
Speaker: Dr Shinya Maeyama (Nagoya University)
• 14:00
Gyrokinetic simulation of tokamak edge plasmas 4h 45m
It has been recently discovered that the trapped electron mode (TEM) may play an important role in the H mode edge plasma for domestic tokamaks such as EAST and HL-2A. The stability and transport for TEM for the edge parameters are studied using large scale gyrokinetic particle simulations. The gyrokinetic simulation reveals the parametric dependences on the wavelength, collisionality and the electron temperature gradients. The un-conventional ballooning mode structure is found the H mode edge parameters, which directly leads to a change in the transport characteristics in the edge. The zonal flow is found by the gyrokinetic simulation to be less important in the edge than in the core. In order to interpret the simulation results, a simplified analytic theory is developed to include both collisional and strong gradient edge characteristics.
Speaker: Prof. Yong Xiao (Institute for Fusion Theory and Simulation)
• 14:00
Gyrokinetic simulations of an electron temperature gradient turbulence-driven current in tokamak plasmas 4h 45m
The so-called “spontaneous” or “intrinsic” rotation driven by ion-scale turbulence has been widely observed in tokamaks. If we turn our attention to the electron parallel momentum balance, it is likely that electron-scale turbulence, e.g. electron temperature gradient (ETG) turbulence, can modify the Ohm’s law, hence providing a current source. However, there has been no serious study of an ETG-driven current in self-consistent simulations using realistic tokamak geometry. In this work, we report results of a gyrokinetic simulation study elucidating the characteristics of an intrinsic current driven by ETG turbulence in toroidal geometry. We focus on effects of the normalized electron gyroradius rho_e* on the ETG-driven current. Our simulations demonstrate that the amount of the ETG-driven current increases with rho_e*, as expected from the gyro-Bohm scaling. In particular, a perturbation of a q-profile by the ETG-driven current becomes visible when a<4000 rho_e. This finding suggests that a significant intrinsic current can be driven inside an H-mode pedestal where the steep gradient of an electron temperature pedestal can excite ETG turbulence in a narrow region.
Speaker: Dr Sumin Yi (National Fusion Research Institute, Daejeon, Republic of Korea)
• 14:00
Gyrokinetic Simulations of Microturbulence in DIII-D pedestal 4h 45m
Present understanding of ELM triggering mechanism is mostly based on the peeling-ballooning theory (PBT) often providing sufficiently good pedestal prediction. While PBT is rather empirical, the more comprehensive kinetic description is still required. In this work we present recent gyrokinetic simulations aimed to identify electromagnetic microinstabilities in the H-mode pedestal region of DIII-D tokamak (discharge #131997 at 3011 ms) using global gyrokinetic code GTC. It was found that dominant instability at the top of the pedestal is the ion temperature gradient mode (ITG). In the middle of a pedestal the kinetic ballooning mode (KBM) becomes the most unstable for the intermediate range of toroidal mode number n~20. For shorter wavelengths the dominant instability is TEM. We have demonstrated the ITG-KBM transition at the pedestal top and TEM-KBM transition in the steep pressure gradient region as plasma pressure increases. One possibility to control drastic ELM activity during H-mode operation is applying resonant magnetic perturbations (RMP) however the detailed mechanism of RMP effect is not completely clear. In our studies we address the direct effect of modified magnetic equilibrium geometry on microturbulence in DIII-D pedestal. By fixing the profiles, and excluding magnetic stochasticity effects, we examine the effect of various strength RMP on KBM stability, and turbulent transport. We have observed the increase of KBM growth rate when RMP is applied; however this change is only detectable for artificially amplified RMP strength. The direct effect of RMP geometry perturbation on zonal flow generation and turbulent transport is found to be insignificant. Work is supported by U.S. DOE theory grant DE-SC0010416, DE-SC0013804, and DOE SciDAC GSEP Center.
Speaker: Dr Ihor Holod (University of California Irvine)
• 14:00
GYROKINETIC SIMULATIONS OF TOKAMAK PEDESTALS- PRESENT EXPERIMENTS AND EXTRAPOLATION TO BURNING PLASMAS 4h 45m
For the first time, electromagnetic gyrokinetic simulations of pedestal transport are reported (inter-ELM). For the JET-ILW (ITER Like Wall) pedestal, nonlinear simulations show that Micro-Tearing Mode (MTM) turbulence produces the bulk of the transport in the steep gradient region, and the combination of MTM, electron temperature gradient (ETG), ion-scale electrostatic turbulence and neoclassical transport reproduces experimental power balance across most of the pedestal. Pedestals with nustar* < 1 are often well into the second stability region, so Kinetic Ballooning Modes do not strongly affect pedestal transport- as indicated by previous linear analysis of JET-Carbon cases [1]. A rho* scan of ITER-like pedestals is performed, keeping other dimensionless parameters constant. Simulations find gyroBohm scaling of transport in the range of rho* of ASDEX/DIIID through low field JET. However, for high field JET and beyond, an insufficiency of velocity shear leads to strong ion scale electrostatic turbulence, and a strong departure from gyroBohm at lower rho* such as ITER. Inclusion of Carbon or Nitrogen greatly reduces this turbulence, so that gyroBohm scaling is reestablished through JET, and the departure at ITER is substantially reduced. Pedestal transport is also strongly affected by the separatrix density, which can be affected by gas puffing. These trends may account for observed differences in pedestal behavior in JET-ILW and JET-Carbon. Unstable electrostatic eigenmodes have an unusual structure in the pedestal, and localize where the velocity shear is low – near the top and bottom. In addition to including low Z impurities, operation with a lower separatrix density can greatly reduce the problem, which may be possible with advanced divertor geometries of Lithium. Finally, initial results indicate that low aspect ratio may have advantages for avoiding shear insufficiency. 1 S. Saarelma et. al., Nucl. Fusion 53 123012 (2013).
Speaker: Dr Mike Kotschenreuther (Institute for Fusion Studies)
• 14:00
Integrated Simulation of Deuterium Experiment Plasma in LHD 4h 45m
The deuterium experiment project from 2017 is planned in LHD, where the deuterium NBI heating beams with the power more than 30MW are injected into the deuterium plasma. Principal objects of this project are to clarify the isotope effect on the heat and particle transport in the helical plasma and to study energetic particle confinement in a helical magnetic configuration measuring triton burn-up neutrons. In this paper, the deuterium experiment plasma of LHD is investigated by applying the integrated simulation code TASK3D and the 5-D drift kinetic equation solver GNET. First, we perform the integrated transport simulation of deuterium plasma, n_D/(n_H+n_D)=0.8, by TASK3D code assuming a typical flat density profiles. We evaluate the heat deposition profiles for the multi-ion species plasma (e, H, He, C) by using the multi-ion version of GNET, which can treat the D and H ion heatings precisely. One-dimensional (1-D) diffusive heat transport equation with multi-ion species (H, He, C) is solved using the heat deposition profiles by GNET. It is found that the deuterium ion temperature reaches more than 6 keV with the isotope effect in the deuterium experiment plasma. On the other hand, the ion temperature reaches about 5 keV if we assume a pure hydrogen plasma. This result indicates that we will obtain about 20% higher ion temperature than that of the hydrogen plasma in the deuterium experiment of LHD if we assume an isotope effect on the turbulent transport based on the He/H experiment results. Next, we perform the triton burn-up simulation of the deuterium experiment of LHD and evaluate the D-T fusion reaction rates to compare with the experimental results of the 14 MeV neutron diagnostic system. It is found that more than 7.0x10^11 m^-3/s of 14MeV neutrons are generated by the D-T fusion reaction at the plasma center. We also find that the confinement of the 1MeV tritons is improved by the strongly inward shifted configuration of LHD (R_ax=3.5m). and that the triton burn-up ratio, which is the ratio of 14 MeV to 2.5 MeV neutron production, is increased to about 0.1%, which is still smaller than that of the large tokamak experiment results.
Speaker: Dr Sadayoshi Murakami (Departement Nuclear Engineering, Kyoto University)
• 14:00
Investigation of Sustainable Reduced-Power non-inductive Scenarios on JT-60SA 4h 45m
Along with the construction and operation of ITER, the design of a demonstration thermonuclear fusion reactor (DEMO) is the main goal of current international fusion research. New generation of tokamaks as JT-60SA are meant to provide important information to allow discriminating between different DEMO designs. In particular JT-60SA will explore the possibility of running steady state plasma scenarios characterised by high fraction of bootstrap current, low flux consumption and sustainable divertor heat-loads. The feasibility of the above scenarios will depend on the simultaneous control of core/divertor/SOL conditions to maintain a peaked pressure profile, clean plasma while ensuring an acceptable heat load on the divertor targets. Preliminary investigations of the SOL/divertor conditions show that sustainment of the steady state scenario without impurity seeding will be challenging due to the large heat loads which are likely to appear when 30 MW of NBI power are employed. Before developing seeding schemes for the full power scenario it will be therefore necessary to prepare a reduced-power optimised scenario where both the fraction of non-inductive current and beta are maximised while the heat flux to the divertor is kept at a sustainable level. The above reduced-power scenario has been investigated with the integrated suite of core/SOL/divertor codes JINTRAC. A scan in NBI power and fuelling rate/location has been performed and found that acceptable levels of power-load on the outer divertor plate can be achieved in the absence of impurity seeding when the NBI power is lowered to 17 MW. The 0-D plasma parameters of this lower power / high fraction of bootstrap current scenario are discussed in this paper, along with the role on performance of the internal transport barrier, and the comparison against the reference values of the JT-60SA research plan. Acknowledgement: This work has been carried out within the framework of the EUROfusion Consortium and has received funding from the Euratom research and training programme 2014-2018 under grant agreement No 633053. The views and opinions expressed herein do not necessarily reflect those of the European Commission.
Speaker: Dr Michele Romanelli (UK Atomic Energy Authority)
• 14:00
ITER Fuelling Requirements and Scenario Development for H, He and DT through JINTRAC Integrated Modelling 4h 45m
The evolution from X-point formation of ITER H, He and DT plasmas with gas and/or pellet fuelling has been studied for the first time self-consistently with the integrated core and edge suite of codes JINTRAC developed at JET. Our results show that understanding how to optimise fuelling performance is vital to operate ITER and to achieve high fusion yield without exceeding operational limits for neutral beam (NB) shine-through and divertor power fluxes (10MW/m^2).In present devices gas can fuel the core as the edge plasma is fairly transparent to neutrals. In contrast the ITER edge plasma will be hotter and denser so more gas will be ionised in the far scrape-off-layer (SOL) and not penetrate to the separatrix.We show that routine use of pellets in ITER is likely to reach the minimum density for safe NB operation in L-mode or for H-modes with Q equals 10. In L-mode with only gas fuelling, we reach a Greenwald density fraction less than 30% before the density build-up in the SOL leads to a MARFE. These can also be triggered by pellets and to avoid them in our simulations we had to fine tune the discrete pellet mass to the core density.To access ELMy H-mode in 15MA/5.3T ITER DT plasmas, we show that alpha heating is crucial. Thus, during the L-H transition fuelling will have to be kept low to allow the ion temperature to rise on-axis to boost the build-up of fusion power. Here, we prove the viability of heating (33MW NB, 20MW ICRH) and fuelling schemes to reach Q of 10, with no need for Ne seeding to ease the divertor power loads while Q is less than 5.For a 15MA/5.3T ITER DT Q equals10 baseline scenario we show that if particle fuelling is too high during the H-L transition, the target power loads in our simulations may stay below the design limit, but a MARFE may occur. In summary our results show that pellets may be crucial to obtain an L-mode density above the NB shine-through limit. We also prove that density control during the L-H and H-L transition is critical. Pellet fuelling should rather be turned off during the L-H transition to aid the access to ELMy H-mode by minimising the density rise to boost the fusion power. Gas fuelling and Ne seeding can be used during the H-L transition to keep the power loads to the divertor tolerable, but precise feedback control over radiation is needed to keep the plasma within the permitted operational range.
Speaker: Dr Elina Militello Asp (CCFE)
• 14:00
Multi-species ITG-TEM driven turbulent transport of D-T ions and He-ash in ITER burning plasmas 4h 45m
Burning plasmas are composed of multiple ion species such as fuel isotopes(D and T) and He-ash produced by the fusion reaction, and more complex turbulent transport processes are expected in comparison to the single-ion plasmas. Since simultaneous measurements of the kinetic profiles for all species are limited even in experiments, systematic studies on the particle and heat transport by the first-principle-based gyrokinetic simulations are indispensable for the prediction of the confinement performance and the optimization of the impurity exhausts and D/T fueling. In this study, the ion-temperature-gradient and trapped-electron-mode (ITG-TEM) driven turbulent transport in realistic ITER plasmas is investigated by means of the multi-species electromagnetic gyrokinetic Vlasov simulation GKV[Watanabe et al., NF2016] with D, T, He, and real-mass kinetic electrons including their inter-species collisions[Nakata et al., CPC2015, Nanami et al., PFR2015], where a good prediction capability has been confirmed against the actual JT-60U tokamak experiment[Nakata et al., IAEA-FEC2014]. The GKV simulations reveal different saturation levels and spatial structures of the turbulent fluctuations in D-T ions and He-ash. For the first time, gyrokinetic-simulation-based quantitative evaluation of a steady burning condition[Reiter et al., NF1990] with He-ash exhaust and fuel inward-pinch is realized by extensive nonlinear scans. Furthermore, the strong impacts of D-T fuel ratio and He-ash accumulations on turbulent energy and particle fluxes are clarified. New findings in this study, which are crucial for the burning plasma performance, are (i)imbalanced D-T turbulent particle transport strongly influenced by He-ash accumulations, and (ii)identification of the steady burning profile regimes with He-ash exhaust and D-T fuel inward-pinch associated with the off-diagonal transport.
Speaker: Dr Motoki Nakata (National Institute for Fusion Science)
• 14:00
New Nonlinear Microtearing Mode Transport Model for Tokamak Plasmas* 4h 45m
Microtearing modes (MTMs) have been identified as a source of significant electron thermal transport in high β tokamak discharges. A model for MTMs that can be installed in integrated whole device predictive modeling codes is needed in order to improve the prediction of electron thermal transport and, consequently, the evolution of the plasma in devices in which MTMs have a significant role. A unified fluid/kinetic approach is used in the development of a nonlinear model for the transport driven by MTMs. The derivation of the model includes the effects of electrostatic and magnetic fluctuations (δB), collisionality, electron temperature and density gradients, magnetic curvature and the effects associated with the parallel propagation vector (k_∥). The electron momentum, electron density, Maxwell equations, Ampere’s law and quasi-neutrality condition are used in the derivation of a nonlinear fluid microtearing dispersion relation. An iterative nonlinear approach is used to calculate distribution function employed in obtaining the nonlinear parallel current and the nonlinear dispersion relation. The third order nonlinear effects in δB are included in the development of microtearing mode model, and the influence of third order effects on a three wave system is considered. For the evolution of the nonlinear microtearing instability in time, the third order effects provide the possibility of saturation of the microtearing instability. In the limit of slab geometry, k_∥ = 0 and strong collisionality, the fluid dispersion relation for nonlinear microtearing modes is found to agree with the kinetic dispersion relation in Ref [1]. An envelope equation for the nonlinear microtearing modes in the collision dominant limit is introduced. The role of MTMs in driving electron transport in the mid-radius region of NSTX-U plasmas is examined, and the dependece of MTMs on plasma parameters including the magnetic shear length, safety factor, electron temperature gradient, electron density gradient, plasma β and collisionality is studied. *This work is supported by the U.S. Department of Energy, Office of Science, under Award Number DE-SC0013977 and DEFG02-92-ER54141 and DE-SC0012174. [1] J.F. Drake et al. Phys. Rev. Letters 44, 994 (1980)
Speaker: Dr Tariq Rafiq (Lehigh University)
• 14:00
On Benchmarking of Simulations of Particle Transport in ITER 4h 45m
We report present status and main results of the ITPA IOS Topical Group activity on the benchmarking of simulations of the core particle transport in ITER baseline ELMy H-mode scenario with the integrated codes which are presently used for the ITER scenario simulations. The ITPA IOS group is pursuing particle transport as an important component of integrated modelling, because the simulations have shown that dynamics of the particle transport plays a key role in the possibility to access and sustain the H-mode and stable burn conditions and to provide controllable shut-down of DT discharge in ITER. Optimisation of the fuelling scenario for ITER requires sufficiently accurate numerical solvers with appropriate description of particle sources, sinks, boundary conditions and integration in the codes for simulations of self-consistent plasma evolutions. Core particle transport is being studied in the frame of code benchmarking within the ITPA IOS group with various integrated modelling codes used for the ITER scenario simulations. The purpose of the benchmark is to verify agreement among various integrated modelling codes by approximating closely the expected scenario on ITER and to predict ITER plasmas more accurately based on knowledge accumulated from the benchmark so to address the critical issues of ITER. It includes comparison of the particle transport solvers, description of the sources and sinks, as well as its implementation in the integrated codes. As a first step, the benchmark is carried out with identical prescribed particle sources, sinks, transport coefficients and boundary conditions for one time slice in the flattop H-mode phase to compare and understand differences among the codes. As a next step, we pursue a series of sensitivity studies and model expansions and improvements. Finally, the impact of particle transport on ITER fusion performance is discussed in time evolution simulations. The results of our benchmarking can be used for the choice of the level of approximation of the particle transport description necessary and sufficient for simulations of the ITER and DEMO scenarios.
Speaker: Prof. Yong-Su Na (Seoul National University)
• 14:00
Physics-based integrated modeling of the energy confinement time scaling laws in tokamaks 4h 45m
As an effort to clarify the physics origin of the global scaling laws of energy confinement time, a new analysis scheme is first proposed in which the total stored energy is divided into the two parts, one being almost directly decided by the marginal stability property and edge boundary condition through profile stiffness and the other by the profile deviation from marginal one through turbulent dynamics under external heating. Initial application to the two parameter cases of plasma current and input power show this scheme is quite effective for identifying the relative role of various physics elements, such as the linear stability, nonlinear turbulent dynamics, pedestal boundary and core-edge coupling, in determining the global scaling law. Particularly, in the plasma current case it is found most of its scaling is originated from the marginal part with the significant role of the pedestal boundary. More detailed analysis results, including the other parameter cases, will be reported in the conference paper.
Speaker: Dr Jin Yong Kim (National Fusion Research Institute, Korea (south))
• 14:00
Predicted fusion performance for ITER and DEMO plasmas using a BALDUR code with predictive tritium influx model 4h 45m
The deuterium and tritium are considered as a fuel for nuclear fusion reactors in the future fusion machine, like ITER and DEMOs. Generally, deuterium is applied by gas puffing or pellet injection; whereas tritium can be internally produced from a blanket of reactors, which relies on reactions between 14.1 MeV neutrons from nuclear fusion reactions and lithium as one composite of the blankets. In this work, a model for predicting tritium flux generated from lithium blanket is developed based on the Monte Carlo code MCNP5, and implemented in the BALDUR integrated predictive modeling code to provide the information of tritium flux coming to main plasma. This suite of code is then used to carry out an evolution of plasma current, densities and temperature in ITER and DEMOs under L-mode and type I ELMy H-mode scenarios. Two designs of DEMOs considered are Chinese design and European design. In these simulations, a combination of NCLASS neoclassical transport and Multi-mode anomalous transport models (either MMM95 or MMM8_1 version) is used to compute a core transport. It is found that the wide range of fusion performance can be achieved, depending on designs and operation modes. The sensitivity of fusion performance due to the variation of plasma parameters, i.e. plasma current, toroidal magnetic field, plasma density, and auxiliary heating power, is also carried out. The ignition test for each design is also conducted. It is found that only plasmas in some of these designs can sustain the plasma and fusion reactions with slightly lower fusion performance after external heating is removed.
Speaker: Dr thawatchai Onjun (Sirindhorn International Institute of Technology)
• 14:00
Progress in the ITER Integrated Modelling Programme and the use and validation of IMAS within the ITER Members 4h 45m
The ITER Integrated Modelling (IM) Programme will not only support the ITER Project in the development and execution of the ITER Research Plan (IRP) but also provide support for the design basis of the ITER facility during construction, in particular for diagnostics. Strategically, the ITER IM Programme is implemented using expertise and technologies developed within the ITER Members’ research programmes with annual reviews by an Integrated Modelling Expert Group (IMEG) comprised of experts from all the ITER Parties. The Integrated Modelling & Analysis Suite (IMAS) is the software infrastructure that has been developed in response to the needs of the IM Programme and which will support the requirements of both plasma operations and research activities. An agile approach is taken to the development of IMAS and a software management framework consisting of linked issue tracking, source code repositories and a continuous integration server to automatically build and regression test revisions has been established. It is essential that results generated for ITER are reproducible and so software hosting and rigorous version control are prerequisites and already ensured, whilst provenance tracking for handling inputs is still in development. The unifying element of IMAS is its use of a standardized data model capable of describing both experimental and simulation data. This enables the development of workflows that can flexibly use different software components as well as being independent of the device being modelled. This makes IMAS an ideal framework for conducting code benchmarking exercises, such as that within the ITPA Energetic Particle Physics Topical Group on the calculation of fast ion distributions. In this paper, some of the initial software adaptations are presented to indicate the use, and consequent validation, of IMAS within the ITER Members. This has been facilitated by the release this year of a local installer for IMAS which has already allowed installation within the research facilities of the majority of the ITER Members including the EU, India, Japan, Korea and the US. For the most part, these workflows are predictive in nature with interpretive workflows expected to follow from the development of plugins to the IMAS data access tools to securely read and map remote experimental data from existing devices into the standardised data model.
Speaker: Dr Simon Pinches (ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St. Paul-lez-Durance Cedex, France)
• 14:00
Progress in the theoretical description and the experimental characterization of tungsten transport in tokamaks 4h 45m
The validation of current models to predict the transport of a heavy impurity like tungsten in tokamaks is confronted with challenges from both the theoretical and experimental standpoints. Both neoclassical and turbulent transport mechanisms are involved, and have to take into account the impact of poloidal asymmetries, produced by both centrifugal effects and temperature anisotropies caused by auxiliary heating. These can significantly modify the neoclassical transport, affecting both its amplitude and the strength of the temperature screening. The size of turbulent impurity transport strongly depends on the ratio of the electron to the ion heat flux and is maximized when this ratio slightly exceeds unity. Moreover, in these conditions, subdominant modes can non-negligibly impact the transport, and lead to turbulent convections which are in opposite direction with respect to the predictions based on the most unstable linear mode only. Theory validation benefits of experiments which are dedicated to testing detailed predictions. However, experiments on heavy impurity transport face limitations on the possibility of diagnosing the impurity density, the accessible domains of plasma parameters and the available heating systems in each device. Experiments have been performed in ASDEX Upgrade to investigate the impact of central wave heating in the avoidance of W accumulation in H-mode plasmas with dominant neutral beam injection heating. Experiments show that ion cyclotron and electron cyclotron heating have similar effects on the W behavior when similar power density profiles are produced with the two wave heating systems, consistent with dominant electron heating produced by ICRH in the H minority scheme in ASDEX Upgrade. Theory-based modelling is performed by combining the GKW and NEO codes, and conditions under which the role of H-minority becomes significant are highlighted. These results are compared with companion experimental and modelling research performed at JET. Finally the implications of the experimental and theoretical results on the prediction of the tungsten behavior in ITER and a future reactor are presented. General parametric dependencies of confinement with increasing size of the device support the favorable expectation for impurities that in a reactor the impact of neoclassical transport is reduced with respect to present tokamaks.
Speaker: Dr Clemente Angioni (Max-Planck-Institut fuer Plasmaphysik)
• 14:00
Recent EUROfusion Achievements in Support to Computationally Demanding Multi-scale Fusion Physics Simulations and Integrated Modelling 4h 45m
Integrated Modelling (IM) of present experiments and future tokamak-reactor requires numerical tools which can describe spatially small-scale and large-scale phenomena as well as dynamically fast transient events and relatively slow plasma evolution within a reasonably fast computational time. The progress in the optimisation and speed-up of the EU first-principle codes and in the development of a basis for their integration into a centrally maintained suite of IM tools achieved by the EUROfusion High Level Support Team (HLST) and Core Programming Team (CPT) is presented here. An overview of the physics phenomena which can be addressed in various areas (core turbulence and magnetic reconnection, collisional transport in non-axisymmetric devices, edge and SOL physics, heating and current drive, pedestal physics, MHD and disruptions, reflectometry simulations) using the improved numerical tools is given. The optimisation of physics codes performed by HLST allowed one to achieve six-fold speed-up of SOLPS-ITER simulations due to OpenMP parallelisation of the B2 part of SOLPS; to investigate kinetic effects in SOL region using the realistic 3D geometry implemented in BIT2/BIT3; to perform the reflectometry simulations (REFMULX/REFMULF) for ASDEX Upgrade or JET much more accurately and to preview with more reality the behaviour of reflectometry in ITER or DEMO; to resolve realistic wall structures enabling the simulation of the precise current patterns required for the prediction of asymmetric forces during disruption events (JOREK-STARWALL). The CPT development activities in support to integrated modelling including a support to local deployment of the IM infrastructure and experimental data access, to the management of releases for sophisticated IM workflows involving a large number of components and to the performance optimization of complex IM workflows are summarised.
Speaker: Dr Irina Voitsekhovitch (CCFE)
• 14:00
Simulation of Neoclassical Tearing Modes in JET 4h 45m
In this work, a simulation of neoclassical tearing mode (NTM) in JET experiments is considered using a 1.5D BALDUR integrated predictive modeling code with an improved ISLAND. An original ISLAND module [1] for calculating the saturated width of magnetic island caused by a magnetic reconnection is obtained from the NTCC Library [2]. This ISLAND module is then modified to improve a consistency and reliability of island width prediction. The modified ISLAND module is still based on Rutherford equation [3] and quasi-linear theory approach. With the improved ISLAND module integrated in 1.5D BALDUR, the effects of neoclassical tearing modes (NTM) can be considered. The effect of NTM is described using the model that both thermal and particle transport within the magnetic island is enhanced, resulting on the flattening of profiles within that region. The BALDUR code with a modified ISLAND module is then used to carry out the time evolution of plasma current, temperature, and density profiles, where the effects of NTM can be real time considered. For example, in JET discharge No. 33131, the simulations with magnetic islands mode (2,1), or with magnetic island mode (3,2), or with both magnetic island mode (2,1) together with mode (3,2), are carried out. It is found that when the magnetic island mode (2,1) is considered, the ion and electron temperature profile, and also the total stored energy profile are decreased the most comparing to the other two scenarios.
Speaker: Dr Nopporn Poolyarat (Department of Physics, Faculty of Science and Technology, Thammasat University, THAILAND)
• 14:00
Statistical validation of transport models on baseline discharges in preparation for the extrapolation to JET D-T 4h 45m
The EUROfusion Consortium is planning deuterium-tritium (D-T) experimental campaigns in 2019 on JET with the ITER-Like Wall (ILW) to address physics issues which are important for ITER-D-T experiments. To achieve the scientific objectives, JET operation should demonstrate 10-15MW of fusion power for at least 5 seconds, a performance never attempted before in fusion-research history. The preparation of the D-T campaign requires, therefore, reliable predictive simulations of this unprecedented JET operational scenarios, providing assessment of the impact of uncertainties resulting from operating with an ILW such as degraded edge confinement and core tungsten accumulation, and from operating with a D-T mixture such as isotopic effects on stability and confinement and alpha heating. Despite the remarkable improvements in present core transport models such as GLF23 and TGLF, the current ability to predict plasma temperature evolution and the resultant fusion power is still limited due to the incompleteness of first principles theories of energy and particle transport in turbulent thermonuclear plasmas and the uncertainty of input data required for predictive simulations such as pedestal temperature, Zeff, and rotation profiles. Thus, for a quantitative assessment of the uncertainties in the D-T performance with ILW, statistical validation of predictive simulations with a large database of D-D discharges is of crucial importance. Predictive TRANSP simulations with advanced transport models such as GLF23 and TGLF for JET experiments is now available using automated input data preparation routines, the JET-TRANSP scripts, which enables one to carry out a large number of predictive TRANSP simulations. In this paper, statistical assessment of the level of agreement of predictive TRANSP simulations with the GLF23 transport model carried out on a large set of well diagnosed JET baseline discharges will be presented, and sensitivity studies on uncertain parameters such as pedestal Ti, Zeff, and rotation profiles will be discussed. The assumption used in the simulation is further investigated for a few representative discharges with the TGLF transport model, which is computationally expensive but more accurate than GLF23. This statistical validation with the assessment of uncertainty level will constitute the basis for TRANSP predictions of JET-ILW-DT experiments.
Speaker: Dr Hyun-Tae Kim (EUROfusion Consortium JET)
• 14:00
Steep gradients in plasma confined at convex-concave magnetic field lines 4h 45m
The formation of large stable plasma gradients, e.g. in form of internal transport barriers, being of a strong both practical and fundamental interest. Normally the larger the gradient the larger the transport, and any deviation due to collective plasma behavior is of great interest. We have predicted theoretically that there is a strong stabilizing action against convective (flute-interchange) perturbations when plasma is confined by magnetic field of alternating-sign curvature – i.e. with convex–concave field lines [Tsventoukh 2014 Nucl. Fusion 54 022004]. The calculations that have been done for simple combinations of axisymmetric mirrors and cusps according to the kinetic stability criterion, give strongly centrally peaked stable plasma pressure profiles instead of shallow ones. We have performed an experimental investigation of the plasma confinement at magnetic confinement device of the alternating-sign curvature [Tsventoukh et al 2015 Nucl. Fusion 55 062001]. For the experimental research of this effect, a compact magnetic confinement device has been modified by adding of the external current coil to fulfil the field-line curvature requirements. The critical convectively-stable plasma pressure profiles calculation in this experimental geometry and the probe measurements of the spatial plasma distribution in the new magnetic configuration of alternating-sign curvature have been performed. The experimental results give some support for a conclusion that there is an increase in the ion saturation current at the region near the minimum of the specific volume min ∫dl/B. This region corresponds to the average minimum in the second adiabatic invariant, and the kinetic description predicts the stable pressure profile peaking here due to reduction of charge separation by particle drift in alternating-sign curvature. For further experimental investigations, a stationary microwave device has been used. A mirror geometry has been created by axisymmetric coils, Langmuir and magnetic probes have been used for the measurements. For the theory developing the effects of a finite plasma beta has been analyzed in axisymmetric equilibrium, and plasma particle kinetics effect on the plasma transport. Work was supported by RFBR grant # 15-38-20617.
Speaker: Dr Mikhail Tsventoukh (Lebedev Physical Institute RAS)
• 14:00
The Development of SOL Transport Model for Integrated Core-SOL Simulation of L-Mode Plasma 4h 45m
Simulations of the plasma in the core and the scrape-off layer (SOL) region are carried out using 1.5D BALDUR integrated predictive modeling code to investigate tokamak plasmas in TFTR reactor operating in low confinement mode (L-mode). In each simulation, the plasma current, temperatures, and density profiles in both core and SOL regions are evolved self-consistency. The plasma profiles in the SOL region is simulated by integrating the fluid equations, including sources, along the field lines. The solutions in the SOL subsequently provide as the boundary conditions of the core plasma region. The core plasma transport model is described using a combination of anomalous transport by Multi-Mode-Model version 1995 (MMM95) and neoclassical transport provided by NCLASS module. Furthermore the calculation of the toroidal velocity used in this work is based on the torque due to intrinsic neoclassical toroidal viscosity (NTV). While the transport coefficients in the SOL region are either determined by fixed constants or neoclassical transport based on NCLASS calculation. By comparing with eight L-mode discharges from TFTR, it was found that the simulations using the transport based on neoclassical theory for SOL transport yields better agreement to experimental results for both density and temperature profiles.
Speaker: Dr Apiwat Wisitsorasak (King Mongkut's University of Technology Thonburi)
• 14:00 18:45
Poster EX/1, TH/1
• 14:00
Enhanced understanding of non‐axisymmetric intrinsic and controlled field impacts in tokamaks 4h 45m
An extensive study of intrinsic and controlled non-axisymmetric field impacts in KSTAR has enhanced the understanding about non-axisymmetric field physics and its implications, as well as demonstrating the importance of optimal 3-D configurations in resonant magnetic perturbation (RMP)-driven control on edge localized modes (ELMs) in tokamaks. The $n=1$ intrinsic non-axisymmetric field was measured to remain as low as ${\langle\delta B/B_0\rangle}_{m/n=2/1} \sim 4\times 10^{-5}$ at high-beta plasmas ($\beta_N\sim2$), which corresponds to approximately 20% below the targeted ITER tolerance level. A systematic survey of $n=1$ controlled resonant field has revealed that KSTAR has a lower power threshold for L-H transition (at least 10 %) than DIII-D (configured with $n=3$ RMP) with similar plasma densities of $n_e=(2 -2.6)\times 10^{19} m^{-3}$, possibly benefiting from a low level of intrinsic error field and toroidal field ripple. As for the RMP ELM control, a high-quality $n=1$ RMP ELM suppression (duration of $\sim40 \tau_E$) was achieved using an operationally ‘reproducible’ approach. Throughout this investigation, we diagnosed edge activities using 3‐D ECE imaging diagnostics (ECEI) on both high-field-side (HFS) and low-field-side (LFS) simultaneously for the first time. According to ECEIs, the RMP ELM suppression was full of lively edge activities, which appears quite challenging to a prevailing theory that ‘peeling‐ballooning’ stability boundary is crossed from unstable to stable regimes due to RMP. While exploring the most favorable 3-D configuration ($n=1$, +90 deg. phasing), we discovered that midplane IVCC coils played a major role in mitigating the ELMs, while two off-midplane IVCCs ($n=1$ odd-parity) appeared insignificant on ELMy behavior change. In contrast, when the off-midplane IVCCs are configured with n=1 even-parity, strong plasma response was observed, even triggering mode-locking at high RMP currents. Considering that the ITER RMP coils are composed of 3-rows, just like in KSTAR, further 3-D physics study in KSTAR is expected to help us minimize the uncertainties of the ITER RMP coils, as well as establish an optimal 3-D configuration for ITER and beyond.
Speaker: Dr Yongkyoon In (National Fusion Research Institute)
• 14:00
Enhancement of helium exhaust by resonant magnetic perturbation fields 4h 45m
Exhaust of helium as a fusion born plasma impurity is a critical requirement for future burning plasmas. We demonstrate in this paper that resonant magnetic perturbation (RMP) fields can be used to actively improve helium exhaust features. We present results from the TEXTOR tokamak with a pumped limiter and from the LHD heliotron with the closed helical divertor. The results show an important additional functionality of the ITER RMP ELM control coils and dedicated experiments on present day devices like DIII-D, EAST or KSTAR which obtained full ELM suppression by RMP field application are motivated. In both devices RMP fields are applied to generate a magnetic island located in the very plasma edge and this magnetic island has a noticeable impact on the helium exhaust. At the TEXTOR tokamak, the effective helium confinement time τp,He is reduced by up to 43% and the actual reduction depends on the coupling of the magnetic island to the pump device. The LHD heliotron device, in contrast, features intrinsically a 3-D boundary and the closed helical divertor was designed for optimal pumping in this geometry. Without RMP field applied, τp,He is a factor of ~4 higher for LHD compared to TEXTOR discharges in a comparable plasma density range. Ion root transport - one out of several different impurity transport regimes at LHD - is the most likely inward transport driver causing the high τp,He. When a magnetic island is seeded into the intrinsic edge stochastic layer, a decrease of τp,He by up to 30% and hence values closer to the tokamak situation are established. This shows that RMP fields are a fine-tuning actuator for the exhaust of helium, which is an attractive additional functionality for the ITER ELM control coils. 3-D fluid plasma edge transport and kinetic neutral gas modeling with the EMC3-EIRENE code shows for LHD that the actual helium concentration in the plasma core is dominated by wall recycling of helium. This points out that the back-fueling of the plasma by helium emitted from the plasma and recycled at the wall elements needs to be controlled. The edge magnetic island induced is shown to be an effective actuator to retain the recycled helium in the plasma periphery where it can be pumped away.
Speaker: Dr Oliver Schmitz (University of Wisconsin - Madison, Department of Engineering Physics)
• 14:00
Optimization of the Plasma Response for the Control of Edge-Localized Modes with 3D Fields 4h 45m
Measurements and modeling of the plasma response to applied 3D magnetic perturbations – specifically its dependence on collisionality, beta, and rotation – yield new insight into the physics of edge-localized mode (ELM) control and better define the criteria needed to achieve ELM suppression in ITER. ELM control depends on the coupling of the applied field to a stable edge mode that drives resonant fields on edge rational surfaces and is directly observed on high-field side (HFS) magnetic sensors. The edge mode amplitude is inversely proportional to pedestal collisionality yet is insensitive to global beta, consistent with a current-driven mode as opposed to a pressure-driven kink, and reinforcing the importance of ITER-like collisionality to resonant field drive [1]. Advances in ideal MHD modeling have identified highly stable, beta-independent plasma response modes that nonetheless drive strong resonant fields – showing a path to ELM control with minimal impact on global stability [2,3]. Onset of ELM-suppression is consistent with a transport bifurcation driven by the penetration of resonant fields, evidenced by sudden changes in: boundary heat flux, visible helical striations, pedestal-top rotation and fluctuations, and HFS magnetic effects [4]. Systematic torque and beta scans reveal a loss of ELM suppression consistent with two-fluid modeling predictions of a reduction in the penetrated field [5]. These results further the development of quantitative models for the conditions necessary to achieve ELM suppression, emphasizing both edge mode coupling for resonant field drive, and low pedestal-top electron rotation for resonant field penetration. Optimization of the equilibrium conditions and discharge evolution, in addition to the applied field structure, will be required to successfully achieve ELM suppression in ITER. [1] C. Paz-Soldan et al, Nucl. Fusion 2016 (in press) [2] C. Paz-Soldan et al, Phys. Rev. Lett. 114, 105001 (2015) [3] N. Logan et al, Phys. Plasmas 2016 (in review) [4] R. Nazikian et al, Phys. Rev. Lett. 114, 105002 (2015) [5] R. Moyer et al, Nucl. Fusion 2016 (in preparation)
Speaker: Dr Carlos Paz-Soldan (Oak Ridge Institute for Science Education)
• 14:00
Penetration and amplification of resonant perturbations in 3D ideal-MHD equilibria 4h 45m
The nature of ideal-MHD equilibria in three-dimensional geometry is profoundly affected by resonant surfaces, which beget a non-analytic dependence of the equilibrium on the boundary. Furthermore, non-physical currents arise in equilibria with continuously-nested magnetic surfaces and smooth pressure and rotational-transform profiles. We demonstrate that three-dimensional, ideal-MHD equilibria with nested surfaces and delta-function current-densities that produce a discontinuous rotational-transform are well defined and can be computed both perturbatively and using fully-nonlinear equilibrium calculations. The results are of direct practical importance: we predict that resonant magnetic perturbations penetrate past the rational surface (i.e. shielding'' is incomplete, even in purely ideal-MHD) and that the perturbation is amplified by plasma pressure, increasingly so as stability limits are approached.
Speaker: Dr Stuart Hudson (Princeton Plasma Physics Laboratory)
• 14:00
Role of MHD dynamo in the formation of 3D equilibria in fusion plasmas 4h 45m
This work investigates the formation of helical core equilibria in toroidal fusion plasmas, focusing on the role of dynamo, or magnetic flux pumping mechanisms in determining the equilibrium current profile. Dynamo effects determine the safety factor profile of the final 3D equilibrium, with important consequences on MHD stability and transport. We compare experimental results from multiple machines (RFX-mod, MST, AUG, DIII-D) and nonlinear MHD modelling. Two paradigmatic cases of helical state formation are considered and common physics is identified, by direct measurements of dynamo effects and MHD simulations: spontaneous formation in high-current reversed-field pinch (RFP) plasmas [1] and the hybrid scenario in high-beta tokamak plasmas [2]. Helical cores form in both cases, either spontaneously via saturation of MHD modes, or due to the marginally-stable ideal MHD response to external 3D fields. Direct measurements of the dynamo emf associated to 3D plasma distortions will be presented for a database of helical RFP plasmas from RFX-mod and MST, covering a wide range of plasma parameters. Similar measurements were also done in helical states forming in response to external 3D fields in Ohmic RFX-mod tokamak plasmas and in DIII-D high-beta hybrid plasmas. Experimental results qualitatively agree with nonlinear MHD modelling performed with the codes SpeCyl [3], PIXIE3D [4], and NIMROD [5]. They indicate that central current is redistributed by a dominantly electrostatic MHD dynamo. The underlying physics common to RFP and tokamak is thus revealed: a helical core displacement modulates parallel current density along flux tubes, which requires a helical electrostatic potential to build up, giving rise to a helical dynamo flow. Similar results were also recently obtained with the M3D-C1 code [6]. [1] R. Lorenzini et al., Nature Phys. 5, 570 (2009); [2] T.C. Luce et al., Nucl. Fusion 54, 013015 (2014); [3] D. Bonfiglio et al., Phys. Rev. Lett. 94, 145001 (2005); [4] D. Bonfiglio et al., Plasma Phys. Control. Fusion 57, 044001 (2015); [5] J.R. King, C.R. Sovinec, V.V. Mirnov, Phys. Plasmas 19, 055905 (2012); [6] S.C. Jardin et al., Phys. Rev. Lett. 115, 215001 (2015).
Speaker: Paolo Piovesan (Consorzio RFX)
• 16:10 16:40
Coffee Break 30m
• 16:40 18:45
Overview 5: Magnetic Fusion OV/5
Convener: Dr Sergei Lebedev (Ioffe Physical-Technical Institute, Russian Academy of Sciences)
• 16:40
3-D effects on transport and plasma control in the TJ-II stellarator 25m
Recent improvements in diagnostics and operation have led to better understanding of 3-D effects on transport and plasma control in the TJ-II stellarator. Impurity transport: Direct measurments of electrostatic potential variations within the same magnetic flux surface in ECRH plasmas are presented. Calculations show that such asymmetries affect impurity accumulation. The asymmetry value and its observed dependency on the electric field are reproduced by neoclassical MC calculations. The dependence of the impurity confinement time on charge and mass has also been studied. Experiments have shown evidence of the influence of ECRH on turbulent mechanisms, increasing both the fluctuation level and the amplitude of Long-Range-Correlations as proxy of Zonal Flows (ZF), as well as affecting NC radial electric fields. Momentum transport and electromagnetic effects: Radial electric fields, ZF-like structures, time memory and radial correlations are modulated by low order rationals. It is shown that magnetic oscillations associated with rational surfaces play an key role in confinement transitions. Furthermore, evidence of the mutual interaction of NC and turbulent mechanisms in qualitative agreement with GK simulations is presented. Innovative power-exhaust scenarios using liquid metals: Novel solutions for plasma facing components based on liquid metals like Li and Sn/Li alloys have been developed. Biasing of Li limiters with respect to carbon ones has evidenced the role of the secondary electron emission of plasma exposed surfaces. Plasma stability studies: It has been shown that a reduction of magnetic well has a direct impact on fluctuations without reducing plasma confinement drastically, suggesting that Mercier stability calculations are missing some stabilization mechanisms. Plasma fuelling experiments and neutral dynamics: First core plasma fuelling experiments using a cryogenic pellet injector system are presented. The radial redistribution of particles can be understood qualitatively from NC predictions. First results on the impact of neutral fluctuations on the observed turbulent structures will be reported. Role of ECRH and iota profile on fast ion confinement: Results show that ECRH and iota-profile are potential tools for AE control. Coherent modes in NBI-heated plasmas are explained as global (GAE) and discrete shear-AEs induced by magnetic islands.
Speaker: Prof. Francisco Castejón (CIEMAT)
• 17:05
Overview of First Results from NSTX-U and Analysis Highlights from NSTX 25m
The National Spherical Torus Experiment (NSTX) has undergone a major upgrade, and NSTX Upgrade (NSTX-U) is now the most capable Spherical Torus/Tokamak (ST) in the world program. NSTX-U mission elements include: exploring unique ST parameter regimes to advance predictive capability for ITER and beyond, developing solutions for the plasma-material interface challenge, and advancing the ST as a possible Fusion Nuclear Science Facility or Pilot Plant. NSTX-U has two major new tools including a new central magnet and new 2nd more tangential neutral beam injector (NBI). Plasma control commissioning and scenario development has proceeded rapidly on NSTX-U. Diverted plasmas with IP = 0.8MA, BT = 0.6T, and tau-pulse ~ 1s are obtained routinely, and sustained H-mode plasmas have been accessed with 2.5MW of NBI heating power. Peak parameters achieved during the first run-month of NSTX-U plasma operation include: NBI power ~4MW, IP = 1MA, stored energy ~ 200kJ, beta-N ~ 4, kappa ~ 2.2, tau-E-tot > 50ms, tau-pulse ~ 1.7s, and a 50% increase in pulse-length from n=1 error field correction. Expected results from the first run campaign include assessments of: core and pedestal confinement at lower collisionality via 60% higher field and current than NSTX, fast-ion confinement and current drive from the new 2nd NBI, and stability and control of high-kappa and high beta-N plasmas. Extensive analysis of NSTX results continued including novel analysis of: edge turbulence data during the L-to-H-mode transition, heat flux footprint narrowing with increasing amplitude of edge-localized modes, and gyrokinetic modeling of core turbulence from dissipative trapped electron mode and electron temperature gradient modes. Further, a unified kinetic resistive wall mode physics model has been developed, and Massive Gas Injection valves similar to proposed ITER valves will be tested on NSTX-U. Lastly, a new method for determining the saturation level for Alfvén Eigenmodes has been developed, and SOL power losses for RF heating modeled and interpreted with the AORSA code. Results from the first research campaign of NSTX-U will be presented, initial comparisons between NSTX-U and NSTX results described, and NSTX analysis highlights presented.
Speaker: Dr Jonathan Menard (Princeton Plasma Physics Laboratory)
• 17:30
Overview of recent physics results from MAST 25m
New results from MAST will be presented that focus on validating models in order to extrapolate to future devices. Particular attention will be given to the areas of scenario development, fast particle physics and plasma exhaust. Understanding filamentary transport across the scrape off layer is a key issue for the design and operation of future devices as it is crucial in determining the power loadings to the divertor and first wall of the machine. A detailed characterisation of the MAST Scrape Off Layer has been performed including results from new diagnostics giving plasma potential and ion temperature measurements. Detailed studies have revealed how filament characteristic are responsible for the broadening of the midplane density profile. These measurements have been compared to extensive modelling, including 3D effects on filaments dynamics with the BOUT++ code, and benchmarking the SOLPS code. Impurity transport studies have shown how the balance between neoclassical and anomalous transport leads to carbon and nitrogen being screened from the core plasma compared to helium which is peaked at the centre. These results, combined with SOLPS modelling, suggest that a stable detachment region can be produced if the impurity puffing is localised. Measurements from a Doppler Backscattering system combined with GS2 simulations have shown that both micro-tearing modes (MTMs) and electron temperature gradient (ETG) modes can be unstable at the top of the pedestal, along with kinetic ballooning modes at the bottom of the steep gradient region. The experimental observations of the relative amplitudes and wavelengths of the density and magnetic field fluctuations at the top of the pedestal are more similar to the linear characteristics of the ETG than the MTM. Comprehensive measurements from a suite of diagnostics on MAST have shown the effect that core MHD modes and resonant magnetic perturbations (RMPs) have on the confinement and redistribution of fast ions arising from neutral beam injection (NBI). Subsequent experiments on MAST demonstrated that by vertically displacing the plasma to achieve off-axis NBI fast ion injection or by changing plasma density or NBI power to vary the fast ion pressure gradient the redistribution could be mitigated.
Speaker: Dr Andrew Kirk (Culham Centre for Fusion Energy)
• 17:55
H-mode and Non-Solenoidal Startup in the Pegasus Ultralow-A Tokamak 25m
Studies at near-unity aspect ratio offer unique insights into the high confinement (H-mode) regime and support development of novel startup scenarios. Ohmic H-mode operation has been attained at A < 1.3. Edge plasma parameters permit probe measurements of the edge pedestal, including the local current density profile, with high spatial and temporal resolution. H-mode plasmas have standard L-H transition phenomena: a drop in D_alpha radiation; the formation of pressure and current pedestals; field-aligned filament ejection during ELMs; and a doubling of energy confinement time from H_98 ~ 0.5 to ~1. The L-H power threshold P_LH increases monotonically with n_e, consistent with the ITPA08 empirical scaling used for ITER and the theoretical FM3 model. Unlike at high A, P_LH is comparable in limited and single-null diverted topologies at A ~ 1.2, consistent with FM3 predictions. The magnitude of P_LH exceeds ITPA scalings by an order of magnitude, with P_LH/P_ITPA08 increasing as A approaches 1. Multiple n modes are observed during two classes of ELMs, consistent with excitation of multiple peeling-ballooning modes. Small, Type III-like ELMs occur at P_OH ~ P_LH with n <= 4. Large, Type-I-like ELMs occur with P_OH > P_LH and intermediate 5 < n < 15. Helical edge current injection appears to suppress Type III ELM activity. J_edge(R,t) measurements across single ELMs show the nonlinear generation and expulsion of current-carrying filaments during the ELM crash. Local Helicity Injection (LHI) offers a nonsolenoidal tokamak startup technique. Helicity is injected via current sources at the plasma edge. A circuit model that treats the plasma as a resistive element with time-varying inductance reasonably predicts I_p(t). The electron confinement governs the power balance. Initial measurements show peaked T_e and pressure profiles, which are comparable to Ohmic-like transport or moderately stochastic confinement. Extrapolation suggests I_p ~ 1 MA may be achievable in NSTX-U. Resistive MHD simulations suggest I_p is built from current rings injected during reconnection between unstable helical current streams. Several experimental observations support this model: imaging of the merging current streams; n=1 MHD activity and discrete current stream localized in the plasma edge; and anomalously high impurity ion heating in the edge region.
Speaker: Prof. Raymond Fonck (University of Wisconsin-Madison)
• 18:20
Overview of Spherical Tokamak Research in Japan 25m
Nationally coordinated research on spherical tokamak (ST) is being conducted in Japan, to strengthen the scientific basis and to broaden future options of ST applications. The research themes to concentrate on are (1) the physics of very high beta plasmas, (2) development of start-up, current drive, and control techniques without the use of the central solenoid (CS), and (3) demonstration of very long pulse operation and the study of steady-state issues. Research elements are developed on several devices optimized for each objective. The basic mechanism of tokamak plasma formation by ECW/EBW was investigated on LATE. The tokamak configuration with closed flux surfaces is formed spontaneously when the equilibrium current changes from the vertical charge separation current to the toroidal return current. Highly over-dense plasmas have been produced, indicating mode conversion to EBW. A maximum plasma current of 66 kA was achieved using 28 GHz on QUEST. Plasma current start-up by LHW is being investigated on TST-2. The most efficient ramp-up was achieved by the capacitively-coupled combline antenna, which excites a traveling LHW with a sharp wavenumber spectrum and high directionality. Experiments with a top-launch antenna, expected to improve single-pass absorption and increase current drive efficiency, have started. The formation of closed flux surfaces by transient coaxial helicity injection (CHI) was verified by internal magnetic probe measurements on HIST. A stable closed flux formation was achieved by high bias flux operation, and the validity of helicity balance was confirmed. CHI electrodes were installed on QUEST under US-Japan collaboration. An RF-driven long pulse discharge of up to 810 s has been achieved on QUEST. Operation with hot metal wall has started, with the aim to control particle recycling by active wall temperature control. Compact toroid injection is being developed as an advanced fueling method. High-power reconnection heating of ST plasmas using axial merging of two ST plasmas was demonstrated in TS-3, TS-4 and UTST. Collaboration on the MAST device has demonstrated that reconnection heating can be extended successfully to larger scale and higher magnetic field. ST plasma stability improvement was accomplished by applying a helical field in TOKASTAR-2, an ST-helical hybrid device equipped with helical field coils.
Speaker: Prof. Yuichi Takase (University of Tokyo)
• Wednesday, 19 October
• 08:30 10:15
In - Vessel Components: FIP/2
Convener: Dr anatoli Krasilnikov (Director Institution @Project center ITER")
• 08:30
Progress of Qualification Testing for Full-Scale Plasma-Facing Unit Prototype of Full Tungsten ITER Divertor in Japan & Progresses on WEST Platform Construction towards First Plasmas 20m
A. R&Ds for starting operation with a full- tungsten (W) ITER (INB-174) divertor have been enhanced by recommendation of the ITER council since 2011. Japan Atomic Energy Agency (JAEA) as Japanese Domestic Agency (JADA) and the ITER organization (IO) have been actively working on the development and demonstration on the full-W ITER divertor under the framework of the task agreement. JAEA is in charge of technology development and demonstration for manufacturing the Outer Vertical Target (OVT) together with Japanese industries. In 2013, as the first phase of the qualification program, JAEA demonstrated the armour heat sink bonding technology with small-scale mock-ups. A high heat flux (HHF) testing for the mock-ups was carried out in the ITER divertor test facility in Efremov Institute, Russia. JAEA succeeded in demonstrating the durability of the W monoblock joint to the Cu-alloy cooling tube against the heat load of 10 MW/m2 × 5000 cycles and 20 MW/m2 × 1000 cycles which are three times higher than a requirement (300 cycles). This result provided one of sufficient materials for the decision to start with the full-W ITER divertor in the baseline. Since 2014, as the second phase, the full-scale plasma-facing unit (PFU) prototypes have been manufactured to demonstrate the scale-up manufactural technology. In this paper, JAEA reports progress of R&Ds on the full-scale PFU prototypes of a full- W ITER Divertor OVT. Under a framework of a W divertor qualification program, JAEA manufactured 7 full-scale PFUs as prototypes. Through the manufacturing, (i) all joint surfaces in four PFUs with a casting Cu interlayer successfully passed the ultrasonic testing and (ii) the surface profile in target part of PFUs stayed within a tolerance. (iii) Moreover JAEA succeeded in demonstrating a durability for the HHF testing of the repetitive heat load of 10 MW/m2 × 5000 cycles and 20 MW/m2 × 1000 cycles under close collaboration with the IO and the Efremov Institute. These results demonstrated the ability of Japanese industries to produce the PFU of full-W ITER divertor enough to meet the technical requirements. B. The WEST platform, which is a major evolution of Tore Supra towards a steady-state tungsten diverted tokamak, is targeted at minimizing risks for ITER divertor procurement and operation. This paper presents an overview of the status and relevant technical issues for the new platform. At the time of the writing, the 4 meter diameter thick casing of the upper and lower divertor in-vessel coils have been manufactured, assembled inside the torus and accurately positioned. The in-situ winding of the water cooled copper conductor requiring about 140 brazing is underway. The complex assembly sequence as well as the resin epoxy impregnation has been simulated and validated on a full scale mock-up. The power supplies which will feed the divertor coils have been produced. Factory acceptance test have been performed and the two power supplies will be installed at Cadarache this summer. The procurement of the ITER-like divertor plasma facing units (PFUs), using the ITER tungsten monoblock technology, is ongoing in collaboration with the European and Japanese Domestic Agencies in charge of providing ITER divertor vertical targets. Prototypes are in preparation and will be tested in WEST before launching series production. Tungsten-coated technologies have been developed and qualified on various substrates to cover the other high heat flux plasma facing components. In particular, inertial graphite PFUs with improved CMSII tungsten coating (15 µm) have been qualified and manufactured in order to complement the ITER-like prototypes of the WEST lower divertor for the first phase of operation. The new CW ELM-resilient ICRH antennas are in manufacturing and the first one will be assembled in spring 2016. The existing LHCD launcher front faces have been reshaped to match the new plasma geometry. The overall diagnostic layout is finalized. Key diagnostics are being upgraded to allow for a proper monitoring of the divertor plasma facing units, the tungsten sources and transport. A new plasma control system prototyping ITER requirements is being implemented. WEST is presently scheduled to be operational in late 2016.
Speaker: Dr Yohji Seki (National Institutes for Quantum and Radiological Science and Technology)
• 08:50
Design and R&D Progress of Chinese HCCB TBS Program 20m
The current design of Chinese Heilum-cooled Ceramic Breeder Test Blanket Module (HCCB TBM) with 1×4 configuration scheme, it includes 4 independent breeding sub-modules with 10mm gap (for thermal expansion) between each other along the poloidal direction. Tthese sub-modules are connected with a big back plate containing auxiliary connection pipes for coolant and purge gas, shear keys and flexible supports to form a whole TBM. And the TBM is connected to the Helium Cooling System (HCS) and Tritium Extraction System (TES) by means of pipes in the back plate. A 3-D neutronics calculation for the updated TBM module design has been completed. Preliminarily a simplified analysis model for the sub-module of TBM is adopted. In order to validate the design of CN HCCB TBS (module and system), a lot of R&Ds on materials has been performed according to technical requirements. The RAFM material of Chinese Low Activation Ferritic (CLF-1) steel has been developed and is scaled up to 5 ton ingot, which is used for the structure material certification. At the same time, 1 dpa neutron irradiation test in high flux test reactor and its PIE experiment has been performed. Based on the CLF-1 steel, some mock-ups have been fabricated by the different techniques and tested. A 1/3-size mockup of TBM module is under fabrication and will be tested soon. The fabrication techniques for the functional materials, such as beryllium and Li4SiO4, have been also developed and the related properties have been obtained. The fabrication of back plate system could be the current largest challenge for HCCB TBM. Several welding technologies and manufacture process are investigated on different size plates, including Laser Welding (LW) process for the BP and the FW, HIP welding for FW. The welding of the CLF-1 with the 316 L(N) IG (ITER Grade) is dissimilar welding process for assembling TBM module and shield block. For the manufacturing solution of FW, Hot Isostatic Pressing (HIP) is a realistic process, and the base welding experimental researches for the HIP joining of RAFMs CLF-1 are being carried out to prepare for the next forming welding practices of FW at present. A small-type Helium Gas Testing Loop (HGTL) is under construction, which will be used for the future component testing and operation testing. The design temperature is 300℃ and the pressure of He gas is 8MPa.
Speaker: Mr Kaiming Feng (Southwestern Institute of Physics)
• 09:10
Lessons learned for the Breeding Blanket designers from the design development of the European Test Blanket Module Systems (He, Tritium, Liquid Metal Systems) 20m
The general objective of the ITER TBM Program is to provide the first experimental data on the performance of the breeding blankets in the integrated fusion nuclear environment. Such data are essential to design and predict the performance of DEMO and future fusion reactors. To achieve this objective, the TBM programme will have to: • test and validate technologies and materials in a fusion relevant environment in view of their further development for DEMO and power plants • validate and qualify predictive tools for the design of the breeding blankets in DEMO and power plants To comply with this mission, the TBM programme will cover the full lifetime of ITER operation, with testing of series of TBM specifically instrumented for maximizing the ROX (Return on Experience) for each ITER operational phase in view of the DEMO breeding blanket design. The design of the European Test Blanket Systems (TBS), Helium Cooled Lithium Lead (HCLL) and Helium Cooled Pebble Bed (HCPB), has concluded its conceptual phase. Particularly during the development of the design of the TBM systems (also known as “TBM ancillary systems”), several lessons learned can be already now considered very important for the designers of the DEMO breeding blanket. They deal with: - the impact of the safety requirements on the design of systems and components - the impact of the licensing procedure on design requirements and implementation - the definition and implementation of the main safety functions - the compliance with the nuclear pressure equipment (ESPN) regulation - the functional analysis of the different systems and components - the selection of technological solutions for the TBM systems which appear relevant for the breeding blanket design - major integration issues, like the management of the tritium contamination due to permeation and leakage. After a synthetic recall of the main design features of the HCLL and HCPB-TBS, all above mentioned topics are discussed in this paper, addressing the analysis onto the original inputs and ROX for the designers of a breeding blanket for DEMO from the current TBM systems design experience.
Speaker: Dr Italo Ricapito (Fusion for Energy)
• 09:30
Development of Sensors for High-Temperature High-Pressure Liquid Pb/ Pb-16Li Applications 20m
Liquid Lead Lithium (Pb-16Li) is of primary interest as one of the candidate materials for coolant fluid and tritium breeder in liquid metal blanket concepts relevant to fusion power plants. For effective and reliable operation of such high temperature liquid metal coolant systems, monitoring and control of critical process parameters like pressure, level, temperature and flow is essential. However, high temperature operating conditions coupled with the corrosive nature of Pb-16Li severely limits the application of commercially available diagnostic tools. This paper illustrates indigenous test facility designs and experimental methods used to develop non-contact configuration radar level sensor and wetted configuration diaphragm seal pressure sensors for high temperature, high pressure liquid Pb and Pb-16Li. Calibration of these sensors at high temperature between 380C-400C and high pressure upto 10 bar was performed. Reliability and performance validation were achieved by continuous long duration testing of sensors in liquid Pb and liquid Pb-16Li environment for over 1000 hour. Estimated error for radar level sensor lies within ±10 mm and estimated error for pressure sensors lies within 1.1% of calibrated span over the entire test duration. Results obtained and critical observations from these tests are presented in this paper.
Speaker: Mr Abhishek Saraswat (InIPR)
• 09:50
Liquid lithium loop system to solve challenging technology issues for fusion power plant 20m
Steady-state fusion power plant designs present major divertor technology challenges, including high divertor heat flux both in steady-state and during transients. In addition to these concerns, there are the unresolved technology issues of long term dust accumulation and associated tritium inventory and safety issues. It has been suggested that radiation-based liquid lithium (LL) divertor concepts with a modest lithium-loop could provide a possible solution for these outstanding fusion reactor technology issues, while potentially improving the reactor plasma performance. The application of lithium (Li) in NSTX resulted in improved H-mode confinement, H-mode power threshold reduction, and reduction in the divertor peak heat flux while maintaining essentially Li-free core plasma operation even during H-modes. These promising results in NSTX and related modeling calculations motivated the radiative liquid lithium divertor (RLLD) concept and its variant, the active liquid lithium divertor concept (ARLLD), taking advantage of the enhanced Li radiation in relatively poorly confined divertor plasmas. To maintain the LL purity in a 1 GW-electric class fusion power plant, a closed LL loop system with a modest circulating capacity of ~ 1 liter/second (l/sec) is envisioned. We examined two key technology issues: 1) dust or solid particle removal and 2) real time recovery of tritium from LL while keeping the tritium inventory level to an acceptable level. By running the LL-loop continuously, it can carry the dust particles and impurities generated in the vacuum vessel to the outside where the dust / impurities can be removed by relatively simple dust filter, cold/hot trap and/or a centrifugal separation systems. With a 1 l/sec LL flow, even a small 0.1% dust content by weight (or 0.5 g per sec) means that the LL-loop could carry away nearly 16 tons of dust per year. In a 1 GW-electric (or ~ 3 GW fusion power) fusion power plant, about 0.5 g / sec of tritium is needed to maintain the fusion fuel cycle assuming ~ 1 % fusion burn efficiency. It appears feasible to recover tritium (T) in real time from LL while maintaining an acceptable T inventory level. Laboratory tests are also planned to investigate the Li-T recover efficiency with the SCT concept and also to assess the viability of the centrifugal Li-T separator with consultation with the manufacturer.
Speaker: Dr Masayuki Ono (PPPL/Princeton University)
• 08:30 12:30
Poster 3: P3
• 08:30
Achievement of Field-Reversed Configuration Plasma Sustainment via 10 MW Neutral-Beam Injection on the C-2U Device 4h
The world’s largest compact-toroid device, C-2, has been upgraded to C-2U at Tri Alpha Energy to achieve sustainment of field-reversed configuration (FRC) plasmas by neutral-beam (NB) injection (NBI) and edge biasing [1,2], and the C-2U experiment is characterized by the following key system upgrades: increased total NB input power from ~4 MW (20 keV hydrogen) to 10+ MW (15 keV hydrogen) with tilted injection angle; enhanced edge-biasing capability inside of each end-divertor for boundary/stability control. C-2U experiments with those upgraded systems have successfully demonstrated dramatic improvements in FRC performance. As anticipated, there are strong effects of the upgraded NB injectors on FRC performance such as: (i) rapid and strong accumulation of fast ions (about a half of initial thermal pressure replaced by fast-ion pressure); (ii) fast-ion footprint largely determines FRC dimensions; (iii) double-humped electron density and temperature profiles; (iv) FRC lifetime and global plasma stability scale strongly with NBI power; and (v) plasma performance correlates with NB pulse duration in which diamagnetism persists several milliseconds after NB termination due to accumulated fast ions. The key accomplishment on C-2U is sustainment of advanced beam-driven FRCs with a macroscopically stable and hot plasma state for up to 5+ ms, limited only by hardware and stored energy constraints such as the NBs’ pulse duration (flat-top ~8 ms) and current sourcing capability of end-on plasma guns. Furthermore, plasma diamagnetism in the best discharges has reached record lifetimes of over 11 ms, timescales twice as long as C-2. In this regime fast ions are well trapped and nearly classically confined, suppressing broadband magnetic turbulence as well as enhancing fusion reactivity via beam driven collective effects. Density fluctuations near the separatrix and in the scrape-off layer have also been dramatically suppressed by a combination of NBI and E×B shearing via plasma-gun edge biasing, thereby improving confinement properties. The demonstrated sustainment of beam-driven FRCs in C-2U is an extraordinary achievement for the FRC and innovative confinement concepts communities, and may lead to intriguing possibilities for fusion reactors. [1] M. Tuszewski et al., Phys. Rev. Lett. 108, 255008 (2012). [2] M.W. Binderbauer et al., Phys. Plasmas 22, 056110 (2015).
Speaker: Dr Hiroshi Gota (Tri Alpha Energy, Inc.)
• 08:30
Adaptive Real-Time Pedestal Control for DIII-D and Prospects for ITER 4h
A comprehensive adaptive real-time (rt) ELM control system that exploits key properties of ELM physics, Resonant Magnetic Perturbation (RMP) ELM suppression physics, and an extensive set of diagnostic inputs to make rt decisions about the control of multiple actuators to sustain ELM suppression/mitigation is demonstrated at DIII D. The control experiments showed the path dependence and hysteresis of plasma recovery: even for the same final perturbing 3D currents, starting with higher initial 3D currents leads to lower recovery down the path. This demonstrates the need for a control system to keep the ITER RMP perturbations close to the ELM suppression threshold at all times. The development at DIII-D initiates progress toward adaptive pedestal control, and includes pedestal profile control as well as ELM suppression/mitigation. 3D coil configuration and phasing for RMP ELM suppression is adjusted in real-time based on SURFMN calculations of the vacuum edge pitch-resonant, and kink-resonant harmonics of the applied 3D magnetic perturbation and offline IPEC data. The amplitude of the 3D coils is regulated to achieve a given ELM frequency (or none) using ELM detection based on the D_α measurements from the divertor region. For pedestal control, the Plasma Control System (PCS) acquires rt Thomson scattering diagnostic data and fits the pedestal width/height for temperature and density profiles. Based on the Thomson fits, PCS regulates the pedestal density by adjusting the gas-puffing rate to increase particle source and RMP density “pump-out” to reduce it. Real-time pedestal stability boundary calculation using a neural network based on EPED1 runs, and a real-time pellet injection control for turn on/off timing and ELM frequency are under development. These developments at DIII-D pave the way for ITER adaptive pedestal control.
Speaker: Prof. Egemen Kolemen (Princeton University)
• 08:30
Analysis and prediction of momentum transport in spherical tokamaks 4h
The inward momentum convection or “pinch” observed in many tokamaks can be explained by the Coriolis drift mechanism, with relatively good quantitative agreement found with gyrokinetic predictions of the ion temperature gradient (ITG) instability. Here we attempt to validate this model over a broader range of beta and aspect ratio by extending into the spherical tokamak (ST) plasma regime using data from NSTX and MAST. Previous perturbative measurements in NSTX H-modes have indicated the existence of an inward momentum pinch with a magnitude similar to that observed in conventional aspect ratio tokamaks. However, linear gyrokinetic simulations run for these cases predict the microtearing mode, which only transports electron energy, is the dominant micro-instability in the region of interest due to the relatively large plasma beta. Although weaker, there is also evidence of a variety of unstable electrostatic and electromagnetic ballooning modes. Quasi-linear calculations for all of these ballooning modes, assuming they contribute substantially to the momentum transport, predict a pinch that is small or directed outward, in contradiction to the experimental results. Additional scans show that the weak pinch is a consequence of how both electromagnetic effects (at relatively large beta) and low aspect ratio influence symmetry-breaking of the instabilities. To minimize electromagnetic effects, similar experiments were performed in MAST L-mode plasmas at relatively low beta using the time-dependent rotation response after the removal of a short n=3 applied magnetic field perturbation. The inferred inward pinch is similar in magnitude to conventional tokamaks and the NSTX H-modes. However, linear gyrokinetic simulations indicate that even for low beta L-modes the predicted momentum pinch is relatively small and cannot reproduce the large experimentally inferred pinch. Based on the above observations and simulations, the Coriolis pinch mechanism predicted from local, linear gyrokinetic theory does not appear to explain perturbative momentum transport at low aspect ratio. Other mechanisms neglected thus far are being investigated as possible solutions to the apparent discrepancy, including nonlinear effects, perpendicular ExB shear driven transport, centrifugal effects and profile shearing.
Speaker: Dr Walter Guttenfelder (Princeton Plasma Physics Laboratory)
• 08:30
Applying the new principles of plasma self-organization to tokamak 4h
Understanding sustainment of stable equilibria with helicity injection in HIT-SI has led to a simple picture of several tokamak features. Perturbations cause a viscous-like force on the current that flattens the j/B profile, which sustains and stabilizes the equilibrium. An explanation of the mechanism is based on the two properties of stable, ideal, two-fluid, magnetized plasma. First, the electron fluid is frozen to magnetic fields and, therefore, current flow is also magnetic field flow. Second, for a stable equilibrium the structure perpendicular to the flux surface resists deformation. This mechanism provides an explanation for the level of field error that spoils tokamak performance, the rate of poloidal flux loss in argon-induced disruptions in DIII-D, why transport barriers depend on the E X B shearing rate, and why a divertor may help in forming a pedestal. This paper is based upon work supported by the U.S. Department of Energy Office of Science, Office of Fusion Energy Sciences under Award No. DE-FG02-96ER54361.
Speaker: Prof. Thomas Jarboe (University of Washington,)
• 08:30
Characteristics of turbulent transport in flux-driven toroidal plasmas 4h
Profile stiffness and intermittent bursts are the basis in understanding L-mode plasmas. However, why and how these different processes coexist and regulate the transport have not been fully clarified. Here, we presented an overall picture of flux-driven ITG turbulent transport which reveals profile stiffness with self-similarity and SOC type intermittent bursts simultaneously using a flux-driven gyrokinetic code by incorporating with statistical analyses. We found that the transport is regulated by four non-diffusive processes, (1) radially localized fast time scale avalanches, (2) radially extended global bursts, (3) slow time scale avalanches with stair-case, (4) transport with long range time correlation. Among them, the process (2) is the key, which results from the instantaneous formation of radially extended ballooning-type structure with long radial correlation length from meso- to macro-scale. Such structures are disintegrated and damped by self-generated zonal flows while the repetitive occurrence of such structure provides a strong constraint on the profile causing stiffness. Zonal flows produced by such global modes becomes the origin of the shear layer of radial electric field and associated pressure corrugation, referred to as ExB staircase. Since they are excited near both edges of global mode, the interspace is determined approximately by the size of the global mode. The staircase is found to evolve dynamically coupled with successive excitation of global mode. This process causes a long time scale breathing in transport and plays a role in sweeping out corrugations appeared on the self-organized stiff profile. To obtain a unified view of transport, we study the spatio-temporal characteristics statistically. Quasi-steady baseline of transport is due to eddies from micro- to meso-scale, which follow a power law scaling, while the busty part to global eddies which release large amount of free energy as a non-power law tail component. The spatio-temporal linkage of such different non-diffusive processes leads to a new turbulent state dominated by long range correlation in time and space. Finally, we found that the magnetic shear is a key parameter, so that the profile stiffness with specific function form and intermittency have revealed in moderate magnetic shear plasmas while weaken in those with weak and reversed magnetic shear.
Speaker: Prof. Yasuaki Kishimoto (Kyouto University)
• 08:30
Co- and Counter Current Rotation in Tore Supra LHCD Plasmas: Neoclassical and Turbulent Transport Processes 4h
Lower hybrid (LH) wave effect on toroidal plasma rotation in L-mode Tore Supra plasmas has been analyzed in more than 50 plasma discharges, with LH input power P_LH up to 4.8MW, plasma current Ip up to 1.4MA, line integrated density n_l up to 6 × 10^19m−2, B_T = 3.8T, and a significant ripple amplitude (up to 5% at the plasma boundary) which makes ripple-induced momentum non negligible. At low plasma current (Ip < 0.95 MA), the rotation change is in the co-current direction and impacts the whole rotation profile. At higher plasma current, an opposite trend is observed, the core plasma rotation incrementing in the counter-current direction, the profile being affected up to r/a < 0.6 only. In both low and high Ip cases, rotation increments are found to increase with P_LH (Fig. 1). Moreover, when Ip increases, at fixed LH power (P_LH = 4.8MW) and plasma density (n_l = 3.8×10^19 m−2), the rotation increases in the counter-current direction, switching from co- to counter-current direction at Ip ∼ 0.95MA. Theoretical investigations show that the rotation evolution results from the competition of different contributions. At high plasma current, the rotation evolution in LHCD plasmas is controlled by the neoclassical friction force due to the trapped ions in banana trajectories through the toroidal diamagnetic velocity. This force results in a counter-current rotation increment as observed in Tore Supra experiments. At low plasma current, the rotation is dominated by momentum turbulent transport when the LH waves are applied. The Reynolds stress grows strongly (through q profile effect) comparing to the high plasma current case, and acts as a co-current force through the residual stress contribution.
Speaker: Dr Christel Fenzi (CEA)
• 08:30
Compact Fusion Energy based on the Spherical Tokamak 4h
Tokamak Energy Ltd, UK, is developing spherical tokamaks (STs) using High Temperature Superconductor (HTS) magnets as a route to fusion power based on high gain, small size power plants. The paper presents an overview of the continuing advances in technology and modeling, which, together with key engineering developments, support this concept. The ST achieved recognition as a high beta plasma research device with many desirable properties. To date it has been shown to be viable as a compact fusion neutron source /Component Test Facility, but not as a viable route to fusion power because of the inefficiency of driving high current in a slender copper centre column. However, significant new advances change the situation substantially. In particular the latest YBCO High Temperature Superconductors (HTS) are now proven to be able to carry large currents in strong magnetic fields in a very compact centre post. Innovative designs of neutron shielding indicate that relatively thin shields could give sufficient protection to an HTS core under significant neutron bombardment, and new engineering designs of the HTS centre column indicate tolerable stresses. Further, recent modeling has shown that, under reasonable operating conditions, tokamak pilot plants and reactors have a power gain Qfus that is only weakly dependent on size, but depends on fusion power as Qfus ~ Pfus H2 where H is the confinement factor relative to ITER empirical scalings. For several reasons - including use of a beta-independent confinement scaling demonstrated as being more appropriate than the IPB98y2 scaling - STs should achieve a specified Qfus at considerably reduced Pfus, reducing wall and divertor loading. These innovations introduce the possibility of a superconducting ST Pilot Plant which can be much smaller than the designs previously considered. An example is given of a low-cost compact fusion pilot plant based on an ST of major radius 1.35m and fusion power 150-200MW with Qfus = 1-10, dependent on the confinement achieved. Higher gain versions would be developed from the insight gained and a fusion power plant would then consist of 10 or so of these modules. This approach offers significant advantages, not least that the small scale of the prototype modules should lead to rapid development at relatively low cost.
Speaker: Dr Alan Costley (Tokamak Energy Ltd)
• 08:30
Compact Toroid Injection Fueling on a Large-sized Field-Reversed Configuration 4h
A repetitively driven compact toroid (CT) injector has been developed for large-sized field-reversed configuration (FRC) facility of the C-2/C-2U primarily for refueling. Pursuit of the FRC as fusion reactor is motivated by highly favorable technological features: extremely high beta (>50%), a natural divertor, and axial mobility allowing separation of start-up and confinement functions. Recently, high confinement performance of FRC has been achieved on the C-2/C-2U facility by neutral beam injection (NBI). However, development of effective fueling method remains as a significant task of FRC fusion reactor core. A CT is formed and injected by a magnetized coaxial plasma gun (MCPG) exclusively developed for the C-2/C-2U FRC. It consists of a set of coaxial cylindrical electrodes, a bias coil and four gas injection ports which are arranged tangentially on the outer electrode. The inner electrode is coated by tungsten to reduce impurity influx. A plasma ring is generated within a gap between the electrodes and is accelerated by Lorenz self-force. During this acceleration process, toroidal current is induced by a poloidal flux interlinked with the plasma ring. Then, the magnetized spheromak-like CT is ejected from the MCPG. To refuel the particles of long-lived FRCs, multiple CT injection is required. Thus, a multi-stage discharge circuit has been developed for multi-pulsed CT injection. Drive frequency of this system can be adjusted up to 1 kHz and the number of CT shots per injector is 2; the system can be further upgraded for larger number of injection pulses. The developed MCPG has achieved supersonic ejection velocity in the range of ~100 km/s. Key plasma parameters of electron density, electron temperature and the number of particles are ~ 5 × 10^21 m^-3, ~ 40 eV, and 0.5 - 1.0 × 10^19, respectively. In this project, single and double pulsed CT injection fueling have been conducted on the C-2/C-2U facility by two CT injectors. The CT injectors are mounted 1 m apart on the vicinity of midplane. To avoid disruptive perturbation on the FRC, the CT injectors have been operated at the lower limit of particle inventory. The experiments demonstrated successful fueling with significant density build-up of 20 - 30% of the FRC particle inventory per single CT injection without any deleterious effects on the C-2/C-2U FRC.
Speaker: Prof. Tomohiko Asai (Nihon University)
• 08:30
Confinement and stability of the ITER Baseline Scenario in DIII-D 4h
Analysis of ~180 ITER Baseline Scenario (IBS) demonstration discharges in DIII-D provides insight into the cause of the 2/1 disruptive instabilities that limit the duration of these plasmas. Raw MSE data and detailed equilibrium reconstructions show that a larger current profile gradient in the region of the q=2 surface characterises the unstable cases, providing the drive for the 2/1 tearing mode onset. Rotation measurements indicate that lower differential rotation at the marginal stability point constitutes the additional separating factor for part of the unstable shots at low injected torque. The current profile is observed to evolve with the plasma rotation, due to modifications in the pedestal transport, and initial transport modeling shows that time dependent predictive simulations can capture these changes. The approach to the instability at low rotation is observed in Active MHD Spectroscopy (AMS) measurements of the plasma response. Drift-kinetic modelling of these measurements indicates that non-ideal effects are significant despite the relatively low βN of these plasmas. The inclusion of collisionality and resistivity is crucial to capture the nature of the modes and ensure predictive capability for ITER plasmas. A combination of the real-time AMS amplitude and phase measurements can be used to detect the onset of the relevant modes, potentially in time for a disruption avoidance or mitigation system to be deployed.
Speaker: Dr Francesca Turco (Columbia University)
• 08:30
Controlling Marginally Detached Divertor Plasmas 4h
A new control system at DIII-D has stabilized the detached divertor plasma state in close proximity to the threshold for reattachment, thus demonstrating ability to maintain detachment with minimum gas puffing. When the same control system was instead ordered to hold the plasma at the threshold, the resulting T_e profiles separated into two groups with one group consistent with marginal detachment, and the other with marginal re-attachment. This shows that a physical bifurcation is taking place, and the plasma dithers between the attached and detached states when the control system attempts to hold to the threshold. The control system is upgraded from the one described by Kolemen, et al. [1] and it handles ELMing plasmas by using real time D_alpha measurements to remove during-ELM slices from real time T_e measurements derived from Thomson scattering. The difference between measured and requested inter-ELM T_e is passed to a PID controller to determine gas puff commands. While some degree of detachment is essential for the health of ITER’s divertor, more detached plasmas come at higher density with more radiation and excessive loss in confinement, making it desirable to limit detachment to the minimum level needed to protect the strike point [1]. However, the observed bifurcation in plasma conditions at the outer strike point with ion B × nabla B into the divertor makes this a significant challenge. The ideal solution without local impurity puffing lies within a narrow (3%) range in upstream density with a steep penalty for going out of bounds; if the divertor plasma were to reattach, there could be a long (depending on delays in the gas puff system) window of high heat flux before detachment could be re-established. Thus, good understanding of detachment behavior near the threshold for re-attachment is required to properly tune an active control system to maintain ideal divertor performance without reattaching. The top-of-pedestal electron densities during dithering across the bifurcation and during stable marginally detached operation are the same within uncertainty, showing the need for local real-time measurements of the divertor conditions. [1] E. Kolemen, et al., J. Nucl. Mater. 463, 1186 (2014), DOI:10.1016/j.jnucmat.2014.11.099
Speaker: Dr David Eldon (Princeton University)
• 08:30
Coupling full-f gyrokinetic studies to experimental measurements of the isotope effect for FT-2 tokamak plasmas 4h
Turbulent transport and flow dynamics in Ohmic FT-2 tokamak plasmas are investigated. Measurements utilize highly localized state-of-the art backscattering while the turbulence simulations are performed with the global full-f nonlinear code ELMFIRE. The role of the geodesic acoustic mode in regulating turbulent transport is studied. Special emphasis is given to the isotope effect observed in tokamak anomalous transport scaling.
Speaker: Susan Leerink (Aalto University)
• 08:30
Developing Disruption Warning Algorithms Using Large Databases on Alcator C-Mod and EAST Tokamaks 4h
To address the challenge of disruption prediction, we have created large disruption warning databases for both Alcator C-Mod and EAST by compiling values for a number of proposed disruption-relevant parameters sampled at many different times throughout all plasma discharges, disruptive and non-disruptive, during the 2015 campaigns on the respective machines. The disruption-relevant parameters include such intuitive quantities as Ip error [= Ip – Ip (programmed)], radiated power fraction [= Prad/Pinput], n/nGreenwald, n=1 mode amplitude, as well as a number of equilibrium parameters derived from EFIT reconstructions (q95, elongation, etcetera). Examples of the evolution of these parameters prior to disruptions on C-Mod and EAST, will be shown. The disruption warning databases for C-Mod and EAST each contain parameter values from well over 100,000 time slices. This allows one to provide quantitative answers to such questions as: (1) Is parameter “X” (e.g. Ip error or n/nG or n=1 mode amplitude) correlated with impending disruptions? If yes, (2) What fraction of disruptions do not show a correlation (i.e. missed disruptions)? (3) What is an appropriate trigger level for each correlated parameter, and how does the number of ‘false positives’ vary with the trigger level? (4) What is the typical warning time, and how does the warning time vary with trigger level? This fundamental quantitative characterisation of disruption-relevant parameters is absolutely crucial for developing any credible real-time disruption warning algorithms. These databases are also amenable to the application of advanced ‘machine learning’ techniques to discern more complicated dependencies on parameters, and the development of more advanced warning algorithms. In principle, the disruption-relevant parameters in the C-Mod and EAST disruption warning databases could be available in real-time, and their plasma control systems could implement a disruption prediction algorithm based on the analysis of these large databases to provide a warning with sufficient lead time that could be used to move the plasma to a less unstable state to avoid a disruption, or to trigger a disruption mitigation system. Acknowledgments: This work supported in part by: US DoE Grants DE-FC02-99ER54512, DE-SC0010720 and DE-SC0010492, using Alcator C-Mod, a DoE Office of Science User Facility
Speaker: Dr Robert Granetz (MIT)
• 08:30
Diamagnetic Plasma Confinement in Linear Traps 4h
A new efficient method of magnetic confinement is suggested for use in linear traps with extremely high plasma pressure. While pressure grows, the equilibrium in a linear trap changes in such a way that the effective mirror ratio increases, and, as a result, the axial particle and energy confinement becomes gas-dynamic and improves linearly with mirror ratio. This effect is due to diamagnetic expulsion of the magnetic field from the plasma volume while beta tends to one. The improved confinement could lead to construction of a compact fusion reactor based on a linear trap, if one could ensure suppression of pressure-driven instabilities, in particular, the ballooning instability. This paper shows how it can be done: one should use magnetic configuration with a stretch of uniform field at its minimum, and place external stabilizers near ends of that stretch. Equilibrium with anisotropic pressure, MHD stability, fast-ion confinement, startup, and the energy balance issues are considered for a linear trap in the diamagnetic-bubble regime. Such type of confinement is shown to be very promising for reactor perspectives.
Speaker: Dr Alexei Beklemishev (Budker Institute of Nuclear Physics SB RAS)
• 08:30
Dimensionless Size Scaling of Intrinsic Rotation 4h
A dimensionless empirical scaling for intrinsic toroidal rotation is given; MA~βNρ*, where MA is the toroidal velocity divided by the Alfvén velocity, βN the usual normalized β value, and ρ* is the ion gyroradius divided by the minor radius. This scaling is in agreement with experimental data from DIII-D, and also incorporates some published data from C-Mod and JET. The velocity used in this scaling is in an outer location in minor radius, outside of the interior core and inside of the large gradient edge region in H-mode conditions, although the scaling result is not very sensitive to the chosen location in H-mode. This scaling establishes the basic magnitude of the intrinsic toroidal rotation and we discuss its relation to the rich variety of rotation profiles that can be realized for intrinsic conditions, that is, minimal injected torque. This scaling has some similarities to existing dimensioned scalings, both the Rice scaling [J.E. Rice et al, Phys. Plasmas 7, 1825 (2000)] and the scaling of Parra et al [F.I. Parra et al, Phys. Rev. Lett. 108, 095001 (2012)].
Speaker: Dr John deGrassie (General Atomics)
• 08:30
Disruption Mitigation in the Presence of Pre-existing MHD Instabilities 4h
Experiments on the DIII-D and Alcator C-Mod tokamaks show that disruption mitigation by massive gas injection (MGI) and shattered pellet injection (SPI) of high-Z impurities remain effective in the presence of large pre-existing MHD instabilities. Rotating and locked magnetic islands will precede a large fraction of disruptions in ITER, making their impact on disruption mitigation a critical concern. Experiments on both machines show that such instabilities do not significantly impede the ability of massive impurity injection to mitigate thermal quench (TQ) and current quench (CQ) loads. On DIII-D, SPI significantly increases peak densities relative to unmitigated disruptions, indicating efficient assimilation of the injected impurities, with or without the presence of the modes. Similar results are found for MGI, on both DIII-D and C-Mod. This efficient assimilation of injected high-Z radiating impurities allows effective TQ mitigation, with enhanced radiation fractions and corresponding decreases in divertor heating, as measured by infrared imaging. MGI and SPI are able to accelerate the CQ even in the presence of pre-existing MHD precursors. Reconstructions of the plasma geometry during the CQ show that vertical displacements of the plasma are reduced relative to unmitigated disruptions, for all cases with MGI or SPI, while halo current impulses and resulting vacuum vessel displacements are also significantly reduced. Peak electron density during the CQ, which is an important metric for runaway electron mitigation, is enhanced in the case of impurity injection and do not differ between stable and MHD unstable discharges. Toroidally distributed measurements of the radiated power on C-Mod indicate that radiation asymmetries during the disruption are not significantly higher in plasmas with locked modes. This result implies that measured radiation asymmetries are likely driven by MHD activity which is initiated by the impurity injection process itself, rather than by the pre-disruption instabilities. Overall, these results on DIII-D and C-Mod increase confidence in the existing physics basis for disruption mitigation, and the resulting design of the ITER disruption mitigation system. This work was supported by the US Department of Energy under DE-AC05-00OR22725, DE-FC02-99ER54512, DE-FC02-04ER54698, DE-FG01-07ER54917, DE-AC52-07NA27344 and DE-SC0006757.
Speaker: Mr Daisuke Shirakid (Oak Ridge National Laboratory)
• 08:30
Divertor and Core Plasma Performance Optimization Enabled by Direct Feedback Control of Surface Heat Flux on Alcator C-Mod’s High-Z Vertical Target Plate Divertor 4h
The C-Mod team has developed a new tool for control of plasma conditions at the divertor target. It is the first heat flux mitigation system to employ real-time measurements of surface heat flux to control impurity seeding. Control of the conditions at the divertor surface is one of the remaining challenges to tokamak fusion energy. Active cooling technology limits the surface heat flux to ~10 MW/m^2 or less and erosion of solid, high-Z targets limits the incident plasma temperature <5 eV. Yet, plasma entering the divertor will be intense: The scrape-off layer heat flux width scales inversely with the poloidal magnetic field and is independent of machine size [1]. This results in a parallel heat flux that scales as q||~PSOL*B/R. Projecting this to ITER results in an unmitigated parallel heat flux of ~5 GW/m^2 and >10 GW/m^2 in a DEMO-class device [2]. Seeding of low-Z impurities (N2 and Ne) into the divertor to mitigate the focused plasma heat flux into a uniform photon heat flux is a viewed as a necessity. Yet excessive seeding comes at a cost, lowering pedestal pressures and increasing core dilution. C-Mod is an excellent test of reactor-relevant plasma control solutions: it has an ITER-like high-Z, vertical target plate and is the only experiment with demonstrated heat fluxes in excess of 1 GW/m^2. During FY15 a system for real-time control of heat flux was deployed [3]. The system uses surface thermocouples and an analog computer to output accurate signals of heat flux, which are used to control the injection rate of nitrogen into the private flux region. It has been used to reduce the surface heat flux from >25 MW/m^2 (corresponding to an unmitigated parallel heat flux q||~500 MW/m^2) to less than 5 MW/m^2 while avoiding degradation of energy confinement, i.e. H98. Yet, even at nearly zero surface heat flux, the divertor Langmuir probes still indicate a plasma temperature too high (>5 eV) to suppress sputtering. In the FY16 campaign a divertor mirror Langmuir probe system [4], which outputs real-time measurements of electron temperature, will be used in an attempt to feedback on the plasma temperature at the divertor target. [1] T. Eich, et al., Nucl. Fusion 53 (2013) 093031; [2] B. LaBombard et al., Nucl. Fusion 55 (2015) 053020; [3] D. Brunner, et al., Rev. Sci. Inst. 87 (2016) 023504; [4] B. LaBombard and L. Lyons, Rev. Sci. Inst. 78 (2007) 073501.
Speaker: Dr Dan Brunner (MIT PSFC)
• 08:30
Dominant role of turbulence in determining particle transport and confinement 4h
In this paper we will show that particle confinement is determined by changes in turbulence characteristic outside mid-radius up to the top of the pedestal in DIII-D H-mode plasmas. We find that the Electron Cyclotron Heating (ECH) density pump-out at low collisionality is the result of an increase in turbulence drive from ρ=0.7-0.9. The frequency of the mode and thus the turbulence type changes from the Ion Gradient Temperature (ITG) to Trapped Electron Mode (TEM) on a much longer time scale. Secondly, we find that particle confinement is strongly reduced in balanced torque injected plasmas, where the ExB shear close to the top of the pedestal drops below the linear growth rate of long wavelength turbulence [1,2]. Both these observations have a strong impact on predictions for ITER, where most of the plasma heating is deposited in the electron channel and the Neutral Beams (NBI) inject relatively low momentum. One option to counter effects of reduced particle confinement at the plasma edge is to increase peaking of the core density profile. An experimental database has shown that density peaking is influenced by collisionality, and the dominant unstable mode, while theory predicts the q-profile should be inversely proportional with the density gradient [3]. We do find, similar as on AUG, that at mid-radius the frequency of the most unstable mode correlates with the inverse density scale length. Experimentally we observe that 1/q ~ grad ne, as predicted by theory. The correlation is stronger when Te=Ti and the plasma is in the ITG regime. When we increase the electron temperature with ECH, the correlation becomes weaker. From these results we can conclude that turbulence plays a dominant role in determining the density profile and particle confinement. However, the predictive capability of particle transport is not well validated and the density profile is assumed to be flat in ITER scenarios [4]. Through comparisons with quasi-linear gyro-kinetic simulations we plan to validate existing codes in order to make better predictions for ITER. [1] X. Wang et al. PPCF 2015 In review [2] S. Mordijck et al. PoP 19 (2012) 056503 [3] C. Angioni et al. Nucl. Fus. 52 (2012) 114003 [4] L. Garzotti et al. Nucl. Fus. 52 (2012) 013002
Speaker: Prof. Saskia Mordijck (The College of William and Mary)
• 08:30
Edge Flow from Momentum Transport by Neutrals 4h
The accessibility and performance of the H-mode are critical to tokamak fusion reactors. While the physics of the pedestal is complicated and far from fully understood, it is clear that flow shear plays an important role. One of the mechanisms that may regulate flow velocity in the plasma edge is momentum transport by neutrals. Due to their high cross-field mobility they may be the most significant momentum transport channel even at low relative densities. There have also been experimental observations suggesting an important role for neutrals in determining pedestal properties and showing strong variation of confinement with the strike point position (as the target configuration changes). We have developed novel numerical tools to determine the radial electric field and plasma flows just inside the separatrix by coupling a kinetic, short mean-free-path model of neutrals to a neoclassical solver. This allows us to compute the momentum flux due to the neutrals that corresponds to specified plasma profiles. The flux must vanish in steady state and this constraint then determines the radial electric field given the density and temperature gradients. As an example we demonstrate the effect of X-point position on edge flow in ITER-like geometry. The major radius of the X-point has a strong effect both on the magnitude of the flow and its collisionality dependence, suggesting that altering the X-point position may offer a means to manipulate the edge flow shear. Beyond this particular scenario, it is clear that in the edge region of the plasma where neutral densities are relatively high the neutrals can have important effects on the radial momentum transport, flow and flow shear. Consequently they are likely to affect both the L-H transition and H-mode confinement, and they should be accounted for in the interpretation of current experiments and in the design of future machines.
Speaker: Dr John Omotani (Department of Physics, Chalmers University of Technology)
• 08:30
Edge- and divertor and plasma behavior in high power high performance double-null plasmas 4h
We identify major challenges to reducing divertor heat flux in high power, high performance near-double null DIII-D plasmas, while still maintaining sufficiently low density to allow for application of RF heating. The plasmas discussed here are characterized by: β_N ≅ 3–4, H_98 ≅ 1.5–1.7, dR_SEP ≅ -5 mm, and P_IN up to 20 MW. The scaling of the peak heat flux (q⊥p) at the outer target of the primary divertor was proportional to P_SOL^0.92 and Ip^0.92 in the range P_SOL = 8 -19 MW and Ip = 1.0–1.4 MA and is consistent with standard ITPA scaling for single null configurations. Three distinct divertor heat flux reduction techniques were tested. First, the puff-and-pump radiating divertor was less effective in reducing divertor heat flux when β_N was raised to 3.7 than occurred for lower values of β_N and P_IN. In the higher β_N case, gas puffing during puff-and-pump resulted in an increase in τ_E and τ_p and led to more rapid fueling of the core. This set an upper limit on the D_2 injection rate that can be tolerated without losing density control, thereby undermining the effectiveness of puff-and-pump. We are investigating how a decrease in ELMing frequency during D_2 injection at higher power and β_N may drive this process. Second, increasing the poloidal flux expansion at the outer target of the primary divertor did not produce the expected reduction in q⊥p that would have been expected from geometrical arguments, e.g., almost doubling the poloidal flux expansion reduced q⊥p by only ~20%. Preliminary analysis suggests that cross-field diffusion effects appear to counteract poloidal flux expansion. Third, we show how q⊥p was reduced by 25-50% when an open divertor is closed on the common flux side of the outer divertor target (“semi-slot” divertor). Steady carbon buildup in the main plasma became significant during higher P_IN operation, and was largely due to sputtered carbon from graphite tiles on the horizontal surface above the pumping plenum entrance (and not from the “divertor floor). Our results strongly suggest the necessity of further study before relying on either radiating divertor or poloidal flux expansion to adequately control divertor heat flux in high power, high performance DN plasmas. *Work supported by the U.S. Department of Energy under DE-FC02-04ER54698, DEAC52007NA27344, DE-AC02-09CH11466, DE-FG02-04ER54761, and DE-AC04- 94AL85000.
Speaker: Dr Thomas W. Petrie (General Atomics)
• 08:30
Effect of energy-non-transporting nonlinear flux on the turbulent plasma transport 4h
Turbulent plasma transport is an important issue in the confinement of the fusion plasmas. E×B nonlinear flux plays a deciding role in redistributing the energy and the momentum. In the case of the isothermal plasma the particle flux Γ is the fundamental measure of the turbulent transport. In the fusion plasma where the particle collisions are relatively rare the plasma is near the adiabatic state. The present work sets out to study Γ near adiabatic state of the electrostatic resistive fluid plasma fluctuations [1]. Simulations are performed in the BOUT++ frame [2] in two-dimensional slab Lx=Ly=80π perpendicular to the equilibrium magnetic field with grids 256×256. The fixed equilibrium density gradient is in the negative x direction and the electron diamagnetic drift is along the y axis. Fluctuations at small scales kρ_s≫1 are cut off by the hyper dissipation. At a quasi-steady state of the energy Γ, in contrast to the energy, is not quite steady exhibiting intermittent peaks as it changes between less than 10 to larger than 40. It is found that secondary broad peak 0.6≤kρ_s≤1.0 is present when Γ is large. The spectra of the E×B energy flux illustrate a similar distribution in the range kρ_s≥0.5 that generally agrees with the difference of Γ's. The E×B energy flux is divided into two parts: one directly involved in the energy redistribution and the other [3]. Preliminary results indicate the strong impact of the latter part on Γ by controlling the phases of the fluctuations. Similar process is believed to be working inside the fusion plasma. Detailed analyses of the correlation between the energy flux and the phases as well as extension to the thermal flux of more realistic plasma models will be discussed at the conference. [1] A. Hasegawa and M. Wakatani, Phys. Rev. Lett. 50, 682 (1985). [2] B. D. Dudson et al, Comput. Phys. Commun. 180, 1467 (2009). [3] B. Min et al, Plasma Phys. Control. Fusion 57, 095009 (2015); B. Min et al, J. Kor. Phys. Soc. 66, 1226 (2015).
Speaker: Mr Chan-Yong An (Soongsil University)
• 08:30
EFFECT OF MAGNETIC SHEAR AND EQUILIBRIUM FLOWS ON COLLISIONLESS MICROTEARING AND MIXED PARITY MODES IN HOT TOKAMAK PLASMAS 4h
Turbulent transport of energy, particles and momentum is one of the important limiting factors for long time plasma confinement. Modern kinetic study using gyrokinetic formalism and simulation has progressed to identify several microinstabilities that cause ion and electron thermal transport. Typically, these have been ballooning parity modes such as the ITG, KBM and ETG modes which cause transport through fluctuations or tearing parity modes such as Microtearing modes (MTM) which change the local magnetic topology and cause transport through stochastization of the magnetic field. Local gyrokinetic simulations have found collisional MTMs unstable in several magnetic confinement configurations such as Spherical Tokamaks, Reverse Field Pinch and Standard Tokamaks. Aditya K Swamy et al. [Phys. Plasmas 21 (2014), 22 (2015)] have found global Collisionless MTMs to be linearly unstable in regions high positive magnetic shear. The collisionless MTM is found to be driven unstable by the magnetic drift resonance of passing electrons. In this work, we address the complex multiscale problem of MTM stability in advanced tokamak scenarios which envisage reversed magnetic shear with observed strong sheared poloidal and toroidal flows in the Internal Transport Barrier. In the first part of this work, safety factor profiles are continuously varied parametrically from standard shear profiles to weak and reverse shear profiles. Multiple MTM modes are found at finite positive shear. As the global safety factor profile is varied, novel mixed parity modes of MTMs are found to become unstable with weak shear. In the second part, the effect of equilibrium flows are studied for their effect on MTM and mixed parity (MP) instabilities and their global mode structures. These and several other characteristics of MTMs and Mixed Parity modes will be reported.
Speaker: Dr Rajaraman Ganesh (Institute for Plasma Researh, Bhat Village, Gandhinagar 382428, Gujarat, INDIA)
• 08:30
Effect of the EC torque on slow plasma rotation under central ECH/ECCD for NTM onset 4h
The modification of low toroidal plasma rotation under application of central EC power injection with possible onset of neoclassical tearing modes (NTM) is an important issue for plasma confinement and for future devices (ITER will be characterized by a low rotation). In low collisionality regime and PEC/Poh >1 TCV experiments central co-ECCD was observed to modify the toroidal plasma rotation both in absence and at the appearance of 3/2 and 2/1 modes, while no modes with cnt-ECCD were observed. The rotation profiles were promptly modified by the central EC power deposition and driven towards the plasma current direction as also observed in similar recent TCV experiments in the framework of the EUROfusion MST1work package. Dedicated MST1 discharges for this study were also done on ASDEX Upgrade and analysis is still ongoing. The understanding of the physical mechanisms acting on the modification of the toroidal plasma rotation can allow the avoidance of the NTM onset and of the loss of confinement. The torques associated with the rotation changes have been generally not associated to a direct action of the EC power absorption, because the EC heating and current drive do not transfer momentum. The toroidal rotation evolution under the effect of EC injection was simulated using a simplified model including an effective momentum diffusivity, scaled on the anomalous ion heat diffusivity, and an effective EC source term describing the effect of the torque due to possible different mechanisms. The physical origin of the torque associated with the EC power absorption in low collisionality regime for PEC/Poh >1, can be associated to different causes such as driven turbulent effects or mechanisms of particles pump-out or based on the asymmetry in the power deposition. In this work the change of rotation under the central EC power injection is investigated using the momentum balance equation and considering these different mechanisms for the interpretation of the experimental results. We consider torque associated to the turbulent effects for the absorption of the EC power related to the enhancement of the turbulent Reynolds stress, torque related to the recycling phenomena at the edge and the occurrence of the EC pump-out and non-vanishing torque driving the toroidal rotation and provided by the surface-averaged displacement current .
Speaker: Dr Silvana Nowak (IFP-CNR)
• 08:30
Effects of Heat and Particle Sources Perturbations on L-H-L Transitions Based on Bifurcation Concept 4h
This work aims to investigate the effects of perturbations of heat and particle sources on the formations of edge transport barrier (ETB) and on the hysteresis properties at the L-H-L transitions in the framework of bifurcation concept. The formation of transport barriers is studied via the combination of thermal and particle transport equations, which also includes neoclassical and anomalous effects. The suppression mechanism based on flow shear stabilization is assumed to affect only on the anomalous channel, where the flow shear is estimated from the force balance equation and couples both transport equations. The main thermal and particle sources are localized near plasma center and edge, respectively. Experimental evidences and theoretical understanding reveal that the formation of an ETB, leading to an L-H transition, is related to the critical heating threshold. Analytical study reveals that the fluxes versus gradients space exhibits bifurcation behaviour with s-curve soft bifurcation type. Evidently, the backward H-L transition occurs at lower values than that of the forward transition, illustrating hysteresis behavior. This work investigates perturbations effects of thermal and particle sources on the formations of both ETB and hysteresis properties. The focus is on the possibility of L-H transition triggering by the fluctuations in heating at marginal point and by pellet injection. It was shown that H-mode can be triggered and maintained so the central plasma pressure can be increased.
Speaker: Dr Thawatchai Onjun (Sirindhorn International Institute of Technology)
• 08:30
Effects of Localized Neoclassical Toroidal Viscosity Effects on the Toroidal Rotation Profile in KSTAR 4h
KSTAR provides a great environment to carry out the NTV study in that the intrinsic error fields and the toroidal field ripples are very small in magnitude, and asymmetric magnetic fields can be added by the in-vessel coil current on demand. In this paper, we report both theoretical and experimental studies on NTV in KSTAR. It is shown that the radial transport of the toroidal angular momentum, , is also proportional to the first order of gyro-radius. In this work, we introduce a different method of the NTV torque estimation, that includes the usual toroidal angular momentum transport besides the NTV torque. It may resolve some known discrepancies between theories and experiments and reveal unknown puzzles at the same time. We show that the inherent neoclassical toroidal viscosity induced by the intrinsic error fields and toroidal field ripple in KSTAR is small enough not to deform the pedestal structure in toroidal rotation profiles, always observed uniquely in H-mode KSTAR plasmas.
Speaker: Dr JaeChun Seol (National Fusion Research Institute)
• 08:30
Effects of the q Profile on Toroidal Rotation in Alcator C-Mod LHCD Plasmas 4h
In future magnetic fusion devices, external momentum input from neutral beam injection will be low, and to reap the benefits of rotation, such as stabilization of deleterious MHD modes and shear suppression of turbulence, utilizing radio frequency drive and understanding self-generated flow would be prudent. Changes in the core toroidal rotation profiles following injection of lower hybrid (LH) waves have been documented in Alcator C-Mod plasmas. Shot by shot scans of LH input power have been performed at fixed magnetic field and electron density for several plasma currents. For sawtoothing target plasmas, if the input power is low enough that the central safety factor q0 remains below 1, the change in the core rotation is in the counter-current direction, consistent in sign, magnitude and LH power scaling with direct momentum input from the LH waves. If the power level is high enough that there are significant changes to the q profile, including the termination of sawtooth oscillations, the change in the toroidal rotation is in the co-current direction, consistent with changes in sign of the momentum flux through the residual stress and its dependence on the current density profile. The direction of the rotation changes depends on the whether q0 is below or above unity, and seemingly not on the magnetic shear, nor the Ohmic confinement regime of the target plasma.
Speaker: Dr John Rice (MIT PSFC)
• 08:30
Electron Cyclotron Heating Modification of Alfvén Eigenmode Activity in DIII-D 4h
Localized electron cyclotron heating (ECH) can have a dramatic impact on neutral beam driven Alfvén eigenmode (AE) activity in DIII-D plasmas. The most common effect, which is explained here for the first time, is a shift in the dominant observed modes from a mix of reversed shear Alfvén eigenmodes (RSAEs) and toroidicity induced Alfvén eigenmodes (TAEs) to a spectrum of weaker TAEs when ECH is deposited near the shear reversal point, q_min. Discharges with weaker RSAE activity also have reduced fast ion transport. A recent experiment to understand the physical mechanisms responsible for this shift in AE stability included variations of ECH steering, power, and timing as well as current ramp rate, beam injection geometry, and beam power. All variations were observed to change the impact of ECH on AE activity significantly. In some cases, RSAE activity was enhanced with ECH near q_min as opposed to near the axis, in contrast to the original DIII-D experiments [1]. It is found that during intervals when the geodesic acoustic mode (GAM) frequency at q_min is elevated and the calculated RSAE minimum frequency is very near or above the nominal TAE frequency (f_TAE), RSAE activity is not observed or RSAEs with a much reduced frequency sweep range are found. This condition is primarily brought about by ECH modification of the local electron temperature (T_e) which can raise both the local T_e at q_min as well as its gradient. A q-evolution model that incorporates this reduction in RSAE frequency sweep range is in agreement with the observed spectra and appears to capture the relative balance of TAE or RSAE-like modes throughout the current ramp phase of over 38 DIII D discharges. Detailed ideal MHD calculations using the NOVA code show both modification of plasma pressure and pressure gradient at q_min play an important role in modifying the RSAE activity. Analysis of a case with ECH near q_min, and no observable RSAE activity, shows the traditional RSAE is no longer an eigenmode of the system. Calculations with the non-perturbative gyro fluid code TAEFL confirms this change in RSAE activity and also shows a large drop in the resultant mode growth rates. Work supported by the U.S. DOE under DE-FC02-04ER54698, SC-G903402, DE-AC05-00OR22725, and DE-AC02-09CH11466. [1] M.A. Van Zeeland, et.al PPCF 50 (2008) 035009
Speaker: Dr Michael Van Zeeland (General Atomics)
• 08:30
ExB Shear and Precession Shear Induced Turbulence Suppression 4h
Starting from the modern bounce-kinetic formalism, [1] a two-point equation which properly describes turbulent eddies associated with trapped electrons is systematically derived in general tokamak geometry. Trapped electron precession shear, as well as ExB shear, is naturally included in the derivation. Our two-point analysis, using moments of separation between the two points, reveals that both precession shear and ExB shear participate on suppressing trapped-electron-related turbulence and their synergism is determined by the relative sign. Our result provides explanations on broad range of experimental observation regarding electron thermal internal transport barrier observed in various tokamaks. [2-4] References [1] B.H. Fong and T.S. hahm, Phys. Plasmas 6, 188 (1999) [2] F.M. Levinton et al., Phys. Rev. Lett. 75, 4417 (1995) [3] G.D. Conway et al., Plasma Phys. Control. Fusion 44, 1167 (2002) [4] T. Fujita et al., Plasma Phys. Control. Fusion 46, A35 (2004)
Speaker: Prof. Taik Soo Hahm (Seoul National University)
• 08:30
Experimental results from three-ion species heating scenario on Alcator C-Mod 4h
Recent experiments on Alcator C-Mod using a small fraction of 3He added to a H(D) plasma have demonstrated efficient ion cyclotron radio frequency (ICRF) heating and indications of MeV 3He tail temperatures. For high toroidal magnetic field B0=8 T discharges with D majority, 3He minority absorption is typically used and has low single pass absorption compared to the H minority absorption scenario. We have observed strong toroidal rotation that is correlated with RF power absorption on thermal 3He ions via mode converted waves. ICRF can also generate high energy ions and this provides a tool to study fast ion dynamics and optimize the quality of plasma confinement. This new scenario has much higher absorption and works by adjusting concentrations of the majority and two minority species to arrange that the polarization of the ICRF wave is favorable for ion heating at the location of the cyclotron resonance of a third trace species. Experiments using a H:D:(3He) three-ion scenario were carried out on C-Mod with an 8T field and an H:D ratio of approximately 2:1. The 3He fraction was varied from 0.4% to 2%. A strong increase in toroidal Alfvén eigenmode (TAE) activity coincided with the addition of 3He to the H(D) plasmas. TAE activity is indicative of the formation of fast ions with a energy on the order of 1 MeV. Increased heating localized around the 3He fundamental cyclotron layer was also observed. We will present analysis of the minority ion temperature using Fokker-Planck calculations coupled with a full wave code over a range of 3He fractions. These temperatures will be compared with theoretical calculations of TAE thresholds. A synthetic PCI diagnostic using the modeled 3D RF fields will be compared to the experimental PCI to determine the breakdown between the two competing absorption mechanisms present in mode conversion layer. These mechanisms can either heat electrons or ions or drive momentum in the ion channel. We will conclude with a discussion of the applicability of this scenario to the upcoming D-T campaign on JET, operations on ITER and as a source of pseudo-alphas on W7-X. * Supported by USDOE grants for SciDAC Center for Simulation of Wave Plasma Interactions DOE DE-FC02-01ER54648, Alcator C-Mod Science user facility DE-FC02-99ER54512, and the Phase Contrast Imaging Diagnostic on C-Mod, DE-FG02-94-ER54235.
Speaker: Dr John Wright (MIT - PSFC)
• 08:30
Experiments on Helicons in DIII-D – Investigation of the Physics of a Reactor-relevant Non-Inductive Current Drive Technology 4h
Experiments have begun in DIII-D to evaluate non-inductive current drive by the Landau absorption of a toroidally-directive spectrum of helicon waves (also known as 'very high harmonic fast waves', 'fast waves in the lower hybrid range of frequencies', or 'whistlers'). Modeling has shown [1] that non-inductive current drive at mid-radius (ρ~0.5) should be achievable in DIII-D with fast waves at 0.5 GHz, with an efficiency twice as high as with non-inductive current drive tools currently available on DIII-D (neutral beams and electron cyclotron current drive) in high-beta conditions. An innovative Traveling Wave Antenna (TWA) of the 'comb-line' type with 12 radiating modules has been constructed, installed in DIII-D, and is currently being tested at very low power (<1 kW) to evaluate the antenna coupling in the linear regime, and to prototype technological aspects of such structures in the tokamak environment. Preliminary results indicate strong antenna/plasma coupling, with detailed 3D modeling underway to quantitatively compare the measurements with theoretical expectations. A key input to this model is the edge and far SOL electron density profile, which is being measured with a microwave reflectometer and with fixed and moveable Langmuir probes. An important issue for wave coupling in this regime is the degree to which (undesired) quasi-electrostatic slow waves are excited by the structure; evaluation of this is a point of emphasis in the ongoing work. A high-power system is presently being prepared for installation later in 2016 in which a single 1.2 MW klystron at 476 MHz will be used to power a TWA with ~36 radiating elements in a structure 2 m wide. The goals of the high-power experiments include evaluation of non-linear effects on excitation of the desired waves (ponderomotive effects, parametric decay) and measurements of the deposition profile and of the current drive efficiency. Ray-tracing predicts [1] an rf-driven current of ~60 kA per coupled MW of helicon power, which should result in an easily measurable driven current in DIII-D in high-beta discharges. This work was supported in part by the US Department of Energy under DE-FC02-04ER54698, DE-FG02-94ER54084, DE-FG02-07ER54917, DE-AC05-00OR22725, DE-AC02-09CH11466, DE-AC04-94AL85000, DE-FG02-08ER54984. [1] Prater, R. et al 2014 Nuclear Fusion 54 083024.
Speaker: Dr Robert Pinsker (General Atomics)
• 08:30
Ferritic Wall and Scrape-Off-Layer Current Effects on Kink Mode Dynamics 4h
The HBT-EP research program aims to: (i) quantify wall-stabilized kink mode dynamics and multimode response to applied magnetic perturbations, (ii) understand the relationship between control coil configuration, conducting and ferromagnetic wall effects, and active feedback control, and (iii) explore advanced feedback algorithms. We present an overview of planned research activities for the next 3 years, along with ongoing experiments in support of upcoming research. Multiple simultaneous kink modes are measured by over 200 local magnetic sensors. We observe that modulation of n=1 halo current depends on phase and amplitude of rotating kink modes and applied RMPs. A low latency (14us) control system uses 96 inputs and 64 outputs for adaptive control of kinks, via a 512-core graphics processing unit. An in-vessel adjustable ferromagnetic wall is used to study ferritic resistive wall modes, and produces increased mode growth rates, RMP response, and disruptivity. Magnetic feedback suppression of modes requires higher actuator gain with the nearby ferritic wall. A biased electrode in the plasma is coupled to the feedback system, and is used to control the rotation of kinks and evaluate error fields. At strong positive bias, the electrode induces a fast plasma rotation state with enhanced poloidal flow shear. A quasi-linear sharp-boundary model of the plasma's multimode response to error fields, including resistive and ferromagnetic effects, is developed to determine harmful error-field structures. Machine upgrades will allow improved measurements and control of scrape-off-layer (SOL) currents. Movable tiles positioned around limiting surfaces will measure SOL and vessel currents during mode activity and disruptions. Biasable plates at divertor strike points will allow control of field-aligned SOL currents for feedback studies. An extreme ultraviolet diagnostic including (i) four sets of 16 poloidal views and (ii) a two-color 16-chord tangential system will allow tomographic reconstruction of the plasma's optical emission and internal structure of kink modes, along with temperature profiles versus time. These measurements will enable feedback on kink modes using only optical sensors and both magnetic and edge-current actuators.
Speaker: Dr Jeffrey Levesque (Columbia University)
• 08:30
Full-f gyrokinetic simulation including kinetic electrons 4h
A new hybrid kinetic trapped electron model [Y. Idomura, J. Comput. Phys. 313, 511 (2016)] is developed for electrostatic full-f gyrokinetic simulations. The model is verified by computing the ion and electron neoclassical transport and the linear ion temperature gradient driven trapped electron mode (ITG-TEM) stability, in which collisional TEM stabilization shows an isotope effect. An impact of kinetic electrons on the ITG turbulence is investigated by comparing ITG turbulence simulations with adiabatic electrons and with kinetic electrons. It is found that in the kinetic electron case, resonant passing electrons transport at mode rational surfaces generates corrugated density profiles, and the resulting microscopic radial electric field Er sustains nonlinear critical temperature gradients above linear ones. This is qualitatively different from the so-called Dimits shift sustained by turbulence driven zonal flows in the adiabatic electron case. In the toroidal angular momentum balance, kinetic trapped electrons enhance the field term stress, which is characterized by the phase difference between the perturbed distribution and the toroidal electric field, and thus reversed between the adiabatic and kinetic electron cases. It is also found that the field term stress and the resulting intrinsic rotation is reversed between the ITG turbulence and the TEM turbulence.
Speaker: Dr Yasuhiro Idomura (Japan Atomic Energy Agency)
• 08:30
Global kinetic effect on the collisionality dependence of the neoclassical toroidal viscosity in the superbanana-plateau regime 4h
The neoclassical toroidal viscosity (NTV) caused by a non-axisymmetric magnetic field perturbation is one of the key issues for the prediction and control of the plasma performance and/or stabilities, since it can play an important role in a momentum balance which determines a plasma rotation. However, there remains a severe discrepancy with regard to the NTV prediction; the so-called superbanana-plateau theory based on a simplified bounce-averaged model predicts collisionality-independent, or the resonant NTV [K. C. Shaing et al., Plasma Phys. Control Fusion 51, 035009 (2009)], while a global drift-kinetic simulation by FORTEC-3D [S. Satake et al., Phys. Rev. Lett. 107, 055001 (2011)] shows a collisionality dependency of the NTV. In this study, we investigate the cause of the discrepancy using two different types of global kinetic simulations; one is FORTEC-3D which is based on drift kinetic equation and solves it using the delta-f Monte Carlo approach, and the other is GT5D which solves the gyrokinetic equation based on the Eulerian full-f approach. We demonstrate that the two global kinetic simulations reproduce similar collisionality dependencies of the NTV over wide ranges of the collisionality, indicating that the collisionality dependency of the NTV is common in the global kinetic simulations. It is found that a theoretically predicted resonant structure in the velocity space, which generates the collisionality-independent NTV in the superbanana-plateau theory, vanishes in the global kinetic simulations. The following two mechanisms are discussed as possible causes for the loss of the resonant structure, which may lead to the non-resonant and collisionality dependent NTV: 1) the magnetic shear dependency of the toroidal precession drift frequency, and 2) trapping/detrapping processes of perturbed particle orbits.
Speaker: Dr Seikichi Matsuoka (Japan Atomic Energy Agency)
• 08:30
Gyrokinetic simulations of electrostatic microinstabilities with bounce-averaged kinetic electrons for shaped tokamak plasmas 4h
Nonlinear bounce-averaged kinetic theory[B.H. Fong and T.S. Hahm, Phys. Plasmas 6, 188 (1999)] is used for magnetically trapped electrons for the purpose of achieving efficient gyrokinetic simulations of Trapped Electron Mode (TEM) and Ion Temperature Gradient mode with trapped electrons (ITG-TEM) in shaped tokamak plasmas. Bounce-averaged kinetic equations are explicitly extended to shaped plasma equilibria from the previous ones for concentric circular plasmas, and implemented to a global nonlinear gyrokinetic code, Gyro-Kinetic Plasma Simulation Program (gKPSP)[J.M. Kwon et al., Phys. Plasmas 21, 013004 (2012)]. Verification of gKPSP with bounce-averaged kinetic trapped electrons in shaped plasmas is successfully carried out for linear properties of ITG-TEM mode and Rosenbluth-Hinton residual zonal flow[M.N. Rosenbluth and F.L. Hinton, Phys. Rev. Lett. 80, 724 (1998)]. Physics responsible for stabilizing effects of elongation on both ITG mode and TEM is identified using global gKPSP simulations. These can be understood in terms of magnetic flux expansion leading to the effective temperature gradient R/L_T(1-E’)[P. Angelino, et al., PRL 102, 195002 (2009)] and poloidal wave length contraction at low field side resulting in the effective poloidal wave number k_theta*rho_i/kappa.
Speaker: Dr Lei Qi (National Fusion Research Institute, Daejeon, Korea)
• 08:30
Helical electric potential modulation via Zonal Flow coupling to Resonant Magnetic Perturbations 4h
Controlling Edge Localized Modes (ELMs) is very important for ITER, and a well-tested way to achieve this is by using external coils to generate Resonant Magnetic Perturbations (RMPs), demonstrated on several tokamaks [1-4]. The working hypothesis for the origin of ELM suppression is that RMPs increase transport in the pedestal, thus lowering the pressure-gradient below the ideal-MHD threshold. In this work, we show that - in presence of RMPS - Zonal Flows can drive a long-lived Vortex-Flow pattern. This finding clarifies the theory of RMP-induced Zonal Flow damping [5]. Note that evidence of such a Vortex-Flow pattern has been observed experimentally [6]. We obtain a dynamical system of coupled 1D equations for Zonal Flows and Vortex-Flow profiles, which we solve numerically (in our model, turbulence acts as a shear-dependent negative eddy-viscosity). As Zonal Flows are turbulence-driven, this shows that turbulence plays a major role in the plasma self-organization towards a 3D quasi-equilibrium. Contrary to Zonal Flows - which act as a benign reservoir of energy - the Vortex Flow pattern has a radial streamer-like flow associated to it and hence can drive convective transport. The associated enhancement in the particle transport - assuming the Vortex Flow has a density component - has a resonant character. This additional transport could act to limit the pressure-gradient, and is therefore a possible candidate to explain ELM suppression. [1] T.E. Evans et al, Nucl. Fusion 48, 024002 (2008). [2] Y. Liang et al, Phys. Rev. Lett. 98, 265004 (2007). [3] W. Suttrop et al, Phys. Rev. Lett. 106, 225004 (2011). [4] Y.M. Jeon et al, Phys. Rev. Lett. 109, 035004 (2012). [5] M. Leconte, P.H. Diamond and Y. Xu, Nucl. Fusion 54, 013004 (2014). [6] N. Vianello et al, Plasma Phys. Control. Fusion 57, 014027 (2015).
Speaker: Dr Michael Leconte (National Fusion Research Institute)
• 08:30
Improved Reproducibility of Plasma Discharges via Physics-model-based q-profile Feedback Control in DIII-D 4h
Recent experiments on DIII-D demonstrate the potential of physics-model-based q-profile control to improve reproducibility of plasma discharges. A combined feedforward+feedback control scheme was employed to optimize the current ramp-up phase by consistently achieving target q profiles (Target 1: qmin=1.3, q95=4.4; Target 2: qmin=1.65, q95=5.0; Target 3: qmin=2.1, q95=6.2) at prescribed times during the plasma formation phase (Target 1: t=1.5s; Target 2: t=1.3s; Target 3: t=1.0s). At the core of the control scheme is a nonlinear, first-principles-driven, physics-based, control-oriented model of the plasma dynamics valid for low confinement (L-mode) scenarios. A partial-differential-equation model of the q-profile dynamics combines the poloidal magnetic flux diffusion equation with physics-based models of the electron density and temperature profiles, the plasma resistivity, and the noninductive current sources (auxiliary and bootstrap). Firstly, a nonlinear, constrained optimization algorithm to design feedforward actuator trajectories is developed with the objective of numerically complementing the traditional trial-and-error experimental effort of advanced scenario planning. The goal of the optimization algorithm is to design actuator trajectories that steer the plasma to the target q profile at a predefined time subject to the plasma dynamics and practical plasma state and actuator constraints, such as the minimum q value and the maximum available auxiliary heating and current-drive (H&CD) power. To prevent undesired L-H transitions, a constraint on the maximum allowable total auxiliary power is imposed in addition to the maximum powers for the individual H&CD actuators. Secondly, integrated feedback control algorithms are designed to keep the q-profile evolution on track by countering the effects of external plasma disturbances, thereby adding robustness to the control scheme. The H&CD system and the total plasma current are the actuators utilized by the feedback controllers to control the plasma dynamics. Experimental results are presented to demonstrate the effectiveness of the combined feedforward+feedback control scheme to consistently achieve the desired target profiles at the predefined times. These results also show how the addition of feedback control significantly improves upon the feedforward-only solution by reducing the matching error.
Speaker: Prof. Eugenio Schuster (Lehigh University)
• 08:30
Investigations of radial high-Z transport mechanisms in ICRF-heated Alcator C-Mod H-mode plasmas 4h
Recent Alcator C-Mod research investigates mechanisms by which ion cyclotron range of frequency (ICRF) heating can effectively mitigate on-axis accumulation of high-Z impurities and explores new techniques to study their interaction with edge transport barriers (ETB). In C-Mod EDA H-modes using D(H) minority heating, modifying the minority concentration and the major radius of the minority resonance layer results in substantive changes in core high-Z impurity transport. Raising the minority fraction is linked to enhanced core peaking of tungsten injected via laser ablation. When the minority resonance layer is moved off-axis to the low-field side (LFS), bridging the q=1 surface, core accumulation is avoided similar to when heating on-axis. In contrast, off-axis heating on the high field side (HFS) at similar minor radii resulted in tungsten accumulation, uncontrolled radiation rise and core electron temperature collapse. These observations differ from recent JET results showing a weak difference in tungsten-driven soft x-ray peaking between LFS and HFS heating. Diffusive and convective transport of high-Z impurities in C-Mod are constrained by STRAHL simulations. Using TORIC in TRANSP to model the minority species and NEO and GKW to model the neoclassical and turbulent transport, a range mechanisms are investigated by which minority heating can impact the core radial impurity transport. While minority heating may modify the core peaking, volume averaged impurity content is controlled by radial flux at the ETB. Modeling suggests that for opaque scrape-off layers as expected in ITER, kinetic profiles will combine to result in outward neoclassical impurity flux between edge localized modes (ELMs). This important result stands in contrast to the widely observed behavior of quasi-stationary impurity flux between ELMs or in ELM-free H-modes to be directed inward, building up core impurity content. Experimental results from Alcator C-Mod suggest that this condition of outward impurity flux may be transiently accessed following a transition from I-mode to ELM-free H-mode. By tracking the time evolution impurities introduced prior to the H-mode transition, the direction of the impurity flux can be estimated from time-evolving STRAHL simulations of impurity spectroscopy. Initial results using this novel pedestal transport analysis technique are presented.
Speaker: Dr Matthew Reinke (ORNL)
• 08:30
Ion heating in magnetosphere plasma device RT-1 4h
While the stable high-beta (~ 1) confinement by a dipole magnetic field was successfully demonstrated with high-temperature electrons (Te > 10 keV) [1, 2], the heating of ions was a challenge. We have made two major progresses in this direction. (i) We developed a system for ion cyclotron resonance of frequency (ICRF) heating, and demonstrated the active heating of ions by launching a slow wave. The ion temperatures in the core region are increased in hydrogen, helium and deuterium plasmas. The differences of temperatures among ion species suggest a strong influence of the charge-exchange loss by which the bulk ions remain relatively cold (< 20 eV) in comparison with impurity ions. (ii) We also found a spontaneous heating mechanism concomitantly occurring with the up-hill diffusion [3, 4]. [1] H. Saitoh et al., Phys. Plasmas 21 (2014) 082511. [2] M. Nishiura et al., Nucl. Fusion 55 (2015) 053019. [3] N. Sato et al., http://arxiv.org/abs/1510.08571, in 2015. [4] Y. Kawazura et al., Phys. Plasmas 22 (2015) 112503.
Speaker: Dr Masaki Nishiura (The University of Tokyo)
• 08:30
ITB formation in gyrokinetic flux-driven ITG turbulence 4h
Profile stiffness is a long standing problem, which may limit the overall performance of H-mode plasmas. In the JET experiment, while strong temperature profile stiffness is observed around the nonlinear threshold of ion temperature gradient, it can be greatly reduced by co-current toroidal rotation in weak magnetic shear plasma. To understand such a mitigation mechanism of the stiffness, we investigate the impact of momentum injection on profile stiffness in flux-driven Ion Temperature Gradient (ITG) turbulence by means of a newly developed toroidal full-f gyrokinetic code GKNET. It is found that momentum injection can change the mean flow through the radial force balance, leading to Internal Transport Barrier (ITB) formation in which the ion thermal diffusivity decreases to the neoclassical transport level. Only co-current toroidal rotation can benefit the ITB formation in weak magnetic shear plasma, showing a qualitative agreement with the observations in the JET experiment. Note that the established ITB is enough stable in the quasi-steady state. The underlying mechanism is identified to originate from a resultant momentum flux. According to the non-local ballooning theory and momentum transport theory, the mean flow shear triggered by co-current toroidal rotation provides the momentum pinch, which can reduce the relaxation of both toroidal rotation and mean flow profiles. On the other hand, the role of counter rotation is opposite so that the relaxation is enhanced. Thus, there exists a positive feedback loop between the enhanced mean flow shear and resultant momentum pinch only in the co-current toroidal rotation case, signifying a favorite trend to ITB formation. Such a momentum pinch effect is also essential for ITB formation around the q_min surface in reversed magnetic shear plasma. We detect that the position of ITB is insensitive to the momentum source profile, which is determined only by the q_min surface. These results show a qualitative agreement with the observations in the JT-60U reversed shear discharges.
Speaker: Dr Kenji Imadera (Kyoto University)
• 08:30
L-H Transition Threshold Physics at Low Collisionality 4h
H-mode operation is the regime of choice for good confinement. Access to and sustainability of the H-mode requires understanding of the L $\rightarrow$ H transition power threshold and the related problem of hysteresis. To predict ITER transitions, one must also understand low collisionality, electron-heated regimes. In this paper, we discuss a.) L $\rightarrow$ H power threshold scaling including the minimum in P_th(n) and elucidate an impact of inter-species energy transfer on threshold physics, b.) transitions in collisionless, electron heated regimes where the electron-ion coupling is anomalous, due to the fluctuation of $\langle E\cdot J\rangle$ work on electrons and ions, c.) new transition scenarios, characterized by the sensitivity of transition evolution to pre-existing L-mode profiles. To study the above phenomena, we have developed a reduced model that independently evolves the collisionally coupled electron and ion temperatures, along with density, turbulence intensity and flow profiles. Our studies have revealed the physics of the power threshold minimum in density as a combined effect of the density dependence of collisionless electron-ion coupling and e-i heating mix. For collisionless regimes, we have included an anomalous power coupling between electrons and ions. Using a recently developed theory of minimum enstrophy relation which predicts a hyper-viscous turbulent flow damping we employ the nonlinear viscous heating of the ions. Our preliminary results on collisionless regimes suggest that L\rightarrow H transition occurs as the endstate of an anomalous electron-ion thermal coupling front. The transition occurs when the front arrives at the edge and impulsively raises T_i there, thus building up the diamagnetic electric field shear. This study highlights the importance of collisionless energy transfer process to transitions in regimes of ITER relevance. Finally, we have explored transitions occurring in the absence of turbulence driven shear flow. The key here is the sensitivity of the transition to the pre-existing L-mode density profile. Ongoing work focuses on elucidating this sensitivity and understanding how to exploit it to optimize the access to H-mode. This work was supported by the Department of Energy under Award No. DE-FG02-04ER54738.
Speaker: Dr Mikhail Malkov (University of California, San Diego)
• 08:30
Modulated heat pulse propagation and partial transport barriers in 3-dimensional chaotic magnetic fields 4h
The quantitative understanding of the role of magnetic field stochasticity on transport is critical for the confinement of fusion plasmas. Specific problems of interest include the control of ELMs by resonant magnetic perturbations and the assessment of heat fluxes at the divertor. Here we present direct numerical simulations of the time dependent parallel heat transport equation modeling heat pulses driven by power modulation in 3-D chaotic magnetic fields. Understanding this problem is important because effective diffusivities, advection velocities, and damping rates are often inferred from local measurements of the amplitude and the phase of the propagation of harmonic temperature perturbations. Heat pulse propagation has also been recently used to study the connection between transport and magnetic field bifurcations in modulated electron cyclotron heating perturbative experiments in LHD and DIII-D. The numerical results presented here provide conclusive evidence that even in the absence of magnetic flux surfaces, chaotic magnetic field configurations exhibit partial barriers to modulated heat transport. In particular, it is shown that high-order islands and remnants of destroyed flux surfaces (Cantori) act as partial “leaky” barriers that slow down or even stop the inward propagation of heat waves where the connection length exhibits a strong gradient. Motivated by recent experimental studies in LHD and DIII-D we also present preliminary results on modulated heat pulse propagation across magnetic islands. The geometry used in this calculation is toroidal and the magnetic field was obtained using the MHD equilibrium code SIESTA. It is shown that magnetic islands, in particular the elliptic (O) and the hyperbolic (X) points, have a direct impact on the spatio-temporal dependence of modulated heat pulses. The main computational challenge in the work presented here stems from the strong asymmetry between the parallel and perpendicular conductivities. To address this problem, we use a Lagrangian Greene's function (LG) method that bypasses known limitations of grid-based methods. To deal with time periodic sources, we present a novel reformulation of the LG method in Fourier space that is significantly more efficient than the original real space version of the method.
Speaker: Diego del-Castillo-Negrete (Fusion Energy Division. Oak Ridge National Laboratory)
• 08:30
Neoclassical Toroidal Plasma Viscosity with Effects of Finite Banana Width in Finite Aspect Ratio Tokamaks 4h
Theory for neoclassical toroidal plasma viscosity is to describe the transport processes, including particle, momentum, and energy transport fluxes in real tokamaks with broken symmetry. The predictions of the theory are in agreement with the numerical results in all collisionality regimes in the large aspect ratio limit. The theory has since been extended to finite aspect ratio tokamaks. The extension is made possible because the perturbed distribution function is localized in the phase space in the low collisionality regimes. Thus, the theory can be used to model transport phenomena including toroidal momentum relaxation in real finite aspect ratio tokamaks. However, there are cases where self-consistent magnetic perturbations have radial variations that are comparable to the width of bananas. To model the transport phenomena, the theory for has to be extended further to include the effects of the finite banana width. To that end, an orbit averaged drift kinetic equation has been developed to describe the transport processes in the low collisionality regimes, when the effective collission frequency is much less than the bounce frequency of bananas. The equation is now solved to calculate the neoclassical toroidal plasma viscosity, and, thus, the corresponding transport fluxes through the flux-force relation, to include the effects of finite banana width in various asymptotic limits. The resultant radial profile for the neoclassical toroidal plasma viscosity varies on the equilibrium scale even though the magnetic perturbations vary rapidly. The reason is that the bounce motion of the finite width of the bananas naturally smoothes out the short scale variations. This result is consistent with the experimental measurements.
Speaker: Dr K. C. Shaing (Engineering Physics Department, University of Wisconsin)
• 08:30
New Results in Negative Viscosity Models for Fusion Plasma Dynamics 4h
Negative viscosity phenomena in which turbulence driven by the heat flux couples its energy to large scale structure, is a familiar and, in fact, necessary element in the success of magnetic fusion. Two prime examples of negative viscosity phenomena are zonal flow formation, where drift wave turbulence drives mesoscale shear flows which regulate large scale eddys, and intrinsic torque, where turbulence with broken symmetry drives toroidal rotation without momentum input. These two negative viscosity phenomena are essential for good confinement and MHD control. Here, we present new results in negative viscosity models, namely condensation of profile corrugations and intrinsic rotation due to dynamical symmetry breaking. These results yield new fundamental insights into zonal flow pattern structure and the origins of intrinsic torque, particularly in weak shear regimes. This material is based upon work supported by the U.S. Department of Energy, Office of Science, Office of Fusion Energy Sciences, under Award Numbers DE-FG02-04ER54738 and DE-SC0008378.
Speaker: Prof. Patrick Diamond (University of California San Diego)
• 08:30
Nonlocal Plasma Response to Edge Perturbation in Tokamak 4h
The transient transport events are observed in toroidally magnetic confinement devices. For example, the cold pulse experiment shows a rapid transient increase of electron temperature in the plasma core in response to an abrupt cooling at the edge. Understanding the nonlocal transport is important to control plasma core and/or fuel supply in ITER and DEMO. The nonlocal particle transport has been investigated based on 4-field reduced MHD model applying edge density source. It is shown that 0/0 and 1/0 modes play an important role for nonlocal transport. The edge cooling is also investigated based on 3-field gyro-fluid model, however, the nonlocal transport is found to be limited in the peripheral region [1]. To identify the effects of density source and temperature sink in the edge region, we have extended 4-field model[2] to 5-field model which consists of the vorticity equation, Ohm’s law, parallel ion momentum equation, electron density equation and electron temperature equation. Simulation study on nonlocal plasma response in tokamak is performed using global fluid code based on the 5-field model. A nonlocal plasma response to edge perturbation is found. The simulation result shows that (i) the mean central electron temperature increases according to the edge cooling (it is shown for the first time), (ii) the magnetic island located at q=2 rational surface plays an important role (acting as internal transport barrier) as well as non-resonant modes such as 0/0 and 1/0 (here m/n implies Fourier mode with poloidal mode number m and toroidal mode number n), (iii) re-distribution of electron temperature occurs after switching off source and sink where meso-scale mode plays a major role. Reference [1] N. Miyato, et al., IAEA FEC2014, TH/P2-12 (2014). [2] M. Yagi, et al., Nucl. Fusion 45 900 (2005).
Speaker: Dr Masatoshi Yagi (National Institutes for Quantum and Radiological Science and Technology, Rokkasho Fusion Institute)
• 08:30
Numerical Diagnostics of Turbulent Transport in Three-Dimensional Magnetic Configurations 4h
Recent simulations in three-dimensional (3-D) magnetically confined plasmas show various aspects of plasma turbulence, and numerical diagnostics using 3-D simulation data of helical plasmas have been carried out. Here we present results of turbulence analyses (i) in simplified geometry for detailed nonlinear mechanism of heat transport, and (ii) in real 3-D geometry for comprehensive understanding of experimental observations. The former topic is on the case with MHD modes with a rather long wavelength. Global nonlinear simulations of drift-interchange modes in helical plasmas are carried out using a reduced MHD model. The model includes various characteristic time-scales. A ‘non-local’ effect has been studied in dynamical transport phenomena by the simulation of the heat source modulation. The nonlinear process plays the key role in the response, which takes a finite temporal duration for the energy redistribution. By conditional averaging the characteristic response can be extracted, such as a spiral 2-D pattern of the heat flux, which is formed nonlinearly and sustained for longer duration than the turbulence decorrelation time. An electromagnetic oscillation also exists, which gives the other time-scale. The latter topic is on the case with 3-D equilibrium and shorter wavelength perturbations. Gyrokinetic simulations of Ion-Temperature-Gradient modes in the LHD configuration have been carried out, and the 3-D data obtained from the simulations are analyzed taking into account of the line of sight of the experimental diagnostics. The method to resolve the local spectrum from the integrated signal (2-D Phase Contrast Imaging) is tested using the helical magnetic configuration. A vertical profile of the k spectrum can be obtained from the integrated signal. The analysis routine can give a fluctuation pattern at an arbitrary position, and comparison　with it gives understanding of the observed integrated signals. The extracted component has a peak at a finite wavenumber in this case. A finite width in the local k spectrum deteriorates the spatial resolution. In this way, variety of the numerical diagnostics from several view-points can give physical understanding and quantitative comparison of turbulent plasmas.
Speaker: Dr Naohiro Kasuya (Kyushu University)
• 08:30
Observation of an isothermal electron temperature profile with low recycling lithium walls in LTX 4h
Discharges with high edge electron temperatures and flat radial electron temperature profiles, extending to the last closed flux surface, and into the low field side scrape-off layer, have now been achieved in the Lithium Tokamak eXperiment (LTX), with lithium-coated walls. Flat temperature profiles are a long-predicted consequence of low recycling boundary conditions [S. Krasheninnikov, L. Zakharov, G. Pereverzev, Phys. Plasmas 10, 1678(2003)]. Temperature profiles are measured in repeated discharges with Thomson scattering; data from several discharges is averaged at each time point to improve accuracy at low density. Modeling indicates that the ion temperature profiles are also flat, which should eliminate temperature gradient–driven instabilities. The confined plasma therefore appears to be (separately) isothermal in the electron and ion populations. The edge density is very low, with a density profile which decreases approximately linearly with the poloidal flux. So far experiments are transient. Gas puffing is used to increase the plasma density. After gas injection stops, the discharge density is allowed to drop, and the edge is pumped by the low recycling lithium wall. The core impurity content, even in low density plasmas without fueling, and edge electron temperatures of 200 eV, remains low. Z(effective) is approximately 1.5, with most of the increase from oxygen, followed by carbon. The smallest fraction of the Z(effective) increase, especially in the core, is from lithium. An upgrade to LTX, which includes a 35A, 20 kV neutral beam injector to provide core fueling and auxiliary heating, is underway. Two beam systems have been loaned to LTX by Tri Alpha Energy. With core fueling provided by the neutral beam, an equilibrium similar to the “Isomak” [P. Catto and R. Hazeltine, Phys. Plasmas 13, 122508 (2006)] – a tokamak discharge in thermodynamic equilibrium, may be accessible in LTX, for the first time. A widened operational window, in both toroidal field and plasma current, is also planned, as well as eventual operation in diverted geometry. Results from the most recent experimental campaign will be described, as well as the upgraded configuration of LTX. This work was supported by USDoE contracts DE-AC02-09CH11466 and DE-AC05-00OR22725.
Speaker: Richard Majeski (Princeton Plasma Physics Lab)
• 08:30
Physics of Unlocked Tearing Modes and Disruption Avoidance by Feedback-Based Electromagnetic Torque Injection 4h
Resistive and quasilinear 3D modeling is improving the understanding of recent experiments [1] carried out in DIII-D and RFX-Mod, in which disruptions were prevented by means of electromagnetic torque applied by non-axisymmetric (3D) coils. We will report the 3D aspects of experimental observations in comparison with finite resistivity (tearing) response by MARS-F and a time dependent torque balance simulation by MARS-Q [2]. Even though these codes are based on linear perturbation theory, many experimental observations are consistent with these 3D model predictions. One is the discovery of stable formation of a multi-layered tearing structure just after avoiding the tearing mode locking. According to magnetic sensors and the internal profile perturbed signals such as toroidal rotation and ion temperature, the mode’s maximum perturbation is poloidally and radially concentrated toward the poloidal angle of the control coil location. This is completely different from the ideal-MHD based RWM response, which is nearly independent of the poloidal structure of the applied 3D fields. But, the resistive (tearing) plasma response by MARS-F in the presence of finite plasma resistivity predicts precisely this type of unique poloidal features. The large flow shear observed in experiments is considered as a possible mechanism to stably sustain the multi-layer structure for long duration. This flow shear buildup seems to be qualitatively consistent with the initial results of time dependent torque balance simulation with MARS-Q. The experimental results and their consistency with 3D MHD models suggest that the use of electromagnetic torque, applied by error field coils or ELM control coils, could help to avoid locked-mode disruptions in ITER. This work was supported in part by the US Department of Energy under DE-AC02-09CH114661, DE-FC02-04ER546984, DE-FG02-04ER547615, DE-AC05-06OR231006, and DE-AC05-00OR227257. [1] M. Okabayashi, IAEA Fusion Energy Conference 2014. (2014 IAEA conference, EX/P2-42, Publication in progress) [2] Y.Q. Liu et al, Phys. Plasmas 20, 042503 (2013)
Speaker: Dr Michio Okabayashi (Princeton Plasma Physics Laboratory)
• 08:30
Plasma profiles and impurity screening behavior of the high-field side scrape-off layer in near-double-null configurations: prospect for mitigating plasma-material interactions on RF actuators and first-wall components* 4h
The improved impurity screening characteristics of the high-field side scrape-off layer to local impurity sources, previously reported for single null geometries, is found to be retained in double null configurations - strengthening the argument for locating current drive and heating actuators on the high-field side. The high-field-side (HFS) scrape-off layer (SOL) is known to exhibit extremely low levels of cross-field transport [1] and excellent impurity screening characteristics [2] in single-null magnetic configurations. It has been proposed that future tokamaks should exploit these remarkable HFS characteristics to solve critical plasma-material interaction (PMI) and sustainment challenges – relocate all RF actuators and close-fitting wall structures to the HFS and employ near-double-null magnetic topologies, to precisely control plasma conditions at the antenna/plasma interface and mitigate the impact of PMI [3]. Dedicated experiments were performed on Alcator C-Mod during the 2015 experimental campaign to quantify impurity screening characteristics and scrape-off layer profiles in near-double-null configurations. Nitrogen screening by the HFS SOL is found to be a factor of 2.5 better than LFS in balanced double-null discharges, despite an extremely thin scrape-off layer. Impurity screening is found to be insensitive to current and Greenwald fraction. HFS impurity screening is least effective (only a factor of 1.5 improvement) in unbalanced double-null discharges that favor the active divertor in the direction of B×∇B. Unbalanced discharges that favor the most active divertor opposite the direction of B×∇B have excellent HFS screening characteristics, a factor of 5 better than LFS. The latter situation is particularly promising for the use of HFS RF actuators in I-mode plasmas – a high confinement, steady state, ELM-free regime that is accessible at high magnetic field to a large range of input power for this magnetic topology [4]. [1] N. Smick, et al., Nucl. Fusion 53 (2013) 02300; [2] G. McCracken, et al., Phys. Plasmas 4 (1997) 1681; [3] B. LaBombard, et al., Nucl. Fusion 55 (2015) 053020; [4] A. Hubbard, et al., IAEA FEC2014, paper EX/P6-22. *This material is based on work supported by U.S. Department of Energy, Office of Fusion Energy Sciences under Award Number DE-FC02-99ER54512 on Alcator C-Mod, a DoE Office of Science User Facility.
Speaker: Dr Brian LaBombard (MIT Plasma Science and Fusion Center)
• 08:30
Plasma Response to Sustainment with Imposed-dynamo Current Drive in HIT-SI and HIT-SI3 4h
The Helicity Injected Torus - Steady Inductive (HIT-SI) program studies efficient, steady-state current drive for magnetic confinement plasmas using a novel experimental method. Stable, high-beta spheromaks have been sustained using steady, inductive current drive which is significantly more efficient than RF or neutral beams when scaled to a reactor. Externally induced loop voltage and magnetic flux are oscillated together so that helicity and power injection are always positive, sustaining the edge plasma current indefinitely. Imposed-dynamo Current Drive (IDCD) theory further shows that the entire plasma current is sustained. The method is ideal for low aspect ratio, toroidal geometries and is compatible with closed flux surfaces. Experimental studies of spheromak plasmas sustained with IDCD have shown stable magnetic profiles with evidence of pressure confinement. New measurements show coherent motion of a stable spheromak in response to the imposed perturbations. On the original device two helicity injectors were mounted on either side of the spheromak and the injected mode spectrum was predominantly n=1 due to the geometry. Coherent, rigid motion indicates that the spheromak is stable and a lack of plasma-generated n=1 energy indicates that the maximum q is maintained below 1 during sustainment. Results from the HIT-SI3 device are also presented. Three inductive helicity injectors are mounted on one side of the spheromak flux conserver. Varying the relative injector phasing changes the injected mode spectrum which includes n = 2, 3, and higher modes. Spheromaks have been sustained with toroidal current three times greater than the quadrature sum of injector currents. This material is based upon work supported by the U.S. Department of Energy, Office of Science, Office of Fusion Energy Sciences, under Award Number DE-FG02-96ER54361.
Speaker: Dr Aaron Hossack (University of Washington)
• 08:30
Predicting Cross-Scale Self-Organization in Turbulent Magnetically Confined Plasmas 4h
Recently, we have developed a comprehensive mathematical and computational framework for the analysis and quantification of self-organization [1]. Application of this method may resolve some of the important issues of fusion plasmas such as prediction of changes in the pattern formation and transport properties, to name a few. We assume that the system self-organizes if its complexity (related to statistical prediction) increases with time. The experimental data consists of the ion-saturation current measurements performed by a moveable Langmuir probe located at the outboard mid-plane on the MAST device. Several different confinement regimes in MAST are analyzed: high-density L-Modes, Dithering H-Modes with heating power close to the threshold for L-H transition with intermittent edge localized modes (ELMs) and H-Modes with ELMs present. We show how the method is used for predicting the occurrence of the particular Mode and we discuss the possible applications of this method for other issues relevant for fusion plasmas. Also, we illustrate how the method is used to predict ELM bursts in the edge plasma, in time and in the spatial location. We further apply this framework in order to predict different bifurcations and dynamic regimes in the model of Stimulated Raman scattering (SRS) in plasma [3], a paradigm of three-wave interaction of great importance for inertial confinement fusion. This three-wave interaction is related to a nonlinear coupling of intense laser light (pump) to the electron plasma wave (EPW) and the scattered light, shifted in wavenumber and frequency. SRS belongs to a family of underdense plasma instabilities which can have a detrimental effect on efficiency of laser energy deposition into a fusion target. This particular SRS model has a very rich spatiotemporal dynamics and exhibits a paradigm of transition to spatiotemporal chaos via quasiperiodic and intermittent stages. We show that our self-organization framework predicts with great accuracy, pattern changes in time and space and occurrence of new dynamic regimes. In addition, the method quantifies each self-organizing state enabling precise characterization of self-organization process under change of different parameters of the system.
Speaker: Dr Milan Rajkovic (University of Belgrade, Institute of Nuclear Sciences Vinca)
• 08:30
Predictions of toroidal rotation and torque sources arising in non-axisymmetric perturbed magnetic fields in tokamaks 4h
Capabilities of the integrated framework consisting of TOPICS, OFMC, VMEC and FORTEC-3D, have been extended to calculate toroidal rotation in fully non-axisymmetric perturbed magnetic fields, for demonstrating operation scenarios in actual tokamak geometry and conditions. It was found for the first time that the toroidally localized perturbed fields due to the test blanket modules (TBMs) and the heating and diagnostic ports in ITER augment the neoclassical toroidal viscosity (NTV) significantly, while do not influence confinement of beam ions and alpha particles. The NTV takes up a large portion of total torque in ITER and fairly decelerates toroidal rotation. It was found that varying the numerical coefficient added to the intrinsic torque model by the residual stress within a factor of unity improves the reproducibility of toroidal rotation in JT-60U H-mode plasmas. This fact opens up access to reliable rotation predictions in H-mode plasmas.
Speaker: Dr Mitsuru Honda (Japan Atomic Energy Agency)
• 08:30
Progress in Theoretical RFP Studies: New Stimulated Helical Regimes and Similarities with Tokamak and Stellarator 4h
Recent theoretical studies of the reversed-field pinch (RFP) have demonstrated the possibility of stimulating new quasi-single helicity (QSH) regimes by allowing a small helical deformation of the magnetic boundary. In particular, 3D nonlinear MHD modeling predicted QSH states based on non-resonant helicities are predicted to be more resilient to magnetic stochasticity induced by secondary modes. These theoretical predictions have motivated a series of experiments in the RFX-mod device with applied magnetic perturbations (MPs). We present here the current state of the art in research on nonlinear MHD modeling of helical RFP regimes, and we highlight similarities of the helical RFP with the tokamak and the stellarator. Helical RFP states with different helicities are compared in terms of their core confinement properties. We consider in particular stimulated helical configurations obtained by applying small m=1 helical MPs with toroidal periodicity corresponding to non-resonant (n=6) and resonant (n=7 and n=8) kink-tearing modes. The safety factor profile as a function of the helical magnetic flux decreases monotonically for the non-resonant case, whereas a region of flat or reversed magnetic shear appears in the core for resonant configurations. When secondary MHD modes are taken into account, a magnetic topology characterized by a conserved helical core enclosed by a stochastic region is typically observed. The width of the conserved core region turns out to be the largest for the non-resonant configuration, in qualitative agreement with RFX-mod experiments suggesting that QSH regimes with n=6 dominant mode are endowed with a larger hot core than spontaneous n=7 QSH states. By using a finite-time Lyapunov exponent method applied for the first time to magnetic confinement configurations, barriers to the diffusion of magnetic field lines are diagnosed in the weakly stochastic region surrounding the conserved helical core in numerical QSH states. Similarly to what observed in the tokamak configuration, the sawtoothing dynamics of RFP plasmas is mitigated with the application of helical MPs. The confinement properties of the RFP edge, characterized by m=0 island chains at the q=0 reversal surface playing a role similar to edge islands in tokamak and stellarator plasmas, are also discussed.
Speaker: Dr Daniele Bonfiglio (Consorzio RFX, Padova, Italy)
• 08:30
Recent Advances in Stellarator Optimization 4h
Computational optimization has revolutionized the field of stellarator design. To date, optimizations have focused primarily on optimization of neoclassical confinement and ideal MHD stability, although limited optimization of other parameters has also been performed. One of the criticisms that has been levelled at this method of design is the complexity of the resultant field coils. Recently, a new coil optimization code - COILOPT++, which uses a spline instead of a Fourier representation of the coils, - was written and included in the STELLOPT suite of codes. The advantage of this method is that it allows the addition of real space constraints on the locations of the coils. The code has been tested by generating coil designs for optimized quasi-axisymmetric stellarator plasma configurations of different aspect ratios. As an initial exercise, a constraint that the windings be vertical was placed on large major radius half of the non-planar coils. Further constraints were also imposed that guaranteed that sector blanket modules could be removed from between the coils, enabling a sector maintenance scheme. Results of this exercise will be presented. New ideas on methods for the optimization of turbulent transport have garnered much attention since these methods have led to design concepts that are calculated to have reduced turbulent heat loss. We have explored possibilities for generating an experimental database to test whether the reduction in transport that is predicted is consistent with experimental observations. To this end, a series of equilibria that can be made in the now latent QUASAR experiment have been identified that will test the predicted transport scalings. Fast particle confinement studies aimed at developing a generalized optimization algorithm will also be discussed.
Speaker: Dr David Gates (PPPL)
• 08:30
Reconnection Heating Experiments and Simulations for Torus Plasma Merging Startup 4h
A series of merging experiments: TS-3, TS-4 and MAST made clear the promising characteristics of reconnection heating for merging formation of high-beta spherical tokamak (ST) and field-reversed configuration (FRC). We found the reconnection outflow produces MW-class (<30MW in TS-3) ion heating power based on the findings: (i) its ion heating energy that scales with square of the reconnecting magnetic field B_rec, (ii) its energy loss lower than 10%, (iii) its ion heating energy in the downstream 10 time larger than its electron heating energy at around X-point and (iv) low dependence of ion heating on the guide (toroidal) field B_g. Based on UK-Japan collaboration, we made the upscaled merging experiment in MAST and documented significant ion heating T_i~1.2keV by increasing B_rec to 0.2T. Its ion heating ~1.2keV and heating time 3-5ms are about four times higher and 50 times shorter than the conventional ion heating ~0.3keV and heating time ~200ms by the CS startup. An important finding is that the B_rec^2 scaling law of reconnection heating energy was successfully extended over 1.2keV under n_e~1.5x10^19 [m^-3]. It depends just on B_rec and with little dependence on the guide (toroidal) magnetic field B_g. During the ST merging, B_rec and B_g are almost equal to poloidal field B_p and tortoidal field B_t, respectively but both components of B_p and B_t reconnect during the two spheromak merging with opposing B_t for FRC formation. Since the reconnection accelerated ions up to 70% of the Alfven speed, the ion velocity scales with B_rec, so that the T_i increment and the reconnection heating energy scale with B_rec^2 under the constant n_e. It is noted that the reconnection heating does not depend on plasma size as long as the reconnection time is shorter than the plasma confinement time. This extended scaling law suggests that the merging startup will possibly realize the burning plasma temperature T_i >10 keV just by increasing B_rec over 0.6T. The merging/ reconnection heating will possibly provide a new direct route to burning plasma regimes without using any additional heating. This promising scaling leads us to new reconnection heating experiments for future direct access to burning plasma regime: TS-U in U. Tokyo and ST-40 in Tokamak Energy.
Speaker: Prof. Yasushi Ono (University of Tokyo)
• 08:30
Relation of plasma flow structures to particle tracer orbits 4h
Turbulence induced transport is one of the outstanding physics problems in plasma physics. In the turbulence induced transport issue, we began with the identification of turbulent flow structures using topological and geometric techniques on the framework of resistive MHD. The structure of the flow is filamentary. The filaments are vortices that are linked to the rational surfaces. At a given time some filamentary vortices located at the lowest rational surfaces close on themselves forming toroidal knots, we characterize them by the time they remain close loops, that is their life. Other filaments are broken and we characterize them by their length. In the case that an averaged poloidal flow is self-consistently included in the calculation, there are some new structures associated with the transport barriers created by the shear in the mean flow. Now, we want to relate these topological structures to properties of tracer particles within a framework of the continuous random walk (CTRW) approach. Vortices may cause some of the trapping of particles, while large scale flows may carry them from vortex to vortex. We focus on the relation between the trapping times and lifetimes of the flow structures and other detrapping mechanisms. The results indicate that most of the trappings that are completed during the calculation correspond to tracers trapped on broken filaments, including possible multiple trappings. The probability distribution function of the trapping times is then a function of the filament length, and has a lognormal character, like the distribution of filament lengths. In the case that an averaged poloidal flow is self-consistently included in the calculation, there is an increase in the tracer trapping due to the transport barriers created by the shear in the mean flow. The tracers trapped in the barriers do not follow the flow filaments linked to the magnetic field lines.
Speaker: Prof. Luis Garcia (Universidad Carlos III de Madrid)
• 08:30
Residual Stress and Momentum Transport in Electromagnetic ITG Turbulence 4h
We study how electromagnetic (EM) fluctuations impact on residual Reynolds stress in the context of the quasi-linear theory. Two-fluid model is employed to describe EM ion temperature gradient turbulence. Analyses show that not only the conventional parallel residual stress but also additional stress due to EM fluctuations strongly increase with plasma beta (=plasma thermal energy/magnetic energy), potentially leading to the strong enhancement of flow generation in high beta plasmas. We identify that this strong increase of residual stress originates from the reinforcement of radial k (=spectrally averaged parallel wavenumber) asymmetry due to the deformation of eigenfunctions near a rational surface.
Speaker: Dr Helen Kaang (National Fusion Research Institute)
• 08:30
Results from the Sheared-Flow Stabilized Z-Pinch and Scaling to Fusion Conditions 4h
The sheared-flow stabilized Z-pinch has been experimentally demonstrated to produce long-lived plasmas that satisfy radial force balance and are stable for thousands of exponential growth times. The sheared-flow stabilized Z-pinch has the potential to lead to a compact plasma confinement device that scales to fusion conditions. The stabilizing effect of a sheared axial flow on the m=1 kink instability in Z-pinches has been studied using ideal MHD theory to reveal that a sheared axial flow stabilizes the kink mode when the shear exceeds a threshold value. Following these theoretical results, the ZaP Flow Z-Pinch group at the University of Washington has been experimentally investigating the connection between flow shear and gross plasma stability. Plasma stability is diagnosed with azimuthal arrays of magnetic probes that measure the plasma's magnetic structure. Large magnetic fluctuations occur during pinch assembly, after which the amplitude and frequency of the magnetic fluctuations diminish. This stable behavior continues for an extended quiescent period. Plasma flow profiles are measured from the Doppler shift of plasma impurity lines. The experimental flow shear exceeds the theoretical threshold during the quiescent period. Scaling relations suggest that high energy density plasma and fusion conditions are possible in a compact design. Recent experiments with the upgraded ZaP-HD device have demonstrated the ability to increase the plasma parameters by compressing the plasma radius to smaller values than achieved with the ZaP device. Based on the successful results of ZaP and ZaP-HD, a new experiment FuZE is designed to scale the plasma parameters to fusion conditions. The project will focus on furthering our understanding of the physics with specific emphasis on the limitations of sheared flow stabilization and on the importance of kinetic effects at large drift speeds.
Speaker: Prof. Uri Shumlak (UsUWash)
• 08:30
Robust Estimation of Tokamak Energy Confinement Scaling through Geodesic Least Squares Regression 4h
The standard scaling law for the global energy confinement in H-mode tokamak plasmas provides a guideline for machine design and planning of operational scenarios. In addition, it is used as a benchmark to assess the quality of confinement in present experiments. However, owing to the complexity of the multi-machine data sets from which the scaling is derived, the coefficient estimates may vary considerably, depending on the regression method used to establish the fit. This necessitates a robust analysis of the confinement scaling law. For this purpose we have developed a new regression technique that is robust in the presence of significant uncertainty on all variables as well as the regression model, and that can handle outliers in the data. The technique, called geodesic least squares regression (GLS), is shown to yield consistent results in estimating the confinement scaling, in contrast to standard ordinary least squares (OLS). Furthermore, some of the coefficients estimated by GLS are significantly different from those given by OLS, particularly those corresponding to the geometrical properties of the plasma. Furthermore, the method is shown to enable direct estimation of the scaling in terms of physical dimensionless variables. Hence, GLS is a powerful regression method that is able to deliver the level of robustness required for scaling in multi-machine databases, and in general in situations with limited knowledge about the regression model or substantial measurement uncertainty on all variables. The method is easily implemented and is hoped to be also of use in many other applications involving regression analysis in fusion science.
Speaker: Prof. Geert Verdoolaege (Ghent University)
• 08:30
Robust H-mode Pedestal Compatibility with SOL and Divertor Plasma Constraints 4h
Experiments on DIII-D have advanced the physics basis for simultaneously achieving a high pressure H-mode pedestal for high core plasma confinement with a highly dissipative divertor for protection of plasma facing components in future reactor tokamaks. These studies show achievement of this goal is governed by the coupling of several pedestal and divertor processes including: 1) the pedestal density profile dependence on the recycling neutral ionization source, 2) the separatrix density required to achieve strong divertor dissipation for high power exhaust, 3) the direct effect of dissipative divertor operation on pedestal performance, and 4) maintaining adequate power flux across the separatrix for robust H-mode confinement. A closed divertor configuration is shown to reduce the core plasma density, even for divertor detachment onset, due to a reduction in the pedestal ionization source. The separatrix density is found to increase with power, resulting in a variation in the accessible ratio of separatrix to pedestal top density. Robust pedestal pressure is found compatible with dissipative divertor operation, as long as collisionality remains low enough for optimal MHD stability, and core radiation from impurity seeding is limited to maintain sufficient power flux across the separatrix. The results suggest that a robust pedestal may be compatible with highly dissipative divertor operation for the lower core collisionality expected in future larger tokamaks. Taken together the pedestal requirements imply that innovative divertor solutions will be required to obtain dissipative operation at lower core density as future tokamaks scale to larger size and higher field. This work is supported by the U.S. DOE under DE-FC02-04ER546981, DEAC52-07NA273442.
Speaker: Dr Anthony W. Leonard (USA)
• 08:30
Single Null Divertor in Negative Triangularity Tokamak 4h
Fusion research has to solve the power handling problem toward fusion demonstration reactor (DEMO). Tokamak plasma with negative triangularity and an outboard divertor X-point may offer such an opportunity as an innovative concept. The present paper extends this concept investigating single null negative triangularity tokamak (SN-NTT). Double null negative triangularity tokamak (DN-NTT) configurations feature quite high stable pedestals in the 1st region of ballooning stability but the vertical stability is an issue for the DN-NTT. Already with one outboard X-point in the SN-NTT internal modes set the pedestal height limit. The changes in ELM regime, pedestal structure and Mercier mode driven turbulence are major issues yet to be investigated for the negative triangularity tokamak concept. Studying such phenomena from first principles would require nonlinear, electromagnetic gyrokinetic simulations. While negative triangularity plasma has some favorable MHD property with respect to ELMs, the beta limit is relatively low. That is connected with the absence of magnetic well for elongated plasma cross-sections. However, negative triangularity tokamak configurations with optimized pressure gradient profiles can be stable for betaN>3 at elongation k=1.8 and internal inductance li=0.9, even in the absence of the magnetic well, with Mercier modes stabilized by magnetic shear in the SN-NTT with the optimal upper triangularity value close to zero. Apart from the ELM mitigation and satisfactory level of beta limits, negative triangularity tokamaks feature other possibilities for power handling such as naturally increased separatrix wetted area due to divertor location at larger radii and more flexible divertor configuration using PF coils inside the TF coil made of NbTi superconductor in the low field region. Negative triangularity experiments in TCV show a reduction in electron heat transport by a factor two compared with positive triangularity D-shaped configurations, which is partly explained by nonlinear gyrokinetic simulations. This configuration also allows better pumping accessibility due to larger conductance. Engineering restrictions on toroidal field (TF) coils at the high field side may not allow the TF shape conformal to negative triangularity plasma: more realistic race-track shaped TF coils are better compatible with the SN-NTT configuration.
Speaker: Dr Sergey Medvedev (Keldysh Institute of Applied Mathematics)
• 08:30
Snowflake Divertor Configuration Effects on Pedestal Stability and Edge Localized Modes in NSTX and DIII-D 4h
Analyses of snowflake (SF) divertor experiments in NSTX and DIII-D show that the SF divertor can increase edge magnetic shear and modify pressure profiles of the H-mode pedestal enabling pedestal stability control while maintaining good H-mode confinement (H_98y2~1). The scrape-off layer (SOL) geometry modifications lead to reduced peak temperature of plasma-facing components (PFC) via significant additional dissipation and partitioning of ELM heat fluxes between additional strike points. In NSTX, where pedestal stability operating condition was close to the kink/peeling boundary with the standard divertor and lithium conditioning, the SF divertor formation led to destabilization of large ELMs and a concomitant reduction of carbon concentration by 30-50% in the pedestal. In DIII-D, kinetic profiles were weakly affected by the SF configuration; a reduction in energy lost per ELM was observed and the ELM frequency was slightly increased. Planned linear MHD stability calculations will help understand the SF effects that appeared to depend on pedestal stability operating point. A reduction of ELM-induced divertor peak temperature Tsurf (and heat flux) in the SF divertor (cf. standard divertor) was observed in both NSTX and DIII-D experiments. The divertor ELM energy density in the SF configuration (cf. standard divertor) is reduced due to a combination of increased ELM ion transit time, power splitting between additional SF strike points, and additional dissipative losses, which are especially large in the high-density radiative SF divertor. The observed pedestal and SOL modifications are generally beneficial, and can be further developed into ELM control scenarios and ELM mitigation techniques. This w