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25th IAEA Fusion Energy Conference - IAEA CN-221

Europe/Moscow
Blue 1-5 (Hotel Park Inn Pribaltiyskaya)

Blue 1-5

Hotel Park Inn Pribaltiyskaya

Saint Petersburg, Russian Federation
Description
The 25th IAEA Fusion Energy Conference is being organized by the IAEA in cooperation with the State Atomic Energy Corporation “Rosatom”, Russian Federation. 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) and San Diego (2012).
    • 16:00 19:00
      Registration Blue Foyer

      Blue Foyer

      Hotel Park Inn Pribaltiyskaya

      Saint Petersburg, Russian Federation
    • 09:00 10:15
      Opening: O/1 Blue 1-5

      Blue 1-5

      Hotel Park Inn Pribaltiyskaya

      Saint Petersburg, Russian Federation
      Convener: Mr Vasily Gusev (Russian Federation)
      • 09:00
        Igor Kurchatov and the Russian Fusion Program 1h 15m
        The 110-th anniversary of Academician Igor Kurchatov was celebrated a year ago. Kurchatov belongs to the pleiad of greatest Russian scientists. His research and managerial activity determined further development of Nuclear Energy and Controlled Fusion both in Russia and in the world. In 1951 Kurchatov recognized the fundamental significance of high temperature plasma research performed by Igor Tamm and Andrei Sakharov and suggested the Government of USSR to use fusion neutrons for Pu239 and U233 breeding from U238 and Th232 as well as tritium breeding from Li6. His historic letter initiated plasma research in devices with straight magnetic field, mirror machines and in toroidal magnetic configurations with plasma current, later named ‘tokamaks’. Kurchatov clearly understood that building a magnetic fusion reactor was impossible without a deep knowledge of high temperature plasma physics. He appointed Academician Lev Artsimovich as the leader of the national fusion program and Academician Mikhail Leontovich as the leader of theoretical research. A national Fusion Program for the period from 2014 to 2030 has been formulated in Russia taking into account the present status of fusion research, results of ITER design and construction activity, trends in development of Nuclear Energy and prospects of Fusion Energy. The program includes two main directions of research and development: 1. active participation in the ITER project and supporting it by theoretical research and experimental activity on national facilities; 2. implementation of magnetic fusion achievements and innovative nuclear technologies for faster development of global Atomic Energy via supplementing it by Fusion-Fission Hybrid Systems (FFHS) capable of extending fissile recourses, improving safety, ecology and nonproliferation regime. The R&D program for hybrid systems and enabling technologies will be realized with the following milestones: 1. Design and construction of the demonstration fusion neutron source DEMO-FNS on the basis of a superconducting tokamak for tests of hybrid blankets and nuclear technologies by 2023; 2. Design and construction of the Pilot Hybrid Plant (PHP) by 2030.
        Speaker: Mr Evgeny VELIKHOV (National Research Centre "Kurchatov Institute")
        Slides
    • 10:15 10:45
      Coffee Break 30m
    • 10:45 12:30
      Overview 1: Magnetic Fusion: OV/1 Blue 1-5

      Blue 1-5

      Hotel Park Inn Pribaltiyskaya

      Saint Petersburg, Russian Federation
      Convener: Mr Aleksandr Burdakov (Russian Federation)
      • 10:45
        Fusion Research in Ioffe Institute 25m
        Overview of activity of the Plasma Physics Division at the Ioffe Institute in support of fusion program is presented. Experiments on LHCD (100kW, 2.5GHz) at the Globus-M tokamak (R=0.36m, a=0.24m, B_tor=0.4T, I_pl=200kA) with poloidally oriented grill resulted in RF driven current of up to 40 kA, in agreement with the theory predictions. At the FT-2 tokamak (R=0.56m, a=0.08m, B_tor=2.3T, I_pl=30kA) experiments with traditional toroidally oriented grill revealed no dependence of LHCD density limit on H/D ratio in spite of 3 times different LH resonance densities. Microwave Doppler Reflectometry (DR) at the Globus-M, and DR and Heavy Ion Beam Probe measurements at the tokamak TUMAN-3M (R=0.53m, a=0.24m, B_tor=1.0T, I_pl=190kA) demonstrated GAM suppression at the L-H transition. Observations at the FT-2 using Doppler Enhanced Scattering showed that GAM amplitude is anti-correlated both spatially and temporally with electron thermal diffusivity. For the first time turbulence amplitude modulation at GAM frequency was found both experimentally and in global gyrokinetic modeling. A model of L-H transition is proposed based on this effect. The loss mechanisms of energetic ions’ (EI) were investigated in the NBI experiments on Globus-M and TUMAN-3M: orbit losses, sawtooth triggered redistribution of EIs and Alfvenic mode excitation. Non-conservation of adiabatic invariant of EI in small aspect ratio configuration was found numerically to play a role in EI losses. Empirical scaling of 2.5 MeV DD neutron rate for the two devices shows strong dependence on toroidal field (B_tor)^1.29 and plasma current (I_pl)^1.34; this justifies B_tor and I_pl increase by a factor of 2.5 in proposed upgrade of Globus-M. Bursts of ~1MHz Alfvenic type oscillations correlating with sawtooth crashes were observed in OH at the TUMAN-3M. Possibility of low threshold parametric excitation of Bernstein and upper hybrid waves trapped in drift-wave eddies resulting in anomalous absorption in ECRH experiments in toroidal plasmas was discovered theoretically. A novel method of radial correlation Doppler reflectometry is shown to be capable of measuring the turbulence wave-number spectrum in realistic 2D geometry. Progress in design and fabrication of three diagnostics for ITER developed in Ioffe institute is reported: Neutral Particle Analysis, Divertor Thomson Scattering and Gamma Spectroscopy.
        Speaker: Mr Leonid Askinazi (Russian Federation)
      • 11:10
        The ITER Project Construction Status 25m
        The ITER Project has visibly made its transition to the construction phase in the two years since the last Fusion Energy Conference in San Diego. By mid-February 2014 commitments to in-kind procurement are approaching 89.6 percentage of the total credit value and 70.7 percentage (99 out of 140) in the number of Procurement Arrangements. Construction is accelerating and the appearance of the site is changing on a daily basis. Vigorous efforts are underway to mitigate some remaining organizational problems, including the alignment of the ITER Organization (IO) and 7 Domestic Agencies (DAs), minimizing any possible delay factor. The seven Parties are well committed to the construction of ITER. The ITER project has gone beyond the turning point. This fact should be understood and shared. We only can go forward together. The total progress being made is so enormous that it is impossible to do justice to everything in a single presentation. Therefore, the buildings, the core tokamak and some of the balance of plant will be reported on here. Most ancillary systems ranging from heating and current drive systems, diagnostics and fuelling systems to remote handling and the hot cell facility are in an advanced state of design. Their status will have to be reported at another time. The same is true of both the integration and assembly efforts.
        Speaker: Mr Osamu Motojima (ITER Organization)
        Slides
      • 11:35
        Overview of the JET Results 25m
        The European fusion programme is moving into the phase of implementation of its Roadmap. In this context, the JET programme has focused on consolidation of ITER design choices and preparation for ITER operation, with a specific emphasis given to the Bulk Tungsten Melt Experiment that has been crucial for the final decision on the material choice for the day-one tungsten divertor in ITER. In the first JET campaigns with the ITER-like Wall (ILW) fuel retention and material migration studies were a high priority. Now the focus of JET experiments has shifted towards integrated scenario development with the goal of addressing issues such as plasma-facing component (PFC) heat loads and W impurity accumulation in conjunction with high performance. In particular, during the last year, the importance of the magnetic geometry in the divertor area, strike point location and divertor pumping were established as key aspects for achieving good H-mode confinement, in combination with avoiding tungsten accumulation using ICRH. Moreover, significant effort was devoted to the use of impurity seeding to produce core-divertor compatible reference scenarios at good confinement which are essential for ITER, as well as high radiative scenarios which are required for DEMO. ITER-relevant conditions for steady-state operation have been achieved for over 7s at 2.5MA/2.7T and 21MW input power with H98(y,2)=0.85 and low divertor target power loads and partial detachment between ELMs. In parallel, post-mortem analyses of the PFCs retrieved from the first ILW campaigns have confirmed the previously reported low fuel retention obtained by gas balance. These studies show that the reduced material erosion and migration leads to reduced trapping of fuel in deposited Be layers which have less incorporated fuel in comparison with Carbon layers. In addition, the pattern of deposition within the divertor has changed significantly with the ILW in comparison with JET carbon wall campaigns due to the much-reduced level of chemical erosion. Transport to remote areas is almost absent, with the only significant Be deposits (15µm) found on the apron of the inner divertor. This work was supported by EURATOM and carried out within the framework of the European Fusion Development Agreement. The views and opinions expressed herein do not necessarily reflect those of the European Commission.
        Speaker: Mr Francesco Romanelli (European Commission)
        Slides
        summary
      • 12:00
        DIII-D Research to Address Key Challenges for ITER and Fusion Energy 25m
        The DIII-D tokamak has addressed key challenges in preparation for ITER and the next generation of fusion devices. The robustness of RMP-ELM suppression was demonstrated using as few of 5 of the usual 12 coils. QH mode was extended to 80% Greenwald density fraction, establishing its viability for ITER. Disruption mitigation led to relatively symmetric non-localized heat loads, while vertical displacement events were better ameliorated by earlier mitigation. Real time ray tracing and spectral mode identification techniques enabled more efficient tearing mode control. Promising candidates for steady state (SS) fusion were demonstrated, with a fully non-inductive hybrid scenario, a sustained high l_i scenario for ITER’s SS goal, a high q_{min} scenario with high β_N potential, and an 80% bootstrap scenario for EAST. Innovative boundary solutions were implemented with good performance achieved using a radiative divertor in ITER baseline and SS scenarios, and adapting snowflake divertor to the SS to further reduce heat flux. New divertor Thomson scattering enabled real time dual detachment/core density control, and characterized detachment in 2-D. Scrape off layer (SOL) profiles exhibited a critical gradient behavior consistent with ideal ballooning limits, while increasing connection length broadened the SOL and lowered divertor heat flux, highlighting the importance of cross-field transport. Low Z gas injection was found to help prevent erosion of high Z plasma facing surfaces. Underpinning this, DIII-D continued to expand the scientific basis of fusion for projection and optimization of future devices. Transport studies characterized turbulence in electron heated regimes to constrain predictive models of burning plasmas, and showed how long wavelength turbulence rises, with transitions from ITG to TEM increasing particle transport and density peaking. Fast ion transport due to Alfvénic modes was found consistent with a critical gradient model, while 3-D fields led to losses even over a single fast ion orbit. H-mode pedestal structure measurements highlighted the need for kinetic effects in full f simulations, while optimization of the pedestal exploited a predicted valley of improved stability to raise performance. Work supported in part by the US DOE under DE-FC02-04ER54698, DE-FG02-89ER54297, and DE-AC02-09CH11466.
        Speaker: Mr Richard J. Buttery (USA)
        Slides
    • 12:30 14:00
      Lunch Break 1h 30m
    • 14:00 16:10
      Overview 2: Inertial & Magnetic Fusion: OV/2 Blue 1-5

      Blue 1-5

      Hotel Park Inn Pribaltiyskaya

      Saint Petersburg, Russian Federation
      Convener: Mr Yong Liu (China)
      • 14:00
        Progress with the ITER Project Activity in Russia 20m
        Russian obligations in the ITER project consist of the development, manufacture, installation and putting into operation at the ITER site of 25 systems. At this stage Institution «Project center ITER» has signed with the ITER Organization 18 procurement arrangements for manufacture and supply of the equipment for ITER. Manufacture of the corresponding systems and development of other 7 systems is carrying out according the schedule of ITER construction. Nine unit lengths of Nb3Sn toroidal field conductor and nine unit lengths of NbTi poloidal field cables were supplied to European Union in 2013-2014 according to schedule. “Vezuvi-11M” gyrotron was tested in bench in NRC ”Kurchatov institute” where all required by ITER parameters (frequency - 170 GHz, power – 1 MW, duration - 1000 sec and efficiency - 52%) were reached in combination. 12 European and 6 Japanese mockups of divertor were tested in Efremov institute at thermal loads of 20 MW/m2. Technology of Russian Beryllium manufacture for ITER first wall was developed in collaboration between Bochvar institute (technology developer) and JSC “Bazalt” with participation of Efremov institute. It was proposed to install gamma-spectrometer behind the NPA at the same line of sight as NPA. Such combination of NPA and gamma-spectrometer will provide additional possibilities to increase accuracy of D/T ratio measurements and fast ion behaviour studies. In collaboration with Institution “Project center ITER” the NPA complex was added by diamond spectrometer of fast atoms that also will increase possibilities of fast ions studies. Results of analysis demonstrate essential increase of high field side reflectometry by adding low field side antenna to provide refractometry measurements. The prototypes of monocrystal Molybdenum mirror were manufactured for hydrogen line and charge exchange recombination spectroscopy. The prototypes of U-235 and U-238 fission chambers and compact diamond neutron spectrometers were developed and tested. The prototypes of spectrometers for CXRS are manufactured and tested.
        Speaker: Mr Anatoly Krasilnikov (Russian Federation)
        Slides
      • 14:20
        Recent ASDEX Upgrade Research in Support of ITER and DEMO 25m
        Research on ASDEX Upgrade is programmatically focused on resolving physics questions that are key to the successful operation of ITER as well as informing design choices for a future DEMO reactor. From 2014 on, a significant part of the ASDEX Upgrade programme is run under the EUROfusion MST programme. Using the flexible set of in-vessel helical perturbation coils, penetration of 3-d fields into the plasma is studied by analysing its impact on the edge plasma in L-mode discharges or examining the interaction with rotating core MHD modes. ELM mitigation by 3-d fields at high collisionality allows access to very small, high frequency ELMs at edge temperatures and pressure gradients higher than in the usual small ELM regime at high density. This cannot be described by the usual peeling-ballooning model for ELM stability which, in general, is shown to be inadequate to explain the whole variety of type I ELM observations. The study of the impact of a full metal first wall on plasma operation and performance continues to be a major research topic. Comparison with JET has confirmed many of our results. Characterisation of the ITER Q=10 scenario (standard H-mode at q95 = 3 and betaN = 1.8) revealed that with full metal wall, confinement quality is marginal for reaching H = 1. In this scenario, type I ELMs are large and first attempts using helical fields or pellets showed limited success in mitigation. With higher betaN, confinement regularly exceeds H = 1. These findings suggest that the optimum operational point for ITER Q = 10 might be at higher q95 and betaN, evolving towards the ‘improved H-mode’ regime under study in ASDEX Upgrade. Concerning research for DEMO, a major emphasis is put on the exhaust problem. Record values of Psep/R exceeding 7 MW/m (total power up to 23 MW) with simultaneous time averaged peak target power load < 5 MW/m2 have been demonstrated under partially detached conditions using feedback controlled N-seeding. Complete detachment at high input power P > 10 MW was achieved and extensive modelling helps to clarify the mechanisms behind the experimentally observed high density zone in the divertor plasma as well as the stable X-point radiation. Using double feedback on N and Ar impurity seeding, a value of Prad,core/Ptot approx. 70 % as will be needed for DEMO was demonstrated at very good plasma performance. Finally, we will report on operational experience with two rings of tiles consisting of ferritic steel on the high field side heat shield to analyse the implications of a possible use of bare EUROFER wall panels on DEMO and the use of newly installed rings of solid W tiles in the outer divertor strike zones. All these studies are accompanied by progress in the understanding of the underlying fusion plasma physics, which is essential to obtain true predictive capability, and will be discussed in the contribution. This project has received funding from the EURATOM research and training programme 2014-2018.
        Speaker: Mr Hartmut Zohm (Germany)
        Slides
      • 14:45
        Overview of Transport and MHD Stability Study and Impact of Magnetic Field Topology in the Large Helical Device 25m
        The progress of physics understanding and concurrent parameter extension since the last IAEA-FEC 2012 [1] in the Large Helical Device is overviewed. Recently the plasma with high ion and electron temperature (Ti(0) ~ Te(0) ~ 6keV) is obtained by the combination of 1) a reduction of wall recycling and neutrals by Helium ICRF discharges and 2) optimization of carbon pellet injection and on-axis ECH. The temperature regime obtained is significantly extended. The stochastic magnetic field in the plasma core which causes flattening of the temperature and the plasma flow can be eliminated by the control of the magnetic shear by NBCD and ECCD and a high central ion temperature (Ti(0) ~ 8keV) discharge is achieved by overcoming the core temperature flattening frequently observed in the plasma with an ion- ITB. After the formation of the ion-ITB, the residual stress switches the sign from the counter- to the co-direction and results in a large toroidal flow in the co-direction [2]. On the other hand, the radial convective velocity of the carbon impurity (Vc) also changes sign from inward to outward and this reversal of convection causes the extremely hollow impurity profile (called impurity hole). A stochastization of the magnetic field affects the MHD instability driven by a pressure gradient. This is an interesting topic because the stochastization of the magnetic field is also a key issue in the resonant magnetic perturbation (RMP) experiment for edge localized mode (ELM) suppression. When the stochastization of the magnetic field is enhanced by a RMP, the pressure driven mode is suppressed even without a change in the pressure gradient itself. In LHD, a low mode (n/m=1/1) magnetic island exists near the plasma periphery and the width of the magnetic island can be controlled by the RMP. By injecting hydrogen pellets into the O-point of the magnetic island, a significant peaked pressure profile inside the magnetic island is produced for a relatively long time, which is similar to the phenomena of a “snake” in a tokamak. Inside the LCFS, the stochastization causes the damping of flows, while it enhances the E×B flow due to the electron loss to the wall along the magnetic field. [1] O.Kaneko, et. al., Nucl. Fusion 53 (2013) 104015. [2] K.Ida, et. al., Phys. Rev. Lett. 111 (2013) 055001.
        Speaker: Mr Katsumi Ida (Japan)
        Slides
        summary
      • 15:10
        Overview of MAST Results 25m
        MAST addresses key issues for ITER and DEMO. Mitigation of ELMs with resonant magnetic perturbations (RMPs) with n=2,3,4,6 has been demonstrated: at higher and lower collisionality; for the first ELM; during the current ramp-up; when a sub-set of in-vessel coils fail; and with rotating n=3 RMPs. n=4,6 fields cause less braking whilst the power to access H-mode is less with n=4 than n=3,6. Refuelling with gas or pellets gives plasmas with mitigated ELMs and reduced peak heat flux whilst less than 10% drop in stored energy. The 3d structure of the post-pellet plasmoid has been imaged, with increased fluctuations during pellet ablation. A synergy exists between pellet-fuelling and RMPs, since mitigated ELMs remove fewer particles. JOREK and CAS3D stability codes show that 3d deformations influence peeling-ballooning stability. ELM precursors strikingly observed with Doppler-backscattering (DBS) and beam emission spectroscopy (BES) are consistent with gyrokinetic simulations of micro-tearing modes (MTMs) in the pedestal. Global gyrokinetic runs show kinetic ballooning modes mediate the pedestal width, whilst nonlinear simulations suggest that MTMs carry significant electron heat flux. A scan in beta at the L-H transition shows that pedestal height scales strongly with core pressure. The observed tilt of low-k turbulent vortices increases with flow shear, due to a decrease in poloidal wave number. Fast ion redistribution by fast particle modes depends on density, and access to a quiescent domain with ‘classical’ fast ion transport is found above a critical density. Highly efficient electron Bernstein wave (EBW) current drive (1A/W) has been achieved in solenoid-free start-up. A new proton detector has characterised escaping fusion products. Langmuir probes and a high-speed camera suggest filaments play a role in particle transport in the private flux region whilst coherence imaging has measured scrape-off layer (SOL) flows. BOUT++ simulations show that fluxes due to filaments are strongly dependent on resistivity and magnetic geometry of the SOL, with higher radial fluxes at higher resistivity. MAST Upgrade is due to operate in 2015 to support ITER preparation and importantly to operate with a Super-X divertor to test extended leg concepts for particle and power exhaust. This work was part-funded by the RCUK Energy Programme and the EU Horizon 2020 programme
        Speaker: Mr Ian Chapman (UK)
        summary
      • 15:35
        Alcator C-Mod: Research in Support of ITER and Steps Beyond 25m
        Alcator C-Mod studies high-field, high-performance diverted plasmas in support of ITER and steps beyond, focussing on RF and microwave tools for heating and current drive, with all metal high-Z plasma-facing components. Stability analysis of pedestals in the I-mode regime finds that pressures are well below the peeling-ballooning limit, as well as expected kinetic ballooning mode thresholds, likely explaining the generally observed lack of ELMs. Experiments using a new field-aligned ICRF antenna, which is rotated to align with the total local magnetic field, show dramatic reductions high-Z metallic impurity generation, and reduced core contanimation. This antenna also shows improvement in load-tolerance, particularly through edge transients. Implementation of novel “mirror-probe” electronics has enabled simultaneous measurements of Te, ne and phi with 1 micro-s time response using a single probe tip. Studies with this setup connected to a fast-scanning probe have revealed important properties of the Quasi-Coherent-Mode (QCM) which regulates edge particle transport in EDA H-mode. An Accelerator-based In-situ Material Surveillance diagnostic has been deployed, and the first between-shot measurements of surface evolution on areas of the inner-wall and above the divertor strike point have been made. Boron surface evolution and deuterium retention have been successfully tracked through a series of tokamak and wall-conditioning experiments. We have observed strong suppression of boundary turbulence and significant E improvement by injecting LH power into high-density H-modes, with H-factor increases up to 30%. An advanced lower hybrid RF (LHRF) actuator component has been implemented in the Integrated Plasma Simulator, and used to simulate modification of sawteeth via LHCD. Gyrokinetic simulations of TEM turbulence find a factor of two nonlinear upshift of the TEM critical density gradient, in close quantitative agreement with experiments. Upgrades which are ready for construction and operation in 2015 include: an actively heated (900 K) advanced tungsten DEMO-like outer divertor; an off-midplane LH launcher to test theories of improved penetration, absorption and current drive at high density, combined with LH source power upgrades to 4MW; and a second magnetic field-aligned ICRF antenna. Work supported by US Department of Energy
        Speaker: Mr Earl Marmar (USA)
        Slides
    • 14:00 18:45
      Overview Posters: OV/P Green 8-9

      Green 8-9

      Hotel Park Inn Pribaltiyskaya

      Saint Petersburg, Russian Federation
      • 14:00
        3D Effects of Edge Magnetic Field Configuration on Divertor/SOL Transport and Optimization Possibilities for a Future Reactor 4h 45m
        This paper overviews recent progress on the experimental identification and physics interpretation of 3D effects of magnetic field geometry/topology on divertor transport in helical devices and tolamaks with RMP. The 3D effects are elucidated as a consequence of competition between transports parallel (//) and perpendicular to magnetic field, in open field lines cut by divertor plates or in magnetic islands. The competition process has strong impacts on divertor functions, density regime, impurity screening, and detachment stability. Based on experiments and numerical simulations, key parameters (indicated with [ ] below) governing the transport process are discussed suggesting demanding issues to be addressed for divertor optimization in future reactors. The divertor density regime, which is known for strong up- and down-stream coupling, high-recycling regime in 2D axi-symmetric configurations, is affected by the 3D configuration. In W7-AS, LHD, TEXTOR-DED and HSX, the dependency is weakened. This is due to enhanced loss of //-momentum or of //-conduction energy. The dependency is functions of magnetic geometry parameters, [field line connection length], [poloidal wave length of RMP] and [Br/Bt]. Impurity screening is observed in Tore Supra, LHD, TEXTOR-DED with edge stochastization, and in W7-AS/X, TJ-II with island divertor. The enhanced outward particle flux due to [Br] provides the screening via friction force exerted on impurity. It is also found that suppression of ion thermal force, in the case of small [Br/Bt]~1.e-4, is responsible for the screening. The systematic study in TEXTOR-DED and LHD has shown that [a thicker stochastic region] provides better screening effects. In W7-AS and LHD, the larger [edge island width] leads to detachment stabilization. This is due to capturing of radiation with the islands and to the decoupling of edge and core plasmas in terms of core fueling of plasma/impurity. In TEXTOR-DED, [rotating RMP] fields result in density limit extension, avoiding MARFE onset. This is caused by spreading of recycling region, preventing edge cooling localization by recycling neutral/impurity. Systematic understandings of the 3D effects of edge magnetic field based on the key parameters shown above will offer new perspectives on divertor optimization for future reactors, which are not available in 2D axi-symmetric configuration.
        Speaker: Mr Masahiro Kobayashi (Japan)
      • 14:00
        Advance of H-Mode Physics for Long-Pulse Operation on EAST 4h 45m
        Since the last IAEA-FEC, significant progress has been made on EAST on both physics and technology fronts towards the long-pulse operation of high-confinement plasma regimes. EAST has been upgraded with more than 25 MW of CW heating & current drive power, along with 70 diagnostics, two internal cryopumps, an ITER-like W monoblock top divertor and resonant magnetic perturbation coils, which will enable EAST to investigate long-pulse H-mode operation with dominant electron heating and low torque input, which will be facing challenges on some of critical issues on ITER. New information has been obtained on the physics of L-H transition. Remarkable efforts have been made in mitigating type-I ELMs in a stationary state H-mode plasma with multi-pulses of supersonic molecular beam injection (SMBI), LHCD, lithium granule and deuterium pellet injection, as well as RMPs, thus potentially offering a valuable means of heat-flux control for next-step long-pulse fusion devices. Long-pulse H-mode discharges with H(98,y2) ~1 have been obtained either with ELM mitigation or in a small ELMy regime accompanied by a new electrostatic edge coherent mode, which appears in the steep-gradient pedestal region and plays a dominant role in driving heat and particles outwards. High peak heat load on the divertor due to type I ELMs, is reduced either by SMBI or LHCD. We find that ELM mitigation with SMBI is due to enhanced particle transport in the pedestal, correlated with large-scale turbulence and strongly anti-correlated with small-scale turbulence, while LHCD induces edge plasma ergodization, broadening the heat deposition footprint. Challenges and progresses on plasma control, effective H&CD, plasma-wall interactions under long-pulse, high-heat flux and high-Z metal wall conditions will also be presented.
        Speaker: Mr Baonian Wan (China)
      • 14:00
        Alcator C-Mod: Research in Support of ITER and Steps Beyond 4h 45m
        Alcator C-Mod studies high-field, high-performance diverted plasmas in support of ITER and steps beyond, focussing on RF and microwave tools for heating and current drive, with all metal high-Z plasma-facing components. Stability analysis of pedestals in the I-mode regime finds that pressures are well below the peeling-ballooning limit, as well as expected kinetic ballooning mode thresholds, likely explaining the generally observed lack of ELMs. Experiments using a new field-aligned ICRF antenna, which is rotated to align with the total local magnetic field, show dramatic reductions high-Z metallic impurity generation, and reduced core contanimation. This antenna also shows improvement in load-tolerance, particularly through edge transients. Implementation of novel “mirror-probe” electronics has enabled simultaneous measurements of Te, ne and phi with 1 micro-s time response using a single probe tip. Studies with this setup connected to a fast-scanning probe have revealed important properties of the Quasi-Coherent-Mode (QCM) which regulates edge particle transport in EDA H-mode. An Accelerator-based In-situ Material Surveillance diagnostic has been deployed, and the first between-shot measurements of surface evolution on areas of the inner-wall and above the divertor strike point have been made. Boron surface evolution and deuterium retention have been successfully tracked through a series of tokamak and wall-conditioning experiments. We have observed strong suppression of boundary turbulence and significant E improvement by injecting LH power into high-density H-modes, with H-factor increases up to 30%. An advanced lower hybrid RF (LHRF) actuator component has been implemented in the Integrated Plasma Simulator, and used to simulate modification of sawteeth via LHCD. Gyrokinetic simulations of TEM turbulence find a factor of two nonlinear upshift of the TEM critical density gradient, in close quantitative agreement with experiments. Upgrades which are ready for construction and operation in 2015 include: an actively heated (900 K) advanced tungsten DEMO-like outer divertor; an off-midplane LH launcher to test theories of improved penetration, absorption and current drive at high density, combined with LH source power upgrades to 4MW; and a second magnetic field-aligned ICRF antenna. Work supported by US Department of Energy
        Speaker: Mr Earl Marmar (USA)
      • 14:00
        Contribution to Fusion Research from IAEA Coordinated Research Projects and Joint Experiments 4h 45m
        IAEA Coordinated Research Projects (CRP) on “Utilisation of a Network of Small Magnetic Confinement Fusion Devices for Mainstream Fusion Research” and “Conceptual development of steady-state compact fusion neutron sources” continue to contribute to the mainstream Fusion Research. These CRPs join participants from 18 IAEA member states, who perform experiments and present results of individual and coordinated research and results of IAEA Joint Experiments (JE) at Research Coordinating Meetings, International Conferences and in publications. These activities also create platform for building long term relationships between scientists from developing and developed countries in fusion science and technology. The objective of the CRP “Conceptual development of steady-state compact fusion neutron sources” (CFNS) is to support the research on the development of steady-state compact fusion neutron sources for scientific, technological and nuclear energy applications. This research provides concepts and conceptual designs for low and high power CFNS; determines operational domains with optimized plasma performance and aimed on the development of a scientific and technological basis and comprehensive safety analysis for the proposed CFNS. The objective of the CRP “Utilization of a Network of Small Magnetic Confinement Fusion Devices for Mainstream Fusion Research” is to streamline results of studies on small fusion devices to mainstream fusion research by establishing a network of cooperation enabling coordinated investigations relevant to physics, diagnostics and technology issues of next fusion devices such as ITER and DEMO. 6 IAEA JEs have been carried out. In the recent JEs, studies have been extended from characterization of plasma turbulence and correlation between the occurrence of transport barriers, improved confinement with electrostatic turbulence, to characterization of the plasma pedestal in ohmic and NB heated H-mode discharges, NBI-induced Alfven Eigenmodes, studies of microwave emission, relation between halo currents and Ip 3D asymmetries during disruptions, evaluation of parallel electron power flux density using Langmuir and Ball-pen probes, RF pre-ionisation and investigation of the use of high-temperature superconductors in tokamak magnets. Outputs of these activities will be overviewed and results of the recent JEs will be presented in detail.
        Speaker: Prof. Mikhail Gryaznevich (Tokamak Energy Ltd; 6Technical University of Denmark, DTU Risø, Roskilde, Denmark; Imperial College, London, UK)
        summary
      • 14:00
        DIII-D Research to Address Key Challenges for ITER and Fusion Energy 4h 45m
        The DIII-D tokamak has addressed key challenges in preparation for ITER and the next generation of fusion devices. The robustness of RMP-ELM suppression was demonstrated using as few of 5 of the usual 12 coils. QH mode was extended to 80% Greenwald density fraction, establishing its viability for ITER. Disruption mitigation led to relatively symmetric non-localized heat loads, while vertical displacement events were better ameliorated by earlier mitigation. Real time ray tracing and spectral mode identification techniques enabled more efficient tearing mode control. Promising candidates for steady state (SS) fusion were demonstrated, with a fully non-inductive hybrid scenario, a sustained high l_i scenario for ITER’s SS goal, a high q_{min} scenario with high β_N potential, and an 80% bootstrap scenario for EAST. Innovative boundary solutions were implemented with good performance achieved using a radiative divertor in ITER baseline and SS scenarios, and adapting snowflake divertor to the SS to further reduce heat flux. New divertor Thomson scattering enabled real time dual detachment/core density control, and characterized detachment in 2-D. Scrape off layer (SOL) profiles exhibited a critical gradient behavior consistent with ideal ballooning limits, while increasing connection length broadened the SOL and lowered divertor heat flux, highlighting the importance of cross-field transport. Low Z gas injection was found to help prevent erosion of high Z plasma facing surfaces. Underpinning this, DIII-D continued to expand the scientific basis of fusion for projection and optimization of future devices. Transport studies characterized turbulence in electron heated regimes to constrain predictive models of burning plasmas, and showed how long wavelength turbulence rises, with transitions from ITG to TEM increasing particle transport and density peaking. Fast ion transport due to Alfvénic modes was found consistent with a critical gradient model, while 3-D fields led to losses even over a single fast ion orbit. H-mode pedestal structure measurements highlighted the need for kinetic effects in full f simulations, while optimization of the pedestal exploited a predicted valley of improved stability to raise performance. Work supported in part by the US DOE under DE-FC02-04ER54698, DE-FG02-89ER54297, and DE-AC02-09CH11466.
        Speaker: Mr Richard J. Buttery (USA)
      • 14:00
        Fast Ignition Realization EXperiment (FIREX) and Prospect to Inertial Fusion Energy in Japan 4h 45m
        Fast ignition has high potential to ignite a fusion fuel with only about one tenth of laser energy necessary for the central ignition. One of the most advanced fast ignition programs is the Fast Ignition Realization Experiment (FIREX). The goal of its first phase is to demonstrate ignition temperature of 5 keV, followed by the second phase to demonstrate ignition-and-burn. Relativistic fast electrons as the energy carrier, however, unfavorably diverge at high laser intensities necessary for significant heating. This difficulty is overcome by kilo-Tesla magnetic field collimating fast electrons towards a compressed fuel. Such super-strong field has been created with a capacitor-coil target driven by a high power laser, and subsequent collimation has also been demonstrated, suggesting that one can achieve ignition temperature at the laser energy available in FIREX. Repetitive creation of fast ignition plasmas has been demonstrated together with the technology development of high-efficient rep lasers and pellet injection, tracking, and beam steering.
        Speaker: Mr Hiroshi AZECHI (Japan)
      • 14:00
        First Experiments in SST-1 4h 45m
        Steady State Superconducting Tokamak (SST-1) has been commissioned after the successful experimental and engineering validations of its critical sub-systems. During the `engineering validation phase’ of SST-1; the cryostat was demonstrated to be leak tight to superconducting magnets system operations in all operational scenarios, the 80 K thermal shield was demonstrated to be uniformly cooled without regions of `thermal run away and hot spots’, the superconducting Toroidal Field (TF) magnets were demonstrated to be cooled to their nominal operational conditions and charged up to 1.5 T of field at the major radius, the assembled SST-1 machine shell was demonstrated to be a graded, stress-strain optimized and distributed thermo-mechanical device and the integrated vacuum vessel was demonstrated to be UHV compatible etc. Subsequently, `field error components’ in SST-1 were measured to be acceptable towards plasma discharges. A successful break-down in SST-1 was obtained in SST-1 in June 2013 assisted with electron cyclotron pre-ionization in second harmonic mode, thus marking the `First Plasma’ in SST-1 and arrival of SST-1 into the league of contemporary steady state devices as well. Subsequent to the first plasma, both physical experiments and boosting of engineering parameters in SST-1 have begun. A successful plasma start-up with E ~ 0.4 V/m, plasma current in excess of 50 kA for 100 ms assisted with ECH pre-ionization in second harmonic at a field of 0.75 T have been achieved. Lengthening the plasma pulse duration with LHCD, plasma current boosting up with ECH assisted pre-ionization in fundamental mode at 1.5 T apart from advance plasma physics experiments are presently being attempted in SST-1. In parallel, SST-1 has demonstrated in unique fashion pure cold gas cooling based nominal operations of its vapour cooled TF current leads up to 4650 A corresponding to 1.5 T of field in the plasma major radius. SST-1 has also achieved the distinction of being the only superconducting Tokamak in the world where the cable-in-conduit-conductor (CICC) based TF magnets are operated with helium cooling in Two-Phase mode during the plasma discharges up to 2.0 T of field at the plasma major radius.
        Speaker: Mr Subrata Pradhan (India)
      • 14:00
        Fusion Research in Ioffe Institute 4h 45m
        Overview of activity of the Plasma Physics Division at the Ioffe Institute in support of fusion program is presented. Experiments on LHCD (100kW, 2.5GHz) at the Globus-M tokamak (R=0.36m, a=0.24m, B_tor=0.4T, I_pl=200kA) with poloidally oriented grill resulted in RF driven current of up to 40 kA, in agreement with the theory predictions. At the FT-2 tokamak (R=0.56m, a=0.08m, B_tor=2.3T, I_pl=30kA) experiments with traditional toroidally oriented grill revealed no dependence of LHCD density limit on H/D ratio in spite of 3 times different LH resonance densities. Microwave Doppler Reflectometry (DR) at the Globus-M, and DR and Heavy Ion Beam Probe measurements at the tokamak TUMAN-3M (R=0.53m, a=0.24m, B_tor=1.0T, I_pl=190kA) demonstrated GAM suppression at the L-H transition. Observations at the FT-2 using Doppler Enhanced Scattering showed that GAM amplitude is anti-correlated both spatially and temporally with electron thermal diffusivity. For the first time turbulence amplitude modulation at GAM frequency was found both experimentally and in global gyrokinetic modeling. A model of L-H transition is proposed based on this effect. The loss mechanisms of energetic ions’ (EI) were investigated in the NBI experiments on Globus-M and TUMAN-3M: orbit losses, sawtooth triggered redistribution of EIs and Alfvenic mode excitation. Non-conservation of adiabatic invariant of EI in small aspect ratio configuration was found numerically to play a role in EI losses. Empirical scaling of 2.5 MeV DD neutron rate for the two devices shows strong dependence on toroidal field (B_tor)^1.29 and plasma current (I_pl)^1.34; this justifies B_tor and I_pl increase by a factor of 2.5 in proposed upgrade of Globus-M. Bursts of ~1MHz Alfvenic type oscillations correlating with sawtooth crashes were observed in OH at the TUMAN-3M. Possibility of low threshold parametric excitation of Bernstein and upper hybrid waves trapped in drift-wave eddies resulting in anomalous absorption in ECRH experiments in toroidal plasmas was discovered theoretically. A novel method of radial correlation Doppler reflectometry is shown to be capable of measuring the turbulence wave-number spectrum in realistic 2D geometry. Progress in design and fabrication of three diagnostics for ITER developed in Ioffe institute is reported: Neutral Particle Analysis, Divertor Thomson Scattering and Gamma Spectroscopy.
        Speaker: Mr Leonid Askinazi (Russian Federation)
      • 14:00
        Overview of DEMO Activities of IFERC Project in BA Activities 4h 45m
        In order to complement ITER and contribute to an early realization of the DEMO reactor, International Fusion Energy Research Centre (IFERC) implements DEMO Design and Research and Development activities. The design basis for DEMO and the sensitivity to underlying physics and technology assumptions has been reviewed through a 3-year collaboration. At this point, the plasma major radius and net electricity have been found to vary in a range of 8-9 m and 0.3-0.5 GWe, respectively. The tentative target is to set the start points of DEMO1 (pulsed) and DEMO2 (steady state) with low fusion power (≤ 2 GW) for compatibility with divertor power handling. So far, DEMO1 with a major radius (R_p) of 9 m, net electricity (P_e,net) of 0.5 GWe and pulse length of about 2 hours, and DEMO2 with R_p ~ 8 m and P_e,net = 0.3-0.5 GWe have been proposed. Comparative studies on the remote maintenance methods are on going, and the modeling and event selection for the safety analysis of upper bounding sequences is in progress. Research and Development of common components for DEMO blanket have been performed in 5 task areas; 1) a long-term exposure test of organic compounds into tritium water for up to 2 years have been carried by JA without serious effects, and analysis of JET tile will be implemented under EU/JA collaboration in 2014, 2) the impact of heat treatment conditions on RAFM (Reduced Martensite Ferric Material) [F84H] steel properties and assessment of irradiation correlation are examined by JA, and the fabrication technology of RAFM [Eurofer] has been studied by EU, 3) using rotating electrode method apparatus, a series of trial tests for pebbles of beryllide for advanced neutron multiplier is carried out by JA to optimize the fabrication conditions, and EU proceeds with the fabrication of Be-Ti rods by hot extrusion of milled material in steel jackets and the characterization of the beryllides, 4) fabrication and the test of advanced tritium breeder pebbles, such as Li_2TiO_3 with excess Li are implemented by JA, and EU proceeds with production and characterization of Lithium orthosilicate pebbles with secondary phase of 20, 25 and 30 mol% lithium metaltitanate, and 5) EU/JA collaboration activities have progressed in the studies of characterization of electrical resistivity, He and H permeability and radiation damage effects of SiC_f/SiC composites progress.
        Speaker: Mr Noriyoshi Nakajima (Japan)
      • 14:00
        Overview of Gyrokinetic Studies on Electromagnetic Turbulence 4h 45m
        Recent results on electromagnetic turbulence from gyrokinetic studies in different magnetic configurations are overviewed, showing the characteristics of electromagnetic turbulence and transport in situations where it is both expected and unexpected, and showing how it is affected by equilibrium magnetic field scale lengths. Ballooning parity ion temperature gradient (ITG) turbulence is found to produce magnetic stochasticity and electron thermal transport through nonlinear excitation of linearly stable tearing parity modes. The process is governed by nonlinear three-wave coupling between the ITG mode, the zonal flow, and the damped tearing parity mode. A significant electron thermal flux scales as beta squared, consistent with magnetic flutter. Above a critical beta known as the nonzonal transition, the magnetic fluctuations disable zonal flows by allowing electron streaming that effectively shorts zonal potential between flux surfaces. This leads to a regime of very high transport levels. A consideration of the residual flow in the presence of magnetic flutter confirms the disabling effect on zonal flows. Tearing parity microtearing modes become unstable in the magnetic geometry of spherical tokamaks and the RFP. They yield a growth rate in NSTX that requires finite collisionality, large beta, and is favored by increasing magnetic shear and decreasing safety factor. In the RFP, a new branch of microtearing with finite growth rate at vanishing collisionality is shown from analytic theory to require the electron grad-B/curvature drift resonance. However, when experimental MST RFP discharges are modeled gyrokinetically, the turbulence is remarkably electrostatic, showing trapped electron mode turbulence, large zonal flows, and a large Dimits shift. Analysis of the effect of the RFP’s shorter equilibrium magnetic field scale lengths shows that it increases the gradient thresholds for instability of trapped electron modes, ITG and microtearing. The stronger magnetic shear increases the beta threshold for kinetic ballooning mode (KBM) instability. This in turn increases the thresholds for magnetic activity, including the nonzonal transition.
        Speaker: Mr Paul Terry (USA)
      • 14:00
        Overview of HL-2A Recent Experiments 4h 45m
        Since the last IAEA FEC, experiments on HL-2A tokamak have been dedicated to address the physics on L-H transition, energetic-particles (EPs) and shear Alfvén waves (SAWs), ELM mitigation, disruption mitigation, edge impurity transport and other MHD related activities. In particular, significant progresses have been made in the following areas: (i) For the first time in experiments, it was found that the phase between normalized radial electric field and the envelope of density fluctuations reverses during the intermediate phase (I-phase) in comparison to the usual predator-prey regime as the plasma approaches H-mode during the L-I-H confinement transition; (ii) The frequency up- and down-sweeping reverse shear Alfvén eigenmodes (RSAEs) were observed in NBI plasmas with qmin~1. By using kinetic AE code (KAEC) simulation, it has been confirmed that the down-sweeping modes are kinetic RSAEs, and the up-sweeping modes are RSAEs that exist in the ideal or kinetic MHD limit; (iii) The transition and interaction among low-frequency MHD modes have been observed during NBI, which suggests profound interaction existing among fishbone mode, long-live mode (LLM) and tearing mode (TM); (iv) With SMBI a runaway electron plateau was observed at a rather low toroidal magnetic field. In addition, progresses have also been made in the analysis of the loss of energetic-ions, the ELM mitigation induced by supersonic molecular beam injection (SMBI), and the onset of neoclassical tearing mode (NTM) during nonlocal effect with SMBI, etc. All these experiments benefited from several newly installed diagnostics, such as Motional Stark Effect (MSE), Charge Exchange Recombination Spectroscopy (CXRS) and a scintillator-based lost fast-ion probe (SLIP), and the upgrade of ECRH heating power to 5 MW.
        Speaker: Mr Min Xu (China)
      • 14:00
        Overview of KSTAR Results in 2013 Campaign 4h 45m
        Since the initial long-pulse H-mode operation in 2012, the H-mode has been sustained longer and the operational regime of plasma parameters has been significantly extended in KSTAR tokamak. The progress in long-pulse operation is mainly due both to the increased NBI heating power of PNBI ~ 3.5 MW and the advance in the shaping control which is not trivial with slow superconducting coils. In 2013 campaign, the duration of H-mode phase has been extended up to 25 sec with 0.5 MA of plasma current and 3 MW of PNBI and, in the coming campaign as main operational goal, it is expected to be extended up to more than 30 sec using 5 MW of PNBI. In addition, in 2014 campaign, the long-pulse operation will be in accordance with ITER requirement, i.e., in ITER similar shape, low safety factor and normalized beta (~2.0). ELM suppression is discovered in wide range of coil configuration and the suppression window in the safety factor q95 has extended from 6.5 to 3.9 depending on the configuration, i.e., q95~6.5 for n=1, q95~5.0 for the mixed n=1 & n=2, and q95~4.0 for n=2 indicating the strong impact of resonant component on ELM suppression. Significant progress has been on the investigation of the underlying mechanism on RMP suppression using measurements of pedestal fluctuations and modeling of plasma response especially for n=1 case where field penetration is global and full response modeling is required including the sielding effect of toroidal rotation. Since the initial 2 segment measurements in 2012, detailed evaluation of error field (EF) has been performed by 4 segment compass scan by measuring maximum current in middle internal coils for each quadrant. In agreement with the previous measurements, the measured level of intrinsic error field is at order of 10-5 at the magnetic axis, which is an order of magnitude lower than other tokamaks. Strong focus is on the extended identification of detailed pattern of n=1 error field utilizing the full poloidal sets of internal coils and its impact on the operational boundary is investigated especially for q95 range below three where effect of error field is critical due to the locked mode and MHD activities. Including above topics, the presentation will address the recent results on rotation & transport physics, newly installed diagnostics, MHD activities and the future plan.
        Speaker: Mr Si-Woo Yoon (Korea, Republic of)
      • 14:00
        Overview of MAST Results 4h 45m
        n=2,3,4,6 has been demonstrated: at higher and lower collisionality; for the first ELM; during the current ramp-up; when a sub-set of in-vessel coils fail; and with rotating n=3 RMPs. n=4,6 fields cause less braking whilst the power to access H-mode is less with n=4 than n=3,6. Refuelling with gas or pellets gives plasmas with mitigated ELMs and reduced peak heat flux whilst less than 10% drop in stored energy. The 3d structure of the post-pellet plasmoid has been imaged, with increased fluctuations during pellet ablation. A synergy exists between pellet-fuelling and RMPs, since mitigated ELMs remove fewer particles. JOREK and CAS3D stability codes show that 3d deformations influence peeling-ballooning stability. ELM precursors strikingly observed with Doppler-backscattering (DBS) and beam emission spectroscopy (BES) are consistent with gyrokinetic simulations of micro-tearing modes (MTMs) in the pedestal. Global gyrokinetic runs show kinetic ballooning modes mediate the pedestal width, whilst nonlinear simulations suggest that MTMs carry significant electron heat flux. A scan in beta at the L-H transition shows that pedestal height scales strongly with core pressure. The observed tilt of low-k turbulent vortices increases with flow shear, due to a decrease in poloidal wave number. Fast ion redistribution by fast particle modes depends on density, and access to a quiescent domain with ‘classical’ fast ion transport is found above a critical density. Highly efficient electron Bernstein wave (EBW) current drive (1A/W) has been achieved in solenoid-free start-up. A new proton detector has characterised escaping fusion products. Langmuir probes and a high-speed camera suggest filaments play a role in particle transport in the private flux region whilst coherence imaging has measured scrape-off layer (SOL) flows. BOUT++ simulations show that fluxes due to filaments are strongly dependent on resistivity and magnetic geometry of the SOL, with higher radial fluxes at higher resistivity. MAST Upgrade is due to operate in 2015 to support ITER preparation and importantly to operate with a Super-X divertor to test extended leg concepts for particle and power exhaust. This work was part-funded by the RCUK Energy Programme and the EU Horizon 2020 programme.
        Speaker: Mr Ian Chapman (UK)
      • 14:00
        Overview of Recent Physics Results from NSTX 4h 45m
        NSTX is currently being upgraded to operate at twice the toroidal field and plasma current (up to 1 T and 2 MA), with a second, more tangentially aimed neutral beam for current and rotation control, allowing for pulse lengths up to 5 s. Recent NSTX physics analyses have addressed topics that will allow NSTX-U to achieve the research goals critical to a Fusion Nuclear Science Facility. These goals include accessing low collisionality and high beta, producing stable, 100% non-inductive operation and assessing Plasma Material Interface (PMI) solutions to handle the high heat loads expected in the next-step devices. Including rotation and kinetic resonances, which depend on collisionality, is necessary for predicting experimental stability thresholds of fast growing Ideal Wall and Resistive Wall Modes. Non-linear gyrokinetic simulations have been performed to study transport of heat, particles and momentum in the core plasma, and its dependence on collisionality and profile shapes. These studies include coupling between low- and high-k turbulence, the effect of rotation and non-local transport. PMI studies have focused on the effect of ELMs and 3D fields on plasma detachment and heat flux handling. DEGAS-2 has been used to study the dependence of gas penetration on SOL temperatures and densities for the MGI system being implemented on the Upgrade. Studies of lithium evaporation on graphite surfaces indicate that lithium increases oxygen surface concentrations on graphite, and deuterium-oxygen affinity, which increases deuterium pumping and reduces recycling. Source studies showed that the low lithium level observed in the core of lithium-coated wall NSTX plasmas was due to both high retention of lithium in the divertor as well as large neoclassical diffusivity. Noninductive operation and current profile control in NSTX-U will be facilitated by Coaxial Helicity Injection as well as RF and NB heating. CHI studies using NIMROD indicate that the reconnection process is consistent with the 2D Sweet-Parker theory. Full wave AORSA simulations show that RF power losses in the SOL increase significantly for both NSTX and NSTX-U when the launched waves propagate in the SOL. TAE avalanches can affect NB driven current through energy loss and redistribution of fast ions. Upgrade construction is moving on schedule with first operation of NSTX-U planned for Autumn 2014.
        Speaker: Mr Stanley Kaye (USA)
      • 14:00
        Overview of Results from the MST Reversed Field Pinch Experiment 4h 45m
        This overview of results from the MST reversed field pinch program summarizes physics important for the advancement of the RFP as well as for improved understanding of toroidal magnetic confinement in general. Topics include energetic particle effects, 3D helical equilibria, beta and density limit studies, microturbulence, ion heating, and magnetic self-organization physics. With neutral beam injection, several bursty energetic particle (EP) modes are observed. The profiles of the magnetic and density fluctuations associated with these EP-modes are measured using an FIR interferometer-polarimeter. Equilibrium reconstructions of the quasi-single-helicity 3D helical state are provided by the V3FIT code that now incorporates several of MST’s advanced diagnostics. A predator-prey theoretical model based on sheared flow and/or magnetic field has been developed that captures key QSH dynamics. Upgraded pellet injection permits study of density and beta limits over MST’s full range of operation, and an MST-record line-average density of 0.9E20 / m^3 (n/n_G =1.4) has been obtained. Plasma beta exhibits saturation at beta_tot ≤ 20% for a wide range of density, 0.2 < n/n_G < 1.6. Gyrokinetic simulations (GENE) based on experimental toroidal equilibrium reconstructions predict unstable trapped electron modes. Nonlinear simulations show that the “Dimits shift” is large and persists at finite beta. Experimentally, small-scale density fluctuations are detected in improved confinement plasmas. Impurity ion temperature measurements reveal a charge-to-mass-ratio dependence in the rapid heating that occurs during a sawtooth crash. Also, a toroidal asymmetry in the ion temperature is measured, correlated with 3D magnetic structure associated with tearing modes. Magnetic self-organization studies include measurements and modeling of the dynamo emf in standard RFP operation as well as with an applied ac inductive electric field to investigate the dynamics of oscillating field current drive (OFCD). Extended MHD computation for standard RFP conditions using NIMROD predicts dynamical coupling of current and plasma flow relaxation. The dynamo emf has also been measured when OFCD is applied, strengthening the understanding of and possibility for steady-state current sustainment using inductive current drive.
        Speaker: Dr Brett Chapman (USA)
      • 14:00
        Overview of the FTU Results 4h 45m
        Since the 2012 IAEA-FEC Conference, FTU operations have been largely devoted to runaway electrons (RE) generation and control, to the exploitation of the 140 GHz EC system and to liquid metal limiter elements. Experiments on RE have shown that the measured threshold electric field is larger than predicted by collisional theory and can be justified considering synchrotron radiation losses. A new RE control algorithm was developed and tested in presence of a RE current plateau, allowing to minimize the interactions with plasma-facing components and safely shut down the discharges. The experimental sessions with 140 GHz EC system have been mainly devoted to experiments on real time control of MHD instabilities using the new EC launcher with fast steering capability. Experiments with EC power modulation have confirmed the possibility to lock the sawtooth period to the EC period, with EC injection inside the q=1 surface, while experiments with central EC injection have shown the onset of 3/2 and 2/1 modes. EC assisted breakdown experiments have been focussed on ITER start-up issues, exploring the polarization conversion at reflection from inner wall and the capability to assure plasma start-up even in presence of a large stray magnetic field. A new actively Cooled Lithium Limiter (CLL) has been installed and tested. The CLL was inserted close to the last closed magnetic surface, without any damage to the limiter surface, and first elongated FTU plasmas with EC additional heating were obtained with the new CLL. Reciprocating Langmuir probes were used to measure the heat flux e-folding length in the scrape-off layer, with the plasma kept to lay on the internal limiter to resemble the ITER start-up phase. Density peaking and controlled MHD activity driven by Neon injection were investigated at different plasma parameters, and a full real-time algorithm for disruption prediction, based on MHD activity signals from Mirnov coils, was developed exploiting a large database of disruptions. New diagnostics were successfully installed and tested, as a gamma camera for RE studies and a diamond probe to detect Cherenkov radiation produced by fast electrons. Laser Induced Breakdown Spectroscopy measurements were performed under vacuum, so demonstrating the possibility to provide useful information on the fuel retention in present and future tokamaks, such as ITER.
        Speaker: Mr Gianluca Pucella (Italy)
      • 14:00
        Overview of the JET Results 4h 45m
        The European fusion programme is moving into the phase of implementation of its Roadmap. In this context, the JET programme has focused on consolidation of ITER design choices and preparation for ITER operation, with a specific emphasis given to the Bulk Tungsten Melt Experiment that has been crucial for the final decision on the material choice for the day-one tungsten divertor in ITER. In the first JET campaigns with the ITER-like Wall (ILW) fuel retention and material migration studies were a high priority. Now the focus of JET experiments has shifted towards integrated scenario development with the goal of addressing issues such as plasma-facing component (PFC) heat loads and W impurity accumulation in conjunction with high performance. In particular, during the last year, the importance of the magnetic geometry in the divertor area, strike point location and divertor pumping were established as key aspects for achieving good H-mode confinement, in combination with avoiding tungsten accumulation using ICRH. Moreover, significant effort was devoted to the use of impurity seeding to produce core-divertor compatible reference scenarios at good confinement which are essential for ITER, as well as high radiative scenarios which are required for DEMO. ITER-relevant conditions for steady-state operation have been achieved for over 7s at 2.5MA/2.7T and 21MW input power with H98(y,2)=0.85 and low divertor target power loads and partial detachment between ELMs. In parallel, post-mortem analyses of the PFCs retrieved from the first ILW campaigns have confirmed the previously reported low fuel retention obtained by gas balance. These studies show that the reduced material erosion and migration leads to reduced trapping of fuel in deposited Be layers which have less incorporated fuel in comparison with Carbon layers. In addition, the pattern of deposition within the divertor has changed significantly with the ILW in comparison with JET carbon wall campaigns due to the much-reduced level of chemical erosion. Transport to remote areas is almost absent, with the only significant Be deposits (15µm) found on the apron of the inner divertor. This work was supported by EURATOM and carried out within the framework of the European Fusion Development Agreement. The views and opinions expressed herein do not necessarily reflect those of the European Commission.
        Speaker: Mr Francesco Romanelli (European Commission)
      • 14:00
        Overview of the Recent Research on the J-TEXT Tokamak 4h 45m
        The experimental research over last two years on the J-TEXT tokamak is summarized, the most significant results including observation of core magnetic and density perturbations associated with sawtooth events and tearing instabilities by a high-performance polarimeter-interferometer (POLARIS), investigation of a rotating helical magnetic field perturbation on tearing modes, studies of resonant magnetic perturbations (RMP) on non-local transport, plasma flows and fluctuations. The POLARIS system has time response up to 1 μs, phase resolution < 0.1o and spatial resolution ~3 cm. Temporal evolution of the safety factor profile, current density profile and electron density profile are obtained during sawtooth crash events and tearing instabilities as well as disruptions. The effects of RMPs in Ohmic plasmas are directly observed by polarimeter for the first time. Particle transport due to the sawtooth crashes is analyzed. Recovery between crashes implies an inward pinch velocity extending to the center. The J-TEXT RMP system can generate a rotating helical field perturbation with a rotation frequency up to 10 kHz, and dominant resonant modes of m/n = 2/1, 3/1 or 1/1. It is found that tearing modes can be easily locked and then rotate together with a rotating RMP. During the mode locking and unlocking, instead of amplifying the island, the RMP can suppress the island width. Further numerical studies extend the understanding of the experimental observations. The effects of RMPs on plasma flows and fluctuations are studied. Both toroidal rotation velocity and radial electric field increase with RMP coil current when the RMP current is no more than 5kA. When the RMP current reaches 6kA, the toroidal velocity profile becomes flatten. Both LFZF and GAM are also damped by strong RMPs. The effects of RMPs on non-local transport in J-TEXT have been studied by using horizontal SMBI and a static RMP. At relatively low density, nonlocal phenomena are easily achieved with SMBI injection, while the rotation response to the SMBI injection is reversed for strong RMPs. SMBI without an RMP makes a change of toroidal rotation in the counter-current direction. We discovered a turbulent acceleration term for parallel rotation which has different physics from the residual stress, and is thus a new candidate mechanism for the origin of spontaneous rotation.
        Speaker: Mr Ge Zhuang (Huazhong University of Science and Technology)
        summary
      • 14:00
        Overview of the RFX-Mod Contribution to the International Fusion Science Program 4h 45m
        The RFX-mod device can be operated both as a Reversed Field Pinch (RFP), where advanced regimes featuring helical shape develop, and as a Tokamak. Due to its flexibility, RFX-mod is contributing to the solution of key issues in the roadmap to ITER and DEMO, including 3D nonlinear MHD modeling, MHD instability control, internal transport barriers, edge transport and turbulence, isotopic effect, high density limit. In RFP configuration, in the last two years advancements in the understanding of the self-organized helical states, featuring strong electron transport barriers, have been achieved; the role of microtearing modes in driving the residual transport at the barrier has been investigated experimentally and by gyrokinetic simulations. First experiments with deuterium as filling gas showed increased temperature and confinement time. New results on fast ion confinement and on the isotope effect on edge transport and turbulence are reported. RFX-mod contributed to the general issue of the high density limit physics, showing that in the RFP the limit is related to a toroidal particle accumulation due to the onset of a convective cell. The experimental program was accompanied by substantial progress in the theoretical activity: 3D nonlinear visco-resistive MHD and non-local transport modelling have been advanced; resistive wall and fast particle modes have been studied by a toroidal MHD kinetic hybrid stability code. In Tokamak configuration, q(a) regimes down to q(a)=1.2 have been pionereed, with (2,1) Tearing Mode (TM) mitigated and (2,1) RWM stabilized: the control of such modes can be obtained both by poloidal and radial sensors with proper control algorithm. Progress has been made in the avoidance of disruptions due to the locking of the (2,1) TM. External 3D fields have been applied to study a variety of physical issues: effect of magnetic perturbations on sawthooth control, plasma flow, runaway electron decorrelation. Probes combining electrostatic and magnetic measurements have been inserted to characterize turbulence and flow pattern at the edge.
        Speaker: Ms Maria Ester Puiatti (Italy)
      • 14:00
        Overview of Transport and MHD Stability Study and Impact of Magnetic Field Topology in the Large Helical Device 4h 45m
        The progress of physics understanding and concurrent parameter extension since the last IAEA-FEC 2012 [1] in the Large Helical Device is overviewed. Recently the plasma with high ion and electron temperature (Ti(0) ~ Te(0) ~ 6keV) is obtained by the combination of 1) a reduction of wall recycling and neutrals by Helium ICRF discharges and 2) optimization of carbon pellet injection and on-axis ECH. The temperature regime obtained is significantly extended. The stochastic magnetic field in the plasma core which causes flattening of the temperature and the plasma flow can be eliminated by the control of the magnetic shear by NBCD and ECCD and a high central ion temperature (Ti(0) ~ 8keV) discharge is achieved by overcoming the core temperature flattening frequently observed in the plasma with an ion- ITB. After the formation of the ion-ITB, the residual stress switches the sign from the counter- to the co-direction and results in a large toroidal flow in the co-direction [2]. On the other hand, the radial convective velocity of the carbon impurity (Vc) also changes sign from inward to outward and this reversal of convection causes the extremely hollow impurity profile (called impurity hole). A stochastization of the magnetic field affects the MHD instability driven by a pressure gradient. This is an interesting topic because the stochastization of the magnetic field is also a key issue in the resonant magnetic perturbation (RMP) experiment for edge localized mode (ELM) suppression. When the stochastization of the magnetic field is enhanced by a RMP, the pressure driven mode is suppressed even without a change in the pressure gradient itself. In LHD, a low mode (n/m=1/1) magnetic island exists near the plasma periphery and the width of the magnetic island can be controlled by the RMP. By injecting hydrogen pellets into the O-point of the magnetic island, a significant peaked pressure profile inside the magnetic island is produced for a relatively long time, which is similar to the phenomena of a “snake” in a tokamak. Inside the LCFS, the stochastization causes the damping of flows, while it enhances the E×B flow due to the electron loss to the wall along the magnetic field. [1] O.Kaneko, et. al., Nucl. Fusion 53 (2013) 104015. [2] K.Ida, et. al., Phys. Rev. Lett. 111 (2013) 055001.
        Speaker: Mr Katsumi Ida (Japan)
      • 14:00
        Progress with the ITER Project Activity in Russia 4h 45m
        Russian obligations in the ITER project consist of the development, manufacture, installation and putting into operation at the ITER site of 25 systems. At this stage Institution «Project center ITER» has signed with the ITER Organization 18 procurement arrangements for manufacture and supply of the equipment for ITER. Manufacture of the corresponding systems and development of other 7 systems is carrying out according the schedule of ITER construction. Nine unit lengths of Nb3Sn toroidal field conductor and nine unit lengths of NbTi poloidal field cables were supplied to European Union in 2013-2014 according to schedule. “Vezuvi-11M” gyrotron was tested in bench in NRC ”Kurchatov institute” where all required by ITER parameters (frequency - 170 GHz, power – 1 MW, duration - 1000 sec and efficiency - 52%) were reached in combination. 12 European and 6 Japanese mockups of divertor were tested in Efremov institute at thermal loads of 20 MW/m2. Technology of Russian Beryllium manufacture for ITER first wall was developed in collaboration between Bochvar institute (technology developer) and JSC “Bazalt” with participation of Efremov institute. It was proposed to install gamma-spectrometer behind the NPA at the same line of sight as NPA. Such combination of NPA and gamma-spectrometer will provide additional possibilities to increase accuracy of D/T ratio measurements and fast ion behaviour studies. In collaboration with Institution “Project center ITER” the NPA complex was added by diamond spectrometer of fast atoms that also will increase possibilities of fast ions studies. Results of analysis demonstrate essential increase of high field side reflectometry by adding low field side antenna to provide refractometry measurements. The prototypes of monocrystal Molybdenum mirror were manufactured for hydrogen line and charge exchange recombination spectroscopy. The prototypes of U-235 and U-238 fission chambers and compact diamond neutron spectrometers were developed and tested. The prototypes of spectrometers for CXRS are manufactured and tested.
        Speaker: Mr anatoli Krasilnikov (Russian Federation)
      • 14:00
        Recent ASDEX Upgrade Research in Support of ITER and DEMO 4h 45m
        Research on ASDEX Upgrade is programmatically focused on resolving physics questions that are key to the successful operation of ITER as well as informing design choices for a future DEMO reactor. From 2014 on, a significant part of the ASDEX Upgrade programme is run under the EUROfusion MST programme. Using the flexible set of in-vessel helical perturbation coils, penetration of 3-d fields into the plasma is studied by analysing its impact on the edge plasma in L-mode discharges or examining the interaction with rotating core MHD modes. ELM mitigation by 3-d fields at high collisionality allows access to very small, high frequency ELMs at edge temperatures and pressure gradients higher than in the usual small ELM regime at high density. This cannot be described by the usual peeling-ballooning model for ELM stability which, in general, is shown to be inadequate to explain the whole variety of type I ELM observations. The study of the impact of a full metal first wall on plasma operation and performance continues to be a major research topic. Comparison with JET has confirmed many of our results. Characterisation of the ITER Q=10 scenario (standard H-mode at q95 = 3 and betaN = 1.8) revealed that with full metal wall, confinement quality is marginal for reaching H = 1. In this scenario, type I ELMs are large and first attempts using helical fields or pellets showed limited success in mitigation. With higher betaN, confinement regularly exceeds H = 1. These findings suggest that the optimum operational point for ITER Q = 10 might be at higher q95 and betaN, evolving towards the ‘improved H-mode’ regime under study in ASDEX Upgrade. Concerning research for DEMO, a major emphasis is put on the exhaust problem. Record values of Psep/R exceeding 7 MW/m (total power up to 23 MW) with simultaneous time averaged peak target power load < 5 MW/m2 have been demonstrated under partially detached conditions using feedback controlled N-seeding. Complete detachment at high input power P > 10 MW was achieved and extensive modelling helps to clarify the mechanisms behind the experimentally observed high density zone in the divertor plasma as well as the stable X-point radiation. Using double feedback on N and Ar impurity seeding, a value of Prad,core/Ptot approx. 70 % as will be needed for DEMO was demonstrated at very good plasma performance. Finally, we will report on operational experience with two rings of tiles consisting of ferritic steel on the high field side heat shield to analyse the implications of a possible use of bare EUROFER wall panels on DEMO and the use of newly installed rings of solid W tiles in the outer divertor strike zones. All these studies are accompanied by progress in the understanding of the underlying fusion plasma physics, which is essential to obtain true predictive capability, and will be discussed in the contribution. This project has received funding from the EURATOM research and training programme 2014-2018.
        Speaker: Mr Hartmut Zohm (Germany)
      • 14:00
        Review of Globus-M Spherical Tokamak Results 4h 45m
        First experiments on noninductive current drive (CD) using lower hybrid waves at 2.45 GHz are described. Waves were launched by a 10 waveguide grill with 120° phase shift between neighboring waveguides. The experimental conditions for a poloidal slowing-down scheme are described. The CD efficiency is found to be somewhat less than for standard tokamak lower hybrid CD. Geodesic acoustic modes (GAM) have been discovered in Globus-M with the help of two Doppler reflectometers. GAMs are localized 2-3 cm inside the separatrix. The GAM frequency agrees with theory whereas the density oscillations is found to have mainly an n=0, m=0 spatial structure. Fast particle confinement during neutral beam injection has been studied and numerically simulated. Alfven instabilities excited by fast particles were detected by a toroidal Mirnov probe array. Their excitation conditions are discussed and the dynamics of fast ion losses induced by Alfven eigenmodes is presented. Unlike for conventional tokamaks, no isotopic effect on confinement has been observed in ohmic discharges comparing similar D and H plasmas. Plasma transport was modelled self consistently from the magnetic axis to the material wall. SOL parameters are compared with experimental results. For plasma-wall interaction studies a specific divertor target consisting of an a-priory damaged tungsten tiles was prepared. The damage was induced by an electron beam or by a plasma gun jet. The damage factor was equivalent to the damage produced by 100 -1000 ELM events in ITER. The first results show a strongly nonuniform temperature field formation on the damaged targets after plasma disruptions. A preliminary explanation is that the initial damage gives rise to a layer with low thermal conductivity right under the surface. We will finish explaining specific engineering design issues of the next step - Globus-M2 (1 T, 500 MA) and describe the status of component manufacture.
        Speaker: Mr Vasily Gusev (Russian Federation)
        summary
      • 14:00
        Status of JT-60SA Project 4h 45m
        In 2009, after a complex start-up phase due to the necessity to carry out a re-baselining effort to fit in the original budget while aiming to retain the machine mission, performance, and experimental flexibility, detailed design of the project was begun. In 2012, with the majority of time-critical industrial contracts in place, it was possible to establish a credible time plan, and now the project is progressing towards the first plasma in March 2019. After focussed R&D and qualification tests, the procurement of the major components and plant are now well underway. In the meantime the disassembly of the JT-60U machine has been completed in 2012.The assembly of JT-60SA started in January 2013 with the installation of the cryostat base, the first item delivered from Europe, and continued in February 2014 with the installation of the three lower superconducting equilibrium field coils. Winding of the TF coils winding packs started in July 2013 in EU. A cold test facility for the TF coils has been installed at the CEA Centre in Saclay. The first TF coil will start tests in 2015. The manufacture of the EF4, EF5 and EF6 coils has been completed. All Vacuum Vessel sectors will be completed by April 2014 and their assembly will soon commence. The manufacturing of the cryostat vessel body has now begun, with final delivery planned in 2017. Contracts for all Magnet Power Supplies are placed, fabrication ongoing with the one for the Quench Protection System completed and their installation to commence in 2014. On-site installation of the cryogenic system will be completed in 2015, and it will be operational late in 2016. The dual frequency 110 and 138 GHz gyrotron has made significant progress towards allowing EC heating (ECH) and current drive (ECCD) under a wide range of plasma parameters. Oscillations of 1 MW for 10 s were successful at both frequencies in a world first for a dual-frequency gyrotron by optimizing electron pitch factor using a triode electron gun. On the N-NB system, the pulse duration and the current density of the negative ion source have been successfully improved from 30 s at 80 A/m2 in the previous operation to 100 s at 120-130 A/m2. The paper will give an overview of the present status of the engineering design, manufacturing and assembly of the JT-60SA machine.
        Speaker: Mr pietro barabaschi (f4e)
      • 14:00
        The ITER Project Construction Status 4h 45m
        The ITER Project has visibly made its transition to the construction phase in the two years since the last Fusion Energy Conference in San Diego. By mid-February 2014 commitments to in-kind procurement are approaching 89.6 percentage of the total credit value and 70.7 percentage (99 out of 140) in the number of Procurement Arrangements. Construction is accelerating and the appearance of the site is changing on a daily basis. Vigorous efforts are underway to mitigate some remaining organizational problems, including the alignment of the ITER Organization (IO) and 7 Domestic Agencies (DAs), minimizing any possible delay factor. The seven Parties are well committed to the construction of ITER. The ITER project has gone beyond the turning point. This fact should be understood and shared. We only can go forward together. The total progress being made is so enormous that it is impossible to do justice to everything in a single presentation. Therefore, the buildings, the core tokamak and some of the balance of plant will be reported on here. Most ancillary systems ranging from heating and current drive systems, diagnostics and fuelling systems to remote handling and the hot cell facility are in an advanced state of design. Their status will have to be reported at another time. The same is true of both the integration and assembly efforts.
        Speaker: Mr Osamu Motojima (ITER Organization)
      • 14:00
        The Science Program of the TCV Tokamak: Exploring Fusion Reactor and Power Plant Concepts 4h 45m
        TCV is acquiring a new 1 MW neutral beam and 2 MW additional third-harmonic ECRH to expand its operational range. Its existing shaping and ECRH launching versatility was amply exploited in an eclectic 2013 campaign. A new high-confinement mode (IN-mode) was found with an edge barrier in density but not in temperature. Density limits close to the Greenwald value were reached – at reduced confinement - by sawtooth regularization with ECRH. The edge gradients were found to be regime-dependent and to govern the scaling of confinement with current. A new theory predicting a toroidal rotation component at the plasma edge, driven by inhomogeneous transport and geodesic curvature, was tested with promising results. The L-H threshold power was measured to be 15-20% higher in both H and He than D, to increase with the length of the outer separatrix, and to be independent of the current ramp rate. Core turbulence was found to decrease from positive to negative edge triangularity deep into the core, consistent with global confinement increase. The geodesic-acoustic mode was studied with multiple diagnostics, and its axisymmetry was confirmed by a full toroidal mapping of its magnetic component. The heat flux profile decay length and heat load profile on the wall were documented as functions of plasma shape in limited plasmas. In the snowflake (SF) divertor configuration we have documented the heat flux profiles on all four strike points. SF simulations with the 3D EMC3-Eirene code, including the physics of the secondary separatrix, underestimate the flux to the secondary strike points, possibly resulting from steady-state ExB drifts. With neon injection, radiation in a SF was 15% higher than in a conventional divertor. The novel triple-X and X-divertor configurations were achieved transiently in TCV. A new sub-ms real-time equilibrium reconstruction code was used in ECRH control of NTMs and in a prototype shape controller. The detection of visible light from the plasma boundary was also successfully used in a position-control algorithm. A new bang-bang controller improved stability against vertical displacements. The RAPTOR real-time transport simulator was applied to current density profile control experiments with ECCD. Shot-by-shot internal inductance optimization was demonstrated by iterative learning control of the current reference trace.
        Speaker: Mr Stefano Coda (Switzerland)
      • 14:00
        Transport, Stability and Plasma Control Studies in the TJ-II Stellarator 4h 45m
        Recent improvements in TJ-II plasma diagnostics and operation have led to a better understanding of transport, stability and plasma control in fusion plasmas. Impurity transport: Observations of asymmetries in impurity parallel flows in TJ-II ion-root plasmas have been interpreted as an indication of the compressible variation of the impurity flow field and hence of in-surface impurity density asymmetries. In addition, first-time observations of electrostatic potential variations within the same magnetic flux surfaces are presented which are reproduced by neoclassical Montecarlo calculations. The dependence of impurity confinement time has been also studied as a function of charge and mass. Momentum transport and isotope physics: TJ-II has provided evidence that three-dimensional magnetic structures convey significant impact on plasma confinement and L-H transitions. Recent observations on the temporal ordering of the limit cycle oscillations at the L-I-H transition show the leading role of the plasma turbulence. Comparative studies in tokamaks and stellarators have provided direct experimental evidence of the importance of multi-scale physics for unravelling the physics of the isotope effect on transport. Power exhaust physics: Novel solutions for plasma facing components based on the use of liquid metals like Li and alloys have been developed on TJ-II. The TJ-II programme on liquid metals addresses fundamental issues like the self-screening effect of liquid lithium driven by evaporation to protect plasma-facing components against heat loads and tritium inventory control, using Li-liquid limiters (LLL) recently installed. Plasma stability studies: Experiments with magnetic well scan on TJ-II suggest that stability calculations, as those presently used in the optimization criteria of stellarators, might miss some stabilization mechanisms. Fast particle control: The TJ-II results show that, upon moderate off-axis ECH power application, the continuous character of the Alfvén eigenmode (AEs) changes significantly and starts displaying frequency chirping. This result shows that ECH can be a tool for AE control that, if confirmed, could become ITER and reactor-relevant.
        Speaker: Mr Joaquín Sánchez (Spain)
    • 16:10 16:40
      Coffee Break 30m
    • 16:40 18:45
      ITER Technology: FIP/1 Blue 1-5

      Blue 1-5

      Hotel Park Inn Pribaltiyskaya

      Saint Petersburg, Russian Federation
      Convener: Mr Oleg Filatov (Russian Federation)
      • 16:40
        Development of Tungsten Monoblock Technology for ITER Full-Tungsten Divertor in Japan 20m
        Through R&D for a plasma facing unit (PFU) of a full-tungsten (W) ITER divertor, Japan Atomic Energy Agency (JAEA) succeeded in demonstrating the durability of the W divertor which endured a repetitive heat load of 20 MW per sqare meter without macroscopic cracks of all W armors. At the beginning of this activity, the bonding technology armor to heat sink was one of the most important key issues in a manufacturing process. JAEA improved the bonding process of the W divertor mock-ups. At first the bonding between the W armor and the copper interlayer (Cu) is performed by using several technologies, such as "Direct casting of Cu" or "Diffusion bonding" or "HIP bonding". Then the brazing between the Cu and the cooling pipe is done. Then the rejection rate due to those bonding processes has been significantly been reduced. As a performance test for the bonding and a heat removal capability, the high heat flux testing was carried out for 6 small-scale mock-ups for the R&D of the full-W ITER divertor. Moreover, a W part of 4 full-scale prototype PFUs were also tested. In the tests, all of the W monoblocks endured the repetitive heat load of 10 MW per sqare meter for 5,000 cycles and 20 MW per sqare meter for 1,000 cycles without the macroscopic crack, which strongly encourages the realization of the full-W divertor target from the start of the operation in ITER. This paper presents the latest R&D activities on the full-W ITER divertor in JAEA.
        Speaker: Mr Yohji Seki (Japan Atomic Energy Agency)
        Poster
        Slides
        summary
      • 17:00
        Surface Heat Loads on Tungsten Monoblocks in the ITER Divertor 20m
        ITER will begin operation with a full W divertor consisting of solid W monoblocks (MB) bonded to CuCrZr water-carrying cooling tubes. We seek to predict heat flux distributions on MBs comparing three models: optical calculations, ion orbit calculations, and PIC. The consequences of five design aspects are analyzed in terms of available models: (1) Divertor target tilting to shadow leading edges at inter-cassette gaps leads to increased local power flux and total power to wetted MBs, as well as to reduced margin against surface melting during fast transients, due to the increased B-field angle (the total power to each target is the same, but it is distributed among fewer MBs). (2) Within a single tilted target, chamfering foreseen to protect leading edges of the MBs against misalignments also leads to increased local power flux due to the increased B-field angle, but that is exactly compensated by a reduction of the wetted area, thus the total power to individual MBs is conserved. ANSYS thermal simulations predict additional heating of MB trailing edges due to non-uniform heat flux deposition. (3) Relative misalignments between plasma-facing units (PFU) play a critical role in the total power incident on a single MB. A MB on a PFU that is radially misaligned by 0.3 mm on a tilted target (see 1) will collect a total convected plasma power that is 50% higher than a perfectly aligned MB on an untilted target. If that power were spread uniformly over the top surface of the MB, it would be equivalent to a power flux that is 50% higher than the specified limits. There is no shaping solution that can remedy this problem. (4) Due to their large Larmor radii, ions released from the pedestal during ELMs can strike magnetically shadowed poloidal edges. The fraction of ELM parallel power flux that penetrates into gaps is independent of ion temperature over the range 0.5keV
        Speaker: Mr James Paul Gunn (CEA Cadarache)
        Slides
        summary
      • 17:20
        Full-Scale Trial Results to Qualify Optimized Manufacturing Plan for ITER Toroidal Field Coil Winding Pack in Japan 20m
        A heat-treated Nb3Sn cable-in-conduit conductor (CICC) must be inserted into a groove of a radial plate (RP), which is designed to maintain the mechanical and electrical reliability of the insulation of ITER Toroidal Field (TF) coil during its 20 years’ operation. The difference between heat-treated conductor length and RP groove length must be controlled with accuracy of +/-0.05%. JAEA developed high accuracy winding system and procedure with the order of +/-0.01% in wound conductor length and performed full-scale winding trials. The target tolerance of +/-0.01% was achieved. In addition, very complicate procedure of RP insertion between upper and lower windings (pancakes), which consist of unit length conductor, is also qualified by using full-scale dummy conductor winding and trial insertion of wound dummy conductor into RP groove was performed, too. Furthermore, proto double-pancake (DP) was successfully heat-treated. These results justify validity of optimized manufacturing plan and allow us to start TF coil winding pack (WP) manufacture.
        Speaker: Mr Norikiyo Koizumi (Japan)
        Slides
        summary
      • 17:40
        Summary of the Test Results of ITER Conductors in SULTAN & Research, Development and Production of ITER Toroidal Field Conductors and Poloidal Field Cables in Russia 20m
        A After completing the qualification tests of the ITER cable-in-conduit conductors (CICC), the series manufacture tests are running in the SULTAN test facility in Villigen, Switzerland, with target completion date in 2015. The key test for the conductor samples is the current sharing temperature, Tcs, at the nominal operating field and current, i.e. the maximum temperature at which the conductors operate before developing an electric field of 10 µV/m. All the TF samples fulfilled the ITER requirement of Tcs ≥ 5.8 K after 1000 load cycles. The Tcs results have a broad scattering among the suppliers, from 5.8 K up to 6.6 K. The assembly of the Nb3Sn based CICC samples (for TF and CS coils) is carried out at CRPP. The NbTi CICC samples (for PF, CC and bus bars) are assembled at the suppliers, with a U-bend replacing the bottom joint. The poor performance of some Main Busbar (MB) conductor samples, caused by poor sample assembly, triggered the effort to assemble a MB sample at CRPP with solder filled terminations and a bottom joint. The superior test results of the MB-CRPP sample, closely matching the performance assessment carried out using 3-D field distribution and n-index behaviour was a successful achievement of the last year of operation. According to the Procurement Arrangement for the ITER coils, the winding companies must qualify the joint and termination manufacture by SULTAN samples. The first joint sample tested in SULTAN was a TF joint from EU, followed by a Correction Coil (CC) joint sample from China. Other joint samples are being assembled in USA (Central Solenoid), in Russia (PF1) and in China (PF6). All the ITER coils use the “twin box” design for joints, except the Central Solenoid. At the first test in SULTAN of a twin-box TF joint sample in 2013, an unexpected resistance increase was observed after an accidental dump of the SULTAN field, causing a large field transient parallel to the joint contact surface, with large eddy currents and electromagnetic loads at the pressure-contact between strand bundle and copper plate of the twin box. The resistance requirement for the TF joint was still fulfilled after the dump. The impact of transient field on resistance and stability was investigated at an additional test campaign of the TF joint sample, with intentional dumps of the SULTAN field. B Russian Federation is the initiator and active participant in development and building of International Thermonuclear Experimental Reactor – ITER. The major element of ITER is its huge superconducting magnet. Special superconducting cables and conductors had to be developed to satisfy very strict demands for such conductors. A lot of Research and Development (R&D) works have been performed in Russia to create our own production of superconductors, cables and conductors. Russian Scientific R&D Cable Institute (known by Russian abbreviation as VNIIKP) has been participating in ITER project since 1993 both at the early stage of R&D and at the following stage of Engineering Design Activity (EDA). Tests of several short samples at Sultan test facility were crowned by successful testing of Toroidal Field (TF) and Poloidal Field (PF) insert coils performed in Japan in 2001 and 2008 correspondingly. After many R&D works bow VNIIKP is actively implementing the final production and delivery stage of cables and conductors. By 2009 the new and modern technological complex has been accomplished to produce PF cables for both the Russian Federation (RF) and European parts and TF conductors for RF part. The complex includes several productions such as chemical technology line, cabling facility and jacketing line. In this review we present a short history of VNIIKP participation in ITER and our current achievements, including some R&D results. The technology used and our production line are described in some details.
        Speaker: Mr Pierluigi Bruzzone (Switzerland)
        Slides
      • 18:00
        Overview of the Design Development, Prototype Manufacturing and Procurement of the ITER In-Vessel Coils 20m
        ITER is incorporating two types of In-Vessel Coils (IVCs): ELM Coils to mitigate Edge Localized Modes and VS Coils to provide Vertical Stabilization of the plasma. Strong coupling with the plasma is required in order that the ELM and VS Coils can meet their performance requirements. Accordingly, the IVCs are mounted on the Vacuum Vessel (VV) inner wall, in close proximity to the plasma, just behind the Blanket Shield Modules (BSM). Fitting the coil systems in between the BSM and the VV leads to difficult integration with diagnostics and cooling water manifolds. This location results in a radiation and temperature environment that is severe necessitating new solutions for material selection, as well as challenging thermo-mechanical analyses and design solutions. Due to high radiation environment, mineral insulated copper conductors enclosed in a steel conduit have been selected. The project is being led and managed by the ITER Organization in close collaboration with the Chinese Academy of Sciences (ASIPP) in Hefei, China, and with the Princeton Plasma Physics Laboratory (PPPL) in Princeton NJ, USA. Prototype manufacturing has been completed by ASIPP. The aim was to develop suitable manufacturing procedures and techniques necessary to fabricate the ELM and VS Coils, and to qualify electrical and mechanical test procedures to meet the acceptance criteria. An extensive set of analyses to evaluate the effects of the high temperatures and electromagnetic loads on the In-Vessel Coils has been carried out at PPPL. The design of the IVCs has been finalized, and it takes into account the results from the prototype manufacturing. The procurement of the IVCs and their conductors will be done via direct call-for-tenders from the ITER Organization and preparation has already started. It is expected that the first call-for-tender will be launched in mid-2014 and the contract signed in early 2015. This paper will give an overview of the detailed design and prototype manufacturing, procurement and schedule for the In-Vessel Coils.
        Speaker: Ms Anna Encheva (Iter Organization)
        Slides
      • 18:20
        Progress in the Design and Manufacture of High Vacuum Components for ITER & Manufacturing Design and Progress of the First Sector for ITER Vacuum Vessel 20m
        A ITER is a large experimental tokamak device being built to demonstrate the feasibility of fusion power. The main scope of this paper is to report the status of the design and manufacturing activities of two major ITER components, the ITER Vacuum Vessel (VV) and the Cryostat. Both components will provide the necessary high-vacuum required in the case of the VV for plasma operation and confinement and to allow for cooldown of the superconducting magnets to cryogenic temperature (Cryostat). The design of the two systems has been developed by the ITER Organization (IO) with the support of many R&D activities carried out by the Parties and is almost complete. Procurement Arrangements (PAs) with four Domestic Agencies (DAs) have been signed to develop the manufacturing design and manufacture the components of these systems. Some detailed design on specific components still needs to be completed. Manufacturing contracts have been placed in 2010-2012 with many preparation and qualification activities. The production of the full-scale VV sectors and cryostat sections has started in the four DAs with the procurement of base materials and manufacture of mock-ups or full-scale components. Realistic manufacturing schedules are being consolidated and the presently expected completion dates will also be reported in this paper. B The ITER Vacuum Vessel (VV) is a torus shaped double wall structure and consists of nine sectors and several ports. Main functions of the VV are to provide high vacuum for plasma operation and to protect radioactive contamination as the first safety barrier. Korea Domestic Agency (KODA) has responsibility for procuring of two sectors including the first sector which will be delivered before others. KODA contracted with Hyundai Heavy Industries Co., LTD (HHI) to product the VV sectors and major ports. The design and fabrication of the VV as nuclear equipment shall be complied with the RCC-MR code and regulations of nuclear pressure equipment in France (ESPN). The manufacturing design has been developed to fabricate the main vessel and port structures in accordance with the design requirements. All manufacturing sequences including welding methods are also established to meet the demanding tolerance and inspection requirement. The manufacturing design of Korean sectors has special design concepts to minimize welding distortion such as self-sustaining support ribs and cup-and-cone type segment joints. Several mock-ups have been constructed to verify and develop the manufacturing design and procedures. Qualifications for welding, forming and NDE have been conducted to verify related procedures according to the requirement. For fabrication of the VV sectors and ports, 3,000 tons of plates and forgings had been produced by European steel companies and delivered to HHI. Four poloidal segments (PS) for the first sector are being fabricated simultaneously in HHI factory. All inner shells were cut, bended and machined for welding. Welding and NDE of inner shells for PS2 and PS4 are finished. To reduce schedule delay machining of forging blocks are on-going in parallel. Some of machined blocks are welded on the inner shell by TIG and electron beam welding.
        Speaker: Mr Carlo Sborchia (ITER Organization)
        Slides
    • 19:30 22:00
      Reception
    • 08:30 10:15
      Overview 3: Inertial & Magnetic Fusion: OV/3 Blue 1-5

      Blue 1-5

      Hotel Park Inn Pribaltiyskaya

      Saint Petersburg, Russian Federation
      Convener: Mr Masahiro Mori (Japan)
      • 08:30
        Fast Ignition Realization EXperiment (FIREX) and Prospect to Inertial Fusion Energy in Japan 25m
        Fast ignition has high potential to ignite a fusion fuel with only about one tenth of laser energy necessary for the central ignition. One of the most advanced fast ignition programs is the Fast Ignition Realization Experiment (FIREX). The goal of its first phase is to demonstrate ignition temperature of 5 keV, followed by the second phase to demonstrate ignition-and-burn. Relativistic fast electrons as the energy carrier, however, unfavorably diverge at high laser intensities necessary for significant heating. This difficulty is overcome by kilo-Tesla magnetic field collimating fast electrons towards a compressed fuel. Such super-strong field has been created with a capacitor-coil target driven by a high power laser, and subsequent collimation has also been demonstrated, suggesting that one can achieve ignition temperature at the laser energy available in FIREX. Repetitive creation of fast ignition plasmas has been demonstrated together with the technology development of high-efficient rep lasers and pellet injection, tracking, and beam steering.
        Speaker: Mr Hiroshi AZECHI (Japan)
        Slides
      • 08:55
        Status of JT-60SA Project 25m
        In 2009, after a complex start-up phase due to the necessity to carry out a re-baselining effort to fit in the original budget while aiming to retain the machine mission, performance, and experimental flexibility, detailed design of the project was begun. In 2012, with the majority of time-critical industrial contracts in place, it was possible to establish a credible time plan, and now the project is progressing towards the first plasma in March 2019. After focussed R&D and qualification tests, the procurement of the major components and plant are now well underway. In the meantime the disassembly of the JT-60U machine has been completed in 2012.The assembly of JT-60SA started in January 2013 with the installation of the cryostat base, the first item delivered from Europe, and continued in February 2014 with the installation of the three lower superconducting equilibrium field coils. Winding of the TF coils winding packs started in July 2013 in EU. A cold test facility for the TF coils has been installed at the CEA Centre in Saclay. The first TF coil will start tests in 2015. The manufacture of the EF4, EF5 and EF6 coils has been completed. All Vacuum Vessel sectors will be completed by April 2014 and their assembly will soon commence. The manufacturing of the cryostat vessel body has now begun, with final delivery planned in 2017. Contracts for all Magnet Power Supplies are placed, fabrication ongoing with the one for the Quench Protection System completed and their installation to commence in 2014. On-site installation of the cryogenic system will be completed in 2015, and it will be operational late in 2016. The dual frequency 110 and 138 GHz gyrotron has made significant progress towards allowing EC heating (ECH) and current drive (ECCD) under a wide range of plasma parameters. Oscillations of 1 MW for 10 s were successful at both frequencies in a world first for a dual-frequency gyrotron by optimizing electron pitch factor using a triode electron gun. On the N-NB system, the pulse duration and the current density of the negative ion source have been successfully improved from 30 s at 80 A/m2 in the previous operation to 100 s at 120-130 A/m2. The paper will give an overview of the present status of the engineering design, manufacturing and assembly of the JT-60SA machine.
        Speaker: Mr Pietro Barabaschi (European Commission)
        Slides
      • 09:20
        Advance of H-Mode Physics for Long-Pulse Operation on EAST 25m
        Since the last IAEA-FEC, significant progress has been made on EAST on both physics and technology fronts towards the long-pulse operation of high-confinement plasma regimes. EAST has been upgraded with more than 25 MW of CW heating & current drive power, along with 70 diagnostics, two internal cryopumps, an ITER-like W monoblock top divertor and resonant magnetic perturbation coils, which will enable EAST to investigate long-pulse H-mode operation with dominant electron heating and low torque input, which will be facing challenges on some of critical issues on ITER. New information has been obtained on the physics of L-H transition. Remarkable efforts have been made in mitigating type-I ELMs in a stationary state H-mode plasma with multi-pulses of supersonic molecular beam injection (SMBI), LHCD, lithium granule and deuterium pellet injection, as well as RMPs, thus potentially offering a valuable means of heat-flux control for next-step long-pulse fusion devices. Long-pulse H-mode discharges with H(98,y2) ~1 have been obtained either with ELM mitigation or in a small ELMy regime accompanied by a new electrostatic edge coherent mode, which appears in the steep-gradient pedestal region and plays a dominant role in driving heat and particles outwards. High peak heat load on the divertor due to type I ELMs, is reduced either by SMBI or LHCD. We find that ELM mitigation with SMBI is due to enhanced particle transport in the pedestal, correlated with large-scale turbulence and strongly anti-correlated with small-scale turbulence, while LHCD induces edge plasma ergodization, broadening the heat deposition footprint. Challenges and progresses on plasma control, effective H&CD, plasma-wall interactions under long-pulse, high-heat flux and high-Z metal wall conditions will also be presented.
        Speaker: Mr Baonian Wan (China)
        Slides
      • 09:45
        Overview of KSTAR Results in 2013 Campaign 25m
        Since the initial long-pulse H-mode operation in 2012, the H-mode has been sustained longer and the operational regime of plasma parameters has been significantly extended in KSTAR tokamak. The progress in long-pulse operation is mainly due both to the increased NBI heating power of PNBI ~ 3.5 MW and the advance in the shaping control which is not trivial with slow superconducting coils. In 2013 campaign, the duration of H-mode phase has been extended up to 25 sec with 0.5 MA of plasma current and 3 MW of PNBI and, in the coming campaign as main operational goal, it is expected to be extended up to more than 30 sec using 5 MW of PNBI. In addition, in 2014 campaign, the long-pulse operation will be in accordance with ITER requirement, i.e., in ITER similar shape, low safety factor and normalized beta (~2.0). ELM suppression is discovered in wide range of coil configuration and the suppression window in the safety factor q95 has extended from 6.5 to 3.9 depending on the configuration, i.e., q95~6.5 for n=1, q95~5.0 for the mixed n=1 & n=2, and q95~4.0 for n=2 indicating the strong impact of resonant component on ELM suppression. Significant progress has been on the investigation of the underlying mechanism on RMP suppression using measurements of pedestal fluctuations and modeling of plasma response especially for n=1 case where field penetration is global and full response modeling is required including the sielding effect of toroidal rotation. Since the initial 2 segment measurements in 2012, detailed evaluation of error field (EF) has been performed by 4 segment compass scan by measuring maximum current in middle internal coils for each quadrant. In agreement with the previous measurements, the measured level of intrinsic error field is at order of 10-5 at the magnetic axis, which is an order of magnitude lower than other tokamaks. Strong focus is on the extended identification of detailed pattern of n=1 error field utilizing the full poloidal sets of internal coils and its impact on the operational boundary is investigated especially for q95 range below three where effect of error field is critical due to the locked mode and MHD activities. Including above topics, the presentation will address the recent results on rotation & transport physics, newly installed diagnostics, MHD activities and the future plan.
        Speaker: Mr Si-Woo Yoon (Republic of Korea)
        Slides
        summary
    • 08:30 12:30
      Poster 1: P1 Green 8-9

      Green 8-9

      Hotel Park Inn Pribaltiyskaya

      Saint Petersburg, Russian Federation
      • 08:30
        A Flute Instability under the ExB Shear Flow in an Open System 4h
        A flute mode is the most dangerous instability in an open system such as the GAMMA10 tandem mirror. So how to stabilize the flute instability is an important problem. This paper studies the stabilizing effects of the ExB shear flow on a flute instability in the Cartesian geometry by the particle simulation and in the GAMMA10 magnetic field by the reduced MHD simulation. The particle simulation uses the (2 and 1/2) dimensional electrostatic implicit code with 128x128 meshes. Ions (electrons) are distributed to have the step functional density profile in x and the ions (electrons) flow in the y-direction with the ExB drift shear as an initial condition. The gravitational acceleration force g is applied in the x-direction. It was found that the flute mode was always unstable to the step functional density profile in x, where conducting boundary condition in x and the periodic boundary condition in y were adopted. The shear flow can excite the Kelvin-Helmholtz (K-H) instability which is stable in the uniform shear flow. A flute instability was transformed into a K-H instability after linearly growing phase. Whether the system is unstable to a flute mode depends on whether the excited K-H mode has an enough power of collapsing the system. It was newly found the suppression condition of plasma collapse (or the stability condition of secondary instability) with discontinuous density step under the uniform shear flow. The flute instability under the shear flow in GAMMA10 was investigated by using the reduced MHD code. Here the centrifugal acceleration of ions resulting from a magnetic field line curvature is included in the radial dependence of magnetic field line specific volume in the code. The shear flow is inputted initially by the equilibrium dynamic plasma vorticity and the resultant electrostatic potential. The reduced MHD simulation adopts the smooth radial pressure profile as an initial condition with uniform azimuthal shear flow. It was found that the linear growth rate of a flute instability was decreased by the ExB shear flow, so that the linear stability analysis was applicable in the smooth pressure radial profile. The reduced MHD simulation found that the flute instability was stabilized by the ExB shear flow in the linear phase. So it was newly determined the stability condition of the flute instability by the ExB uniform shear flow in GAMMA10.
        Speaker: Prof. Isao Katanuma (University of Tsukuba)
      • 08:30
        Analysis of Plasma Behavior in the Localized Non-Axisymmetric B Region of the GAMMA 10 Tandem Mirror 4h
        Ion current in ion drift direction as well as the plasma shift in that direction in the localized non-axisymmetric magnetic field region of the anchor cell of the GAMMA 10 tandem mirror were measured. In addition, electron current perpendicular to the ion drift direction was found in this region. To understand these plasma behaviors, relevant kinetic properties of plasma particles are analyzed numerically. Numerical analysis indicates that the ion drift due to the non-axisymmetric magnetic field and electron displacement due to small angle Coulomb scattering during passing through drifted ions are the possible candidates of the observed plasma phenomenon.
        Speaker: Dr Md. Khairul Islam (Bangladesh Atomic Energy Commission)
        Poster
        summary
      • 08:30
        Computation of Resistive Instabilities in Tokamaks with Full Toroidal Geometry and Coupling Using DCON 4h
        Precise determination of resistive instabilities is an outstanding issue in tokamaks, remaining unsatisfactory for a long time despite its importance for advanced plasma control. This paper presents the first successful computation of such resistive instabilities including full mode coupling and multiple singular surfaces, by upgrading DCON [1] with a resonant-Galerkin method [2] using improved basis functions such as Hermite cubics and high-order Frobenius power series. Incorporating the resistive layer model of Glasser-Greene-Johnson (GGJ) [3] and matching the inner-layer solutions into full outer-layer solutions in DCON, a complete picture of resistive instabilities in tokamaks can be obtained and studied. Excellent quantitative agreement with the MARS-F code [4], for both growth rate and outer-layer solutions, has been achieved. Convergence is also a distinguished property in DCON as tested with challenging NSTX equilibria with strong shaping, high-β, and multiple rational surfaces up to 10. Another important advantage in DCON is the separation of the inner-layer from the outer-layer regions, which will allow us to extend inner-layer model efficiently to more advanced fluid equations including drift kinetic effects and to perform more precise calculations of non-ideal stability and 3D perturbed equilibria in the future. [1] A. H. Glasser and M. S. Chance, Bull. Am. Phys. Soc. 42, 1848 (1997) [2] A. Pletzer and R. L. Dewar, J. Plasma Physics 45, 427 (1991) [3] A. H. Glasser, J. M. Greene, and J. L. Johnson, Phys. Fluids 18, 7 (1975) [4] Y. Liu et al., Phys. Plasmas 19, 172509 (2012) This research was supported by U.S. DOE contracts #DE-AC02-09CH11466.
        Speaker: Dr Yueqiang Liu (UK)
      • 08:30
        Current Drive by Electron Temperature Gradient Turbulence in Tokamak Pedestal Region 4h
        In this paper, the quasilinear version of the current evolution equation in the presence of ETG turbulence in the tokamak pedestal region is written down. It has been shown that the current drive has to fight the conventional resistive dissipation mechanism as well as new dissipation mechanisms, such as a turbulence driven hyper-resistivity coefficient associated with the ETG turbulence. It is likely that the ETG turbulence tends to saturate at amplitudes much larger than what the mixing length theory would predict, primarily because of nonlinear radial streamer like mechanisms, which encourage big radial steps across the magnetic fields and give appropriate and reasonable magnitudes of the cross field transport due to this instability. We have used these saturated ETG turbulence levels to estimate the magnitudes of the spontaneous source of toroidal current injection as well as the anomalous hyper resistivity coefficient. These estimates of turbulence driven current are compared with the background bootstrap current in the pedestal region. It is concluded that significant modification of the equilibrium currents as well as current profiles may arise in the pedestal region as a consequence of the turbulent injection of current in the basic pedestal plasma.
        Speaker: Dr Sanat Kumar Tiwari (Institute for Plasma Research, Bhat, Gandhinagar, Gujarat)
        summary
      • 08:30
        Density Limit Studies in the Tokamak and the Reversed-Field Pinch 4h
        Both in the tokamak and the reversed-field pinch (RFP), new finds show that the high density limit, which often disrupts tokamak discharges and slowly terminates RFP ones, is not governed by a unique, theoretically well-determined physical phenomenon, but by a combination of complicated mechanisms involving two-fluid effects, electrostatic plasma response to magnetic islands and plasma-wall interaction. In this paper we will show that the description in terms of the unique "Greenwald density" nG = Ip/pi a^2 should be reinterpreted in terms of edge critical density, and related to the amplitude of the equilibrium magnetic field, the resonance of islands next to the edge, and input power. Recent results in FTU point out that in discharges with a variable density peaking the line-averaged central density scales as n0 ~ B^1.5, which is a scaling with the magnetic field. The usual Greenwald-like scaling nedge = 0.35 nG holds for the edge density. The density limit depends also on the input power: recent experiments in the RFX-mod RFP with a lithized wall show that the central density increases linearly with the ohmic input power and that larger densities can be accessed, for the same input power, with better wall conditioning. An important point to raise is the role of the thermal instabilities in setting the environment for the development of the density limit: in both machines, FTU and RFX, the density limit is associated with the appearance of the multifaceted asymmetric radiation from the edge (MARFE), which is triggered by MHD activity (m/n=2/1 in FTU and 0/1 in RFX). In the RFX case, the MARFE is also linked to a well-defined flow pattern. In fact, the 0/1 island, which resonates at q = 0 in the RFP edge (r/a = 0.9), is destabilized at high density, and generates an electrostatic response in the form of a convective cell, with the same 0/1 symmetry. The toroidal flow reverses direction along the toroidal angle, with the formation of two null points of v_phi (or, equivalently, radial electric field Er): a source and a stagnation point, with the latter corresponding to the toroidally localized MARFE. The association between flow patterns and MARFE can be tested in FTU, by investigating the effect of ERCH on the MARFE and the density peaking. Initial results, on FTU as well as in ASDEX, indeed show a dependence of the disruption phenomenology on the ECRH.
        Speaker: Dr Gianluca Spizzo (Consorzio RFX, Associazione EURATOM-ENEA sulla Fusione, Padova, Italy)
        Poster
        summary
      • 08:30
        Determination of the System Function for the Particle Circulation Process Using Perturbation Technique in QUEST 4h
        A new approach to realize steady state tokamak operation SSTO has been demonstrated in QUEST with all metal wall baked at 100°C. Using particle flux perturbations driven by particle source H2 and plasma-wall interaction PWI the system functions of processes of retention and release into/from the wall are determined both in time and frequency domains. The system function for the particle circulation has been determined by perturbation technique and independent measurement of partial pressure and permeation flux. The temporal evolution of system function, especially very low frequency component, must be controlled in order to sustain the steady-state discharge.
        Speaker: Dr Arseniy Kuzmin (Kyushu University, RIAM, AFRC, Japan, Kasuga)
        Poster
      • 08:30
        Developing and Validating Predictive Models for Fast Ion Relaxation in Burning Plasmas 4h
        The performance of the burning plasmas is limited by the confinement of superalfvenic fusion products, alpha particles, resonating with the Alfvénic eigenmodes (AEs). Two techniques are developed to evaluate the AE induced fast ion relaxation. Both rely on linear instability theory and are confirmed by experiments. The first is the reduced quasilinear technique or critical gradient model (CGM) where marginally unstable (or critical) gradient of fast ion pressure is due to unstable AEs. It allows the reconstruction of fast ion pressure profile and computed their losses. The second technique is called hybrid that is also based on NOVA-K linear stability computations of TAE (or RSAE) mode structures and growth rates. AE amplitudes are computed from the nonlinear theory perturbatively and used in the numerical runs. With the help of the guiding center code ORBIT the hybrid model relaxes the fast particle profiles. We apply these models for NSTX and DIII-D plasmas with the neutral beam injections for validations. Both methods are relatively fast ways to predict the fast ion profiles in burning plasmas and can be used for predictive modeling prior to building experimental devices such as ITER.
        Speaker: Prof. William W. Heidbrink (USA)
        Poster
        summary
      • 08:30
        Development of Lithium CPS Based Limiters for Realization of a Concept of Closed Lithium Circulation Loop in Tokamak 4h
        Development of commercially attractive project of a tokamak based fusion neutron source, intended for the progress in fusion power reactor and fission technology, requires the possibility of plasma facing elements (PFE) to steady-state operation under extremely high power loads. Cooling of tokamak boundary plasma owing to radiation of none fully stripped lithium ions is considered as a promising way for PFE protection. It may be effectively realized when the main part of lithium ions are involved in the closed circuit of migration between plasma and PFE surface. Such power exhaust process may be implemented with the use of in-vessel lithium device based on lithium capillary-pore system (CPS) as plasma-facing material. CPS ensures the PFE self-regeneration during steady-state operation, provides the returning of collected and condensed lithium to emitting zone that prevents lithium accumulation outside the plasma interaction area. Several types of limiters based on lithium CPS with different ratio of emitting/collecting area and collectors for T-11M tokamak have been developed, created and tested with the aim of experimental substantiation for the concept of lithium closed circulation loop, investigation of lithium behavior in tokamak plasma and estimation of lithium CPS stability under high power flux. Design and main experimental test results are presented and discussed for horizontal, vertical, longitudinal lithium limiters and also for devices of lithium collection and extraction from the tokamak chamber - ring-type collector, cryogenic target.
        Speaker: Dr Alexey Vertkov (JSC “Red Star”)
      • 08:30
        Experimental Investigation of the System of Vertical and Longitudinal Lithium Limiters as a Prototype of Plasma Facing Components of a Steady State Tokamak-Reactor on T-11M Tokamak 4h
        During operation on the tokamak T-11M it was achieved almost complete (up to 80%) closing of lithium circulation circuit between the edge of hot (106K) plasma and the chamber wall. Lithium, emitted by the vertical capillary Li limiter during operations of T-11M has been collected by the cryogenic target and removed outside the tokamak vacuum chamber without disturbing of tokamak operation cycle, what is a key requirement for use of lithium in a steady state tokamak-reactor. In T-11M it was tested a new functional model of the prototype of closed lithium circuit for the protection of chamber wall by a simultaneous using of the vertical lithium limiter T-11M as an emitter of lithium and new longitudinal lithium limiter as its collector. Such technological scheme can be suggested for the steady state fusion neutron source (FNS) on the tokamak basis.
        Speaker: Prof. Sergey Mirnov (TRINITI)
      • 08:30
        Experimental Quantification of the Impact of Large and Small Scale Instabilities on Confined Fast Ions in ASDEX Upgrade 4h
        The transport of fast, suprathermal ions as generated by neutral beam injection (NBI) is an important topic in fusion research. In unperturbed plasmas the fast-ion transport is expected to be neoclassical, i.e. dominated by collisions, while an anomalous transport is observed in the presence of instabilities. This anomalous fast-ion redistribution must be investigated in detail because it may limit the heating and current drive performance in future fusion devices and could even damage the first wall. At the ASDEX Upgrade tokamak, fast ions are generated by up to 8 different NBI sources and their distribution function can be measured by a comprehensive set of diagnostics: A multi-view fast-ion D-alpha (FIDA) spectroscopy diagnostic, neutral particle analyzers, neutron measurements and fast-ion loss detectors (FILD) permit studies in velocity space and real space. A reduction of the central fast-ion density of up to 50% has been measured in experiments with strong sawtooth activity. Outside the q=1 surface, a corresponding increase is observed while the FILDs show no significant fast-ion redistribution to unconfined orbits. Simulations that assume flux aligned transport can explain the radial shape of this measured internal fast-ion redistribution well, but they underestimate its magnitude which could be explained by ExB drifts. Measurements during strong activity of reversed shear Alfvén eigenmodes (RSAE) show significant differences to neoclassical predictions. The observed radial fast-ion profiles are strongly flattened and broadened compared to the predictions. Non-linear simulations of the RSAE-induced fast-ion transport using the HAGIS/LIGKA code will be compared with the experimental data. In MHD-quiescent plasmas, a clear change of the radial fast-ion profiles is measured when replacing on-axis NBI with off-axis NBI. Neoclassical predictions are in very good agreement with this observation while simulations that assume an anomalous transport of 0.5m2/s do not fit the experimental data. This is, however, in contradiction with previous off-axis NBI current drive experiments that related the absence of a measurable fast-ion driven current to a turbulence-induced fast-ion redistribution. New off-axis NBI current drive experiments are, hence, being conducted to resolve this contradiction and the results of this investigation will be discussed.
        Speaker: Dr Benedikt Geiger (Max-Planck-Institut für Plasmaphysik, Garching, Germany)
        summary
      • 08:30
        Fast Particle Behavior in Globus-M 4h
        Behavior of the fast particle population, arising during 20-30 keV hydrogen and deuterium neutral beam injection in the hydrogen and deuterium plasmas, is investigated. Experiments revealed large fast ion losses. Experimental results are confirmed by different types of modeling: simulation with the NUBEAM module, solution of Boltzmann kinetic equation with Landau collision term and full 3D fast ion tracking algorithm. Dynamics of the energetic particle losses during the sawtooth oscillations and Alfven eigenmodes is investigated. Losses, induced by these instabilities, may exceed 25%. A way to decrease fast ion losses in present conditions is shown. Modeling for the Globus-M2 tokamak is performed. Simulations predict essential improvement in the fast ion confinement.
        Speaker: Mr Nikolay Bakharev (Ioffe Physical-Technical Institute of the Russian Academy of Sciences)
        Poster
        summary
      • 08:30
        Fast-Ion Response to Externally Applied 3D Magnetic Perturbations in ASDEX Upgrade H-Mode Plasmas 4h
        The fast-ion response to externally applied 3D Magnetic Perturbations (MPs) has been investigated on ASDEX Upgrade (AUG) in H-mode plasmas with a wide range of collisionalities / densities and MP spectra. MPs have little effect on kinetic profiles, including fast-ions, in high collisionality plasmas with mitigated ELMs while a strong plasma (including fast-ions) response is observed in H-mode regimes with low collisionality / density and low q95. Multiple, absolutely calibrated, fast-ion loss detectors (FILDs) located at different toroidal and poloidal positions measure significant changes in escaping ion phase-space when MPs are applied. Fast-ion losses can be up to an order of magnitude larger with MPs than the nominal NBI prompt losses measured without MPs. The application of the 3D fields is followed by a rapid rise (within ms) of the associated fast-ion losses while the measured fast-ion losses exhibit a slow decay, ~100 ms, down to the nominal NBI prompt loss level, after the MP coils are switched off. The heat load associated to the MP induced fast-ion losses have been measured with infrared cameras imaging the divertor as well as FILD and the surrounding first wall. The measured heat load can be up to 6 times larger with MPs than without MPs. The impact the 3D fields have on the confined fast-ions have been monitored by means of Fast-Ion D-Alpha (FIDA) spectroscopy. FIDA measures an enhancement of the fast-ion content in plasma with a visible impact on the gradients of the fast-ion profiles when RMPs are applied and density pump-out is observed. A strong fast-ion response is typically accompanied by an apparent displacement of the outboard separatrix, 1-3 cm, as measured by Beam Emission Spectroscopy (BES) that modifies significantly the NBI deposition profile. The accurate fast-ion measurements presented here are used to test models of 3D fields using full orbit simulations. The perturbed equilibria are calculated in vacuum, using the 3D free boundary VMEC / NEMEC code as well as including the plasma response with the M3D-C1, MARS-F and JOREK codes.
        Speaker: Dr Manuel Garcia-Munoz (University of Seville, Spain)
      • 08:30
        Feedback of a Neoclassical Tearing Mode on Drift Wave – Zonal Flow Turbulence 4h
        We study the feedback loop of a spontaneous magnetic perturbation (neoclassical tearing mode) on a background of drift wave – Zonal Flows, in the framework of a 1D predator-prey model for the evolution of turbulence intensity I(x,t), Zonal Flow energy U(x,t), electron temperature gradient T(x) and island-width. A modified Rutherford equation describes the magnetic island dynamics. The magnetic island is driven by the neoclassical bootstrap current, and acts to damp turbulence-driven Zonal Flows, while turbulence also affects the island-chain evolution by flattening the temperature profile, thus depleting the bootstrap current. A critical issue in fusion devices is what determines the threshold island width, set by a competition between parallel heat conduction along tilted field lines v.s. perpendicular diffusion across field lines. Since perpendicular heat diffusion is mostly turbulent, the threshold island width is determined by turbulence. As turbulence is regulated by Zonal Flows, this threshold is ultimately tied to Zonal Flow intensity and thus Zonal Flow damping. As the island grows, Zonal Flows are strongly (but locally) damped and turbulence remains un-quenched in the vicinity of the island. In effect, they are two plasma regions divided by the island separatrix: Outside of the magnetic island, the plasma is ZF-dominated, whereas inside the island, the plasma is pushed back to a saturated turbulence regime (L-mode-like regime). Turbulence spreading across the island-separatrix occurs at quasi-periodic intervals and triggers temperature profile collapse, due to turbulent heat diffusion. Our results suggest that the threshold island-width is modulated by the microturbulence at the Limit-Cycle Oscillation frequency of the predator-prey system. Hence, the threshold island-width is ultimately set by ZF damping.
        Speaker: Dr Michael Leconte (NFRI)
        Poster
        summary
      • 08:30
        Filament Transport in the SOL of ASDEX Upgrade 4h
        At the edge of fusion plasmas, intermittently expelled density filaments, so-called blobs [1], are propagating through the scrape-off layer (SOL) perpendicular to the magnetic field. Due to its higher density and temperature compared with the background SOL plasma, they can lead to a significant degradation of plasma facing components in the main chamber. Since this degradation is critical for the first wall in future fusion devices, an understanding of the generation and the propagation of blobs is needed. Therefore, the dynamics of blob filaments is investigated in the SOL of ASDEX Upgrade by means of Lithium beam emission spectroscopy (Li-BES) [2], Langmuir probes, and gas puff imaging. This way, the density, velocity, lifetime, frequency and size of the blobs perpendicular to the magnetic field are determined. A comparison of the measurements with a recently developed analytical blob model based on a drift-interchange-Alfvén fluid model [3] indicates an influence of a finite ion temperature on the blob dynamics which has typically been neglected in other blob models. The blob dynamics agree well with the sheath-connected regime at lower plasma densities, and inertial effects play only a minor role [4]. At higher densities, a transition into another regime with large blob amplitudes and increased transport is found [5]. This points to a prominent role of blob transport at higher Greenwald fractions and has implications for the gross erosion of wall material in reactor relevant operation scenarios with a detached divertor near the density limit. [1] D. A. D’Ippolito, J. R. Myra, and S. J. Zweben, Phys. Plasmas 18, 060501 (2011) [2] M. Willensdorfer et al., Plasma Phys. Control. Fusion 56, 025008 (2014) [3] P. Manz et al., Phys. Plasmas 20, 102307 (2013) [4] G. Birkenmeier et al., Proceedings of the 40th EPS Conference on Plasma Physics 2013, Espoo, Finland [5] D. Carralero et al., Proceedings of the 40th EPS Conference on Plasma Physics 2013, Espoo, Finland
        Speaker: Dr Gregor Birkenmeier (Max Planck Institute for Plasma Physics, Garching, Germany)
        summary
      • 08:30
        From Micro to Macro: L-H Transition Dynamics and Power Threshold Scaling 4h
        It is believed that L-H transition occurs via coupling of turbulence to low frequency shear flows by Reynolds work. As a consequence, turbulence and turbulent transport collapse, enabling the growth of diamagnetic electric field shear and the transition. This work focuses on the missing link between microscopics and macroscopics, and its critical role in power threshold scaling. The major goal is the understanding of the observed occurrence of a minimum in the power threshold. We pursued a model which separates electron and ion temperature evolution by extending a recent 1D, five-field model which captures the transition layer evolution well, but does not have this capability. In the new model, density appears as an electron-ion coupling parameter, as well as in ZF damping. We propose and examine the explanation that: (i) the initial trend of decrease in the power threshold is due to stronger collisional electron-ion coupling which enables the development of stronger diamagnetic electric field. This scaling trend reflects the role of the mean shear in locking-in of the transition, (ii) the subsequent increase in the threshold is due to the increase in damping of shear flows with ion collisionality. This scaling trend reflects the role of the turbulence generated shear flow as a trigger mechanism. Our studies reveal a clear power threshold minimum in density scans ran for a fixed, electron dominated heating mix, but an even more distinct minimum is predicted for the fixed density when scanning the ion to total heating ratio. Here we see that the power threshold minimum appears as an interplay of electron-ion coupling. In addition to the basic scaling trends, model studies reveal: (a) the threshold power increases for off-axis electron heat deposition. This follows from the fact that electron-ion coupling is reduced in this instance, (b) a minimum power is predicted for a heating mix scan as well as for a density scan. This points towards the possibility of a global minimum in the threshold power in terms of a number of relevant parameters, (c) no clear threshold minimum is predicted by this model for pure ion heat deposition. Ongoing work is concerned with quantifying the strength of hysteresis in terms of multiple macroscopic parameters and with relating this to observed back-transition shear flow and turbulence dynamics.
        Speaker: Dr Mikhail Malkov (University of California, San Diego)
        summary
      • 08:30
        Fully Non-Inductive Current Drive Experiments Using 28 GHz and 8.2 GHz Electron Cyclotron Waves in QUEST 4h
        28 GHz Electron Cyclotron Current Drive (ECCD) effect was clearly observed in Ohmically heated plasmas with feedback regulation of center solenoid coil current in 2nd harmonic inboard off-axis heating scenario. In non-inductive current drive experiments only by the 28 GHz injection, 54 kA plasma current was sustained for 0.9 s. Higher plasma current of 66 kA was non-inductively obtained by slow ramp-up of vertical field using the 28 GHz ECH/ECCD. Non-inductive high-density/ current plasma start-up, which is a key issue for fusion reactor design has been demonstrated using 2nd harmonic ECH/ECCD. Density jump across 8.2 GHz cutoff density was observed in superposed 28 GHz / 8.2 GHz injections. The 50 kA plasmas were sustained by the 8.2 GHz injection into the 28 GHz target plasma if the stable plasma shaping was obtained.
        Speaker: Dr Hiroshi Idei (Research Institute for Applied Mechanics, Kyushu University)
        summary
      • 08:30
        Generation of Energetic Electrons by Magnetic Reconnection with Presence of High Guide Field 4h
        Magnetic reconnection allows highly-conducting plasma to change its magnetic topology in nuclear fusion plasmas, such as sawtooth crash, internal reconnection event, and so on. Recent theoretical and numerical works revealed that the presence of guide field (GF) greatly changes the collisionless reconnection in a qualitative way. One of the essential changes is that electrons are efficiently accelerated near the X-point to achieve high kinetic energy. Those fast electrons then cause secondary modification on reconnection structure, sometimes involving excitation of waves by electron beam instability. In this paper, we report some evidences of accelerated electrons in the UTST device, which provides well-controlled reconnection condition with toroidal GF 20 times higher than the reconnection field. During the non-steady reconnection process in the UTST, a sharp ring-shape emission was found near the X-point. Toroidally accelerated electrons by reconnection electric field are supposed to ionize singly charged carbon impurities in the middle phase of reconnection. The SXR from the X-point region was observed simultaneously with the reconnection electric field. Generation of electron high energy tail up to 300 eV was confirmed by comparison of SXR signals through various filters. The SXR emission showed almost linear increasing trend with the toroidal GF when it exceeds the threshold value of GF ~ 0.12 T, suggesting that the number of accelerated electrons is determined by the duration of electrons remaining near the X-point. As a consequence, highly-efficient electron acceleration takes place in the magnetic reconnection with high GF even though the released magnetic energy was not very large. In contrast, the ion flow acceleration was observed only in the cases with reconnection electric field higher than 100 V/m. Thus, in the high GF case, the released magnetic energy is mainly converted to the kinetic energy of bulk ions, which then is converted to thermal energy; however, some electrons are effectively accelerated by the reconnection electric field to form a high energy tail. The accelerated fast electrons could excite low frequency modes near the reconnection region, which may cause bad influences on confinement property of tokamak plasmas.
        Speaker: Prof. Michiaki Inomoto (The University of Tokyo)
      • 08:30
        Geodesic Acoustic Mode Investigation in the Spherical Globus-M Tokamak using a multi-diagnostic approach 4h
        Owing to the active researches of intermediate regime which is known as limit cycle oscillation (LCO) regime, it is important to investigate oscillations of velocity of zonal flows, plasma density and magnetic field simultaneously. The geodesic acoustic mode (GAM) investigations using multi-diagnostic were carried out on the spherical tokamak Globus-M (R = 0.36 m, a = 0.24 m, Ip = 150 kA, BT = 0.3 T). Variation of GAM intensity showed features of LCO regime. The key implement for researches was a method of Doppler microwave backscattering, or Doppler reflectometry (DR). The diagnostic consisted of two Doppler reflectometers, which were apart in poloidal direction. Oscillations of plasma density at GAM frequencies were investigated using electrostatic probes and detectors of Dα intensity emission, which were detected at different lines of sight. Structure of magnetic field perturbation has been investigated using poloidal and toroidal Mirnov probe arrays. Correlation analysis shown that the structure of ExB velocity oscillations had corresponded to n=0, m=0. The main structure of density fluctuations at GAM frequency had n=0, m=0 mode numbers. Besides the n=0, m=1 component is also presented. The mode structure for magnetic field disturbance was established. To interpret an origin of plasma density global perturbation at GAM frequency plausible model was suggested.
        Speaker: Dr Viktor Bulanin (St.Petersburg State Polytechnical University)
        summary
      • 08:30
        Global Gyrokinetic Modeling of Geodesic Acoustic Modes and Shear Alfvén Instabilities in ASDEX Upgrade 4h
        In this work, we investigate theoretically the dynamics of global instabilities observed in ASDEX Upgrade (AUG) by means of collisionless numerical simulations. We focus in particular on geodesic acoustic modes (GAM) and shear Alfven instabilities. The numerical tools we use are the codes NEMORB (nonlinear global gyrokinetic PIC), LIGKA (linear global gyrokinetic) and XHMGC (nonlinear global hybrid). In the first part of this work, results of axisymmetric simulations of GAMs with gyrokinetic codes NEMORB and LIGKA with AUG equilibrium profiles are shown. In the second part, we show results of axisymmetric electromagnetic simulations with NEMORB and LIGKA in the presence of an EP population. Finally, in the third part, we show results of single-n (with n the toroidal mode number) numerical simulations of shear Alfven instabilities with NEMORB, LIGKA and XHMGC. Comparisons with analytical theory and experimental data are also shown, for each case of interest.
        Speaker: Dr Alessandro Biancalani (Max-Planck-Institut für Plasmaphysik Euratom Association, Garching, Germany)
        summary
      • 08:30
        Identification of Intrinsic Torques in ASDEX Upgrade H-Mode Plasmas 4h
        Previous work performed has amassed a substantial database of intrinsic rotation measurements in various tokamak devices. However, as our understanding of momentum transport has evolved, it has become clear that a reliable prediction of the rotation in future devices requires a more complete momentum transport model and a more fundamental understanding of the mechanisms driving the intrinsic rotation. Initial estimates from this database project a large intrinsic velocity (~300 km/s) for ITER. The next step in intrinsic rotation studies is to characterize the “intrinsic torque” associated with its generation. The primary goal of this paper is to clarify whether or not the edge localized intrinsic torque scales with the pedestal strength on the ASDEX-Upgrade tokamak. In the q-profile scan resulting effectively in a pedestal strength scan (factor of 2-3 variation in pedestal top values of Ti, Te and ne), the integrated intrinsic torque profiles show clearly that the intrinsic torque increases with increasing plasma current on the outer half of the plasma radius. All the cases at different currents have in common that the intrinsic torque has a rather broad profile with the main contribution coming from outside r/a=0.4 which is somewhat different from what has been observed previously on DIII-D where the intrinsic torque is more pronouncedly peaked at the edge. In the ECRH power scan, the intrinsic torque during the low ECRH power phase is in the co-current direction and increases toward the plasma edge. In the high power ECRH phase, however, a negative (counter-current) torque source is present from the centre of the plasma up to mid radius consistent with previous work on AUG. The ECRH power deposition is centrally located. In addition to the counter intrinsic torque, the rotation modulation data cannot be explained by any other mechanism than outward convection at r/a < 0.4. The first intrinsic torque experiments on AUG show that co-torque is driven in the outer part of the plasma radius and that counter-torques can develop in the inner half radius when sufficient ECRH is applied to alter the heat transport and the local plasma turbulence.
        Speaker: Dr Tuomas Tala (VTT)
        summary
      • 08:30
        Impact of Isotopic Effect on Density Limit and LHCD Efficiency in the FT-2 Experiments 4h
        Current drive by lower hybrid waves (LHCD) is the most effective method to keep the plasma current, but it is feasible only at the plasma density not exceeding some density limit n_DL. In the present work the main attention is paid to investigation of this effect on the FT-2 (R=0.55 m, a=0.08 m, B_T ≤ 3T, I_p=19÷40 kA, f_0 = 920МHz) tokamak. The dependence of LHCD efficiency on isotopic plasma content (hydrogen/deuterium) is studied. On the FT-2 tokamak, where a large experience has been accumulated in the area of plasma – LH wave interaction observation, the long-continued experimental run on LHCD efficiency study has been realized. Characteristic features of such experiment are strong influence of the isotope plasma composition on the LH resonance density nLН. For hydrogen plasma n_LН ~ 3.5 10^13 см^-3, whiles for deuterium n_LH ~ 10^14см^-3. The suppression of the LHCD and beginning of the interaction of LH waves with ions is controlled by the hydrogen/deuterium plasma density rise. In the hot hydrogen plasma (Te(r=0cm) ≈ 700eV) the density limit n_DL of LHCD is approximately equal to the resonance value n_LH at which the interaction of the LH wave with the electron component is replaced by direct absorption by plasma ions (nLH≈ nLC =3.5•10^13 cm^-3 is the point of linear conversion). In the hot deuterium plasma one could expect an increase of nDL because of a much higher value of n_LH ≥ n_LC ≈ 10^14 cm^-3. However it appeared that the observed density limit for LHCD generation n_DL ≈ (3.5÷4)•10^13 cm^-3 is not determined by nLC. Role of parametric instabilities in CD switch-off is considered in both cases. The cooling of the plasma column and density rise could lead to a reduction of the threshold for the parametric decay of f0 and result in the earlier suppression of LHCD. In both cases the LHCD was inversely proportional to the density, which corresponds to the theoretical predictions. In order to analyze the experimentally observed effects the GRILL3D and FRTC codes have been used. The important role of the synergetic effect caused by the interaction of different spectral components of the excited RF waves was revealed. Next step of LHCD modeling is devoted to a dynamic modeling of LHCD plasma shots at rather low plasma densities <n_e> = 0.5 ÷ 2•10^13 cm^-3, when role of runaway electrons is significant at the FT-2 conditions.
        Speaker: Dr Sergey Lashkul (Ioffe Physical-Technical Institute, RAS, Russia)
        summary
      • 08:30
        Impurities Removal during Central ECR Heating in T-10 4h
        Experiments on impurities removal with central ECR heating on T-10 were carried out with various plasma parameters. CXRS and Zeff measurements show removal of carbon nuclei from plasma during central ECRH. There is a complex of high-power gyrotrons for ECRH experiments on T-10. Spectroscopic diagnostics of T-10 allows to measure carbon concentration by CXRS diagnostics [1] and to measure effective ion charge (Zeff) radial distribution from bremstrahlung intensity with subtraction of background molecular linear spectra. Radial distribution of C+5 passive line intensity was measured simultaneously with CXRS measurements. This linear radiation allows to estimate ionization flux from C+5 into C+6 ions for calculation of C+6 particle confinement time τparticle: Value of τparticle can be successfully defined only for impurities with relatively high Z (like carbon) due to the ionization flux into the nuclei in toroidal and poloidal coordinates for carbon impurity is homogeneous in T-10 conditions. Carbon confinement time in OH regimes rises with increasing of line averaged electron density and decreasing of plasma current. In ECR heated regimes with PECR=1 MW one observes sharp decrease of carbon confinement time to a value τparticle≈23 ms, which is almost the same for various plasma parameters within the error limits. The most contrast fall of τparticle is observed at high plasma densities when one observes sharp decrease of total carbon concentration during central ECRH, although carbon ionization flux in ECR regimes is about twice higher than in OH plasma. Work was carried out by "Rosatom" 13.05.2013 № H.4x.44.90.13.1101
        Speaker: Mr Leonid Klyuchnikov (Institute of Tokamak Phisycs, National Research Centre "Kurchatov Institute", Moscow, Russia)
        Poster
        summary
      • 08:30
        Integrated Modeling of the Globus-M Tokamak Plasma 4h
        In the present paper the results of integrated modeling of Globus-M tokamak plasma with the help of recently coupled core transport code ASTRA and edge transport code B2SOLPS are presented. In the modeling taken into account are the neoclassical transport, auxiliary heating and current drive by the NBI, 2D drift fluxes, currents and electric field in the edge plasma in a real geometry of magnetic flux surfaces and first wall constructions of a spherical tokamak. It is demonstrated that the modeling results are in a satisfactory agreement with laser and probe measurements and fast neutral particle analyzer and neutron analyzer signals for various plasma current values both in ohmically heated and NBI-heated discharges. The dependence of the scrape-off layer structure and the heat loads to the divertor targets on the plasma current and the discharge power is investigated. It is found that the heat flux decay length agrees with a predictions of the scaling [T.Eich et al., Phys.Rev.Lett., 107, 215001 (2011)] stemming from a large multi-machine experimental database. Thus this scaling passed through the tests versus experimental data from a small spherical tokamak with small magnetic field, plasma current and power, and the database is expanded correspondingly.
        Speaker: Dr Ilya Senichenkov (Saint Petersburg State Polytechnical University)
        summary
      • 08:30
        Investigation of a Phenomenology of the Improved Confinement Regime in T-11M Tokamak 4h
        Regimes of discharge of improved plasma confinement were found in the experiments with a vertical lithium limiter on T-11M tokamak, which was manifested in spontaneous growth of the electron density up to the limit Greenwald and above. Previously, such modes arising after chamber lithiization were observed in the tokamak FTU. The analysis of data obtained in T-11M has showed that this regime of improved confinement is differed from the ordinary mode (L-mode) by the sharp profile of the plasma density N_e(r), relatively high values of N_e(0) in the center and ordinary density values N_e(a) in the edge of plasma, as if there is an internal transport barrier in the center. Another visible difference of this mode from L-mode is a significant increase of soft X-rays power from the center, while maintaining or even reducing Z_eff(0). In the central region of plasma the lifetime of particles is increased by approximately twice. The energy lifetime is increased by 30-40%. Major differences of the regime of improved confinement from remarkable limiter H-mode observed in T-11M, for example, after boronization are peaking of the density profile, absence of active ELMs and collection of impurities in the plasma center. An analogy of such regimes is carried out with previously detected on the Alcator C-Mode I-mode.
        Speaker: Mrs Anastasia Shcherbak (SRC RF TRINITI)
      • 08:30
        Investigation of Progression from Low to High Hydrogen Recycling during Long Duration Discharges on a Spherical Tokamak, QUEST 4h
        Progression from low (LR) to high recycling (HR) was observed in full non-inductive long duration discharges up to 5 minutes on QUEST. Transitional repetitive behavior between LR and HR was induced by periodic gas puffing and the period to recover to LR, tau_rec, was gradually prolonged. The period, tau_rec normalized by gas rate has a linear relation to time-integrated H_alfa. As the prolongation of tau_rec was also induced by higher gas rate even in the start-up phase, the value of tau_rec is an index of the amount of recycled hydrogen. The experimental observation indicates hydrogen recycling rate is dominantly depending on hydrogen fluence to the wall. To understand the dependence, deuterium storing capability of the specimen exposed to QUEST plasmas during an experimental campaign was investigated by implantation of deuterium molecule ions of 1keV and subsequent thermal desorption spectrum (TDS) as a post-mortem analysis. The important desorption in the obtained TDSs appeared around 420 and 470K, and these peaks can be reconstructed by a model including diffusion, recombination, trapping, and plasma induced desorption. The model calculation was applied to the QUEST long duration discharges and shows that recycling ratio has a clear dependence on fluence and the fluence in the QUEST long duration discharges is sufficient to make a saturation in recycling ratio of unity. These results indicate that hydrogen recycling has the capability to provide a clear effect on plasma in long duration discharges and the progression is driven by enhanced hydrogen recycling with high fluence to the wall.
        Speaker: Prof. Kazuaki Hanada (Advanced Fusion Research Center, Research Institute for Applied Mechanics, Kyushu University)
      • 08:30
        Magnetic Island and Plasma Rotation under External Resonant Magnetic Perturbation in T-10 Tokamak 4h
        The experimental comparison of the m=2 mode and plasma rotation velocities at the q=2 magnetic surface in a range of the mode amplitudes is presented in this paper. The phase velocity of the mode rotation is measured with a set of poloidal magnetic field sensors located at the inner side of the vacuum vessel wall. The plasma rotation velocity at the q=2 magnetic surface in the direction of the mode phase velocity is measured with the heavy ion beam diagnostics. In the presence of a static RMP the rotation is irregular that appears as cyclical variations of the mode and plasma instantaneous velocities. The period of these variations is equal to the period of the mode oscillations. In each period the velocities depend on the angular shift between the mode and RMP. A non-monotonic dependence of the mode rotation irregularity on the mode amplitude is observed. The rotation irregularity increases in both cases of sufficiently big and small amplitudes. In the case of big mode amplitude the rotation irregularity of the mode coincides with the rotation irregularity of the resonant plasma layer. On the contrary, the observed rise of the mode rotation irregularity in the case of sufficiently small mode amplitude is not followed by an increase of the rotation irregularity of the resonant plasma layer. It means that a decoupling between the mode and plasma rotations is observed for small islands. The experimental results are simulated with the TEAR code based on the two-fluid MHD approximation. The effects of plasma resistivity, viscosity, RMP and the current induced in the resistive vacuum vessel are taken into account. The calculated irregularities of the mode and plasma rotation depend on the mode amplitude similar to the experimental data. For large islands, the rotation irregularity is attributed to variations of the electromagnetic torque applied to the resonant layer. For small islands, the deviations of the mode rotation velocity from the plasma velocity take place due to the effect of finite plasma resistivity.
        Speaker: Dr Nikolay Ivanov (Kurchatov Institute)
        summary
      • 08:30
        Modelling of Pulsed and Steady-State DEMO Scenarios 4h
        An intensive programme has been started in the EU, aiming at a more and more refined selection of the DEMO design. The general strategy adopted consists in developing two DEMO concepts in parallel: a pulsed tokamak, characterised by rather conventional physics and technology assumptions (DEMO1) and a steady-state tokamak, with moderately advanced physics and technology assumptions (DEMO2). The physics assessment part of this programme involves three main steps: i) the analysis of the general physics guidelines of a tokamak DEMO; ii) the search for optimum working points, performed by means of systems codes, i.e., 0-D codes combining both physics and technology constraints; iii) space and time dependent simulations of plasma scenarios, performed by means of integrated modelling codes with various levels of assumptions. In this last area of work, a coordinated effort has been undertaken at the EU level, as an EFDA Task Agreement during 2012 and 2013. The general goal of this Task was the analysis of working points produced by the systems code PROCESS for both DEMO1 and DEMO2 by means of various integrated modelling codes. Iterations between systems codes and scenario modelling should eventually converge to the definition of optimum working points that are consistent with the physics guidelines. The main results of this work on scenario modelling are reported here. The computational tools used for these studies are: - The 0.5-D integrated modelling code METIS - The coupled core-edge code COREDIV - The 1.5-D integrated modelling codes ASTRA, JINTRAC and CRONOS Starting from the 0-D outputs of the PROCESS code for both DEMO1 and DEMO2 working points, the following steps have been performed: - test of the consistency of the PROCESS working points by exploratory runs of METIS. Iterations with PROCESS in order to improve the working points - assessment of density and temperature profiles consistent with the reference working points and with first-principle particle transport models (TGLF, GLF23), by ASTRA - assessment of impurity and radiation profiles consistent with the reference working points and with suitable impurity transport assumptions by COREDIV - global scenario assessment by 1.5-D simulations with JINTRAC and CRONOS - sensitivity analysis for variations of selected plasma and machine parameters and of transport models (by METIS, ASTRA, COREDIV)
        Speaker: Dr Gerardo Giruzzi (IRFM, CEA)
        summary
      • 08:30
        Nonlinear and Toroidal Mode Coupling Effects on m=1, n=1 Instabilities 4h
        Instabilities with poloidal and toroidal mode numbers m/n=1/1 remain an important concern for fusion in toroidal plasmas. Sawtooth crashes can periodically reduce the central plasma pressure and fusion rate or trigger more dangerous instabilities. Recent experimental results[1-3] have identified new types of 1/1 modes around and inside the q=1 magnetic surface. Nonlinear full MHD numerical simulations with M3D and their analysis[1,2,4,5] demonstrate that these modes are dynamic and strongly influenced by toroidal and nonlinear mode coupling, effects that have been ignored in most nonlinear theories. As in experiment, multiple 1/1 structures can appear simultaneously around and inside q=1. They include long-lived 1/1 helical density concentrations or ``snakes'' and a variety of internal kink like modes. Background plasma toroidal rotation is important; 1/1 modes tend to rotate coherently with the plasma. Some snakes, such as those due to heavy impurity ions, can form around q=1 without a magnetic island. The states tend to minimize the free energy, since the 1/1 helical temperature develops opposite sign to the helical density, reducing the non-axisymmetric pressure. Snakes can coexist with and partially stabilize periodic sawtooth crashes inside q<1. At low resistivity, the 1/1 resistive internal kink and sawtooth crash are shaped by terms in the momentum balance that are higher order in inverse aspect ratio, even at small r_1/R_o=1/10. The narrow Sweet-arker-like reconnection layer of reduced MHD rarely develops. Instead, a fast crash phase driven by toroidal nonlinearity, enhanced by these terms, matches experimental crash times and the observed temperature redistribution. The crash does not follow the Kadomtsev sawtooth model because the density is not tightly tied to the magnetic field lines. The higher order terms and mode coupling can affect other instabilities, such as m>1 magnetic islands and plasma edge instabilities. Work supported by the US DOE Office of Fusion Energy Sciences and SciDAC programs. [1] L. Delgado-Aparicio, L. Sugiyama, et al., Phys. Rev. Letters 110, (2013) 65006. [2] L. Delgado-Aparicio, L. Sugiyama, et al., Nucl. Fusion 53, (2013) 043019. [3] L. Delgado-Aparicio, et al., submitted to Phys. Rev. Letters (2014). [4] L.E. Sugiyama, Phys. Plasmas 20, 032504 (2013). [5] L.E. Sugiyama, Phys. Plasmas, to appear (2014).
        Speaker: Dr Linda Sugiyama (Massachusetts Institute of Technology)
        summary
      • 08:30
        Nonthermal Microwave Emission Features under the Plasma Ohmic Heating and Lower Hybrid Current Drive in the FT- 2 Tokamak 4h
        Results of studying the abnormal microwave emission (ME) arising under ohmic heating (OH) of the moderately dense plasma and generation of the low-hybrid current drive (LHCD) in the FT-2 tokamak are presented. The ME appearance is due to the «fan» instability development and the substantial local magnetic ripples existence. It was found that the ME arises continuously during OH in the frequency range (10 – 40)GHz and is accompanied by short, «gaint» flashes, which are greatly larger than the pedestal and have the narrow frequency spectrum. The synchrotron emission (SE) growth and less intensive flashes appear also in the range (57-75)GHz. As known they arise under the maser amplification of SE during interaction of AE with harmonics of magnetic ripples in the electron cyclotron autoresonance. Owing to the non-linear transformation of excited electron plasma waves into electromagnetic ones collective emission, CE, appears. It becomes possible the maser amplification of both SE and CE . The less intensive ME flashes arise apparently in the SE maser amplification only. The «gaint» flashes may be initiated under suitable conditions by transition of the maser – amplifier into the self–excitation regime, when the short powerful flashes of low-frequency coherent ME are generated. The plasma HF-pumping at the low-hybrid frequencies during the quazistationary OH stage provides the effective LHCD. The first high fast electron heating from 300 eV up to 550 eV at Phf = 90 kW was registered in this regime together with the radiation losses growth due to the SE intensity increase in the (53-156)GHz range. It was accompanied by short ME spikes of the maser nature observed in the more narrow frequency range (53 ÷ 78)GHz. The such ones were observed earlier under the plasma OH. It is possible this additional electron heating is due to the SE and spikes of ME absorbtion in the black plasma layers.
        Speaker: Mr Vladimir Rozhdestvensky (Ioffe-Institute)
      • 08:30
        On Anomalous Dissipation and Relaxation in ELMs 4h
        We present a new dynamical model of pedestal ELM phenomena based upon the multi-scale interaction between low-n MHD ballooning mode and short scale ETG turbulence. ELM dynamics are determined by the few basics process results from multi scale interaction. These includes: generation of hyper resistivity ( ) in coupled ballooning mode - ETG turbulence; excitation of hyper resistive BM near ideal MHD threshold; regulation of via feedback loop between hyper resistive and ETG mode; formation of steep current and pressure gradients between primary resonances by process of gradient pinching. It is argued that gradient pinching, which occurs as primary modes grow, will destabilize dissipative convective cells throughout the pedestal. In particular, these cells will be driven in the region between the primary helicity. This ensemble of BMs and dissipative BMs will result in fast relaxation throughout the pedestal. Note that the multi-helicity interaction effectively spreads the relaxation throughout the region of the pedestal.
        Speaker: Prof. Raghvendra Singh (NFRI-Korea / IPR-India)
      • 08:30
        Poloidal Inhomogeneity of Turbulence in the FT-2 Tokamak by Radial Correlation Doppler Reflectometry and Full-f Gyrokinetic Modeling 4h
        The drift-wave turbulence responsible for anomalous transport of energy and particles in tokamak plasma is widely studied nowadays both experimentally and in theory. An interesting and important prediction of the numerical approach based on full-f gyrokinetic modeling is statistical inhomogeneity of trapped-electron-mode turbulence typical for ohmic discharge. Though the radial variation of the mean turbulence characteristics is well known to the experimental community the data on their poloidal dependence are rare. In this paper we address the problem using radial correlation reflectometry (RCR) technique utilizing simultaneous microwave plasma probing at different frequencies in the presence of a cut off and based on correlation analysis of backscattering signals. Oblique plasma probing as a method to cope with contribution of small angle scattering off long scale turbulence component have been justified recently. It was proved that the radial correlation Doppler reflectometry (RCDR) version of the diagnostic provides a way for determination of the turbulence radial wave number spectra and its detailed investigation. Following this approach the RCDR scheme in the 50-75 GHz frequency range from high magnetic field side have been assembled and detailed measurements in different poloidal octants were performed in FT-2 ohmic hydrogen and deuterium discharges. The turbulence parameters were measured at variable incidence angle (±10-30 degrees) corresponding to different turbulence poloidal wave numbers (6-16 cm-1) and frequency changing from 50 kHz to 400 kHz. The cut-off layer minor radius was varied in the range from 3 to 6 cm. Both frequency resolved and integrated RCDR CCFs were determined. The correlation length corresponding to the frequency averaged CCF is smaller than that, obtained in the Doppler shift frequency thus leading to its underestimation. A new antennae set from low field side and interferometer antennae are utilized for other poloidal octants. As a result of measurements using these antennae the variation of radial correlation length from 0.25 cm to 0.45 cm when moving from high-field side to low-field side of the torus was demonstrated in agreement to the results of gyrokinetic modeling performed with ELMFIRE code. A well pronounced excess of the turbulence correlation length in deuterium over its value in hydrogen discharges was shown.
        Speaker: Mr Alexey Altukhov (Ioffe Institute)
        summary
      • 08:30
        Quantifying Self-Organization in Magnetically Confined Fusion Plasmas 4h
        Plasma self-organization is the frontier research area in plasma physics and its understanding is extremely important for the construction of innovative fusion configurations. Emergence, an outcome of self-organization, implies the appearance of certain large scale structures, forms or patterns, formed from a large number of simple interactions of smaller parts of the system. Motivated by these requirements, we have developed a framework (mathematical and computational), which, at the same time, makes choice of the optimal wavelet for analysis of data, enables optimal prediction of the dynamics, quantifies stages of self-organization and removes the effects of noise, when required [1]. It also includes the role of scales in the process. A spatiotemporal data of the gyrokinetic Vlasov simulation results for the ion temperature gradient turbulence, where the standard and the inward-shifted configurations of the Large helical Device are considered in this study. Although fluctuations of the electrostatic potential for zonal flows exhibit spatiotemporal chaos in both configurations, we show that self-organization is different in the two cases. Specifically, we show that complexity is more intense in the standard configuration, however the increase in time of complexity is higher in the inward-shifted configuration implying faster relaxation. These results are shown to be consistent with the results of the analysis of the spatiotemporal chaotic dynamics in the two configurations. We illustrate how this method may be used to test various confinement configurations in order to achieve the optimal self-organization under given circumstances. We also analyze the ion-saturation current measurements of three different confinement regimes, namely the L-mode, the H-mode and the dithering H-mode, in the scrape-off layer of several devices. We show how self-organization in each of the regimes may be compared and also how changes in configuration may be predicted. Finally, we discuss the versatility of the method and other potential uses in the realm of fusion plasmas.
        Speaker: Dr Milan Rajkovic (University of Belgrade, Institute of Nuclear Sciences Vinca)
        summary
      • 08:30
        Radial Electric Field and Poloidal Impurity Asymmetries in the Pedestal of ASDEX Upgrade: Quantitative Comparisons between Experiment and Theory 4h
        The formation of the H-mode transport barrier is strongly connected to the existence of a sheared plasma flow perpendicular to the magnetic field caused by a local radial electric field E_r. The strong gradients in E_r and the associated ExB velocity shear play a fundamental role in edge turbulence suppression, transport barrier formation and the transition to the H-mode. This contribution describes the nature and structure of the E_r well and its connection to H-mode confinement and discusses the impact of poloidal impurity asymmetries on the pedestal. A detailed analysis of the edge E_r and kinetic profiles revealed that in H-mode E_r and the main ion pressure gradient term, grad(p_i)/(en_i), are identical within the uncertainties. This relation corresponds to the cancellation of the poloidal components of the ion diamagnetic and ExB drifts and suggests that in the pedestal the perpendicular main ion flow is close to zero. This result is confirmed by direct measurements of the main ion temperature, density and flow velocities in helium plasmas. The main ion poloidal rotation exhibits very small values at the plasma edge, as expected from neoclassical theory. The edge poloidal flow measurements of both main ions and impurities have been compared to a hierarchy of neoclassical models. In all cases, the measurements are found to be in quantitative agreement with neoclassical theory demonstrating that in the pedestal the E_r well is sustained by the gradients of the main ion species. New charge exchange measurements at ASDEX Upgrade reveal the existence of a poloidal asymmetry in the flow pattern at the pedestal. The flow asymmetry can be explained by an excess of impurity density at the high-field side following the postulate of divergence-free flows on a flux surface. Comparison of the measured flows to theoretical predictions based on the parallel momentum balance reveals the nature of the parallel impurity dynamics. The key features of the experimental data including the shape of the rotation profiles and the poloidal impurity density asymmetry are reproduced quantitatively for the first time. The impact of these findings with respect to impurity transport at the plasma edge are presented. This project has received funding from the EURATOM research and training programme 2014–2018.
        Speaker: Dr Eleonora Viezzer (Max-Planck-Institut für Plasmaphysik)
        summary
      • 08:30
        Real-Time Control of NTMs Using ECCD at ASDEX Upgrade 4h
        In high performance plasmas, Neoclassical Tearing Modes (NTMs) are regularly observed at large beta-values. NTMs reduce the achievable normalized beta and degrade the fusion reactor performance which scales as beta_N squared. A widely used method, also foreseen for ITER, for avoiding and controlling NTMs is the deposition of electron cyclotron current drive (ECCD) on the relevant rational surface. ASDEX Upgrade is making a large effort to develop, operate and evaluate an ECCD based, generic solution to MHD control, easily portable to new devices like ITER. A number of real-time diagnostics and intelligent controllers achieve precise control of ECCD deposition inside the O-point, thus minimizing the power required for NTM stabilization. For robustly operating our feedback loop, proper integration into and coordination of several diagnostic systems with the discharge control system (DCS) was essential, hence a reliable framework was designed, developed and tested. The relevant cooperating diagnostic systems are “Equilibrium”, “ECE/Mirnov” and “rt-TORBEAM” with “ECRH” as the main actuator, centrally coordinated by DCS at timescales faster than typical current diffusion times (AUG: tau approx. 150 ms). All subsystems adhere to cycle times of 15 ms (rt-TORBEAM) or less than 5 ms (all others), such that the full control loop can be executed every 20 ms, making sophisticated experiments possible. Using the feedback loop, we can reliably target and – provided sufficient ECCD power is applied – stabilize NTMs triggered at beta_N larger than 2. The system can target either 2/1 or 3/2 NTMs, whichever is detected to have the higher amplitude. The amplitude can also be used to refine the target by searching for the deposition location which minimizes the amplitude. To avoid a mode developing at all, preemptive stabilisation can be performed by tracking rational surfaces from a real-time equilibrium. In our experiments we are developing and testing a variety of control strategies, to preempt and/or stabilize existing NTMs. We have started to integrate ECCD based control of other MHD instabilities, like sawteeth, into the control scheme, thus capitalizing on the generic approach of mapping all relevant information into the same reference coordinate system given by the real-time magnetic equilibrium. This strategy can directly be adopted by next step machines such as ITER.
        Speaker: Dr Matthias Reich (Max-Planck-Institut für Plasmaphysik, Garching, Germany)
        summary
      • 08:30
        Redistribution of Energetic Particles Due to Internal Kink Modes 4h
        The internal kink modes associated with sawtooth oscillations can produce a redistribution of the energetic particles population, thus modifying the power deposition profile and increasing particle losses and wall loading. We study the effect of internal kink modes on the confinement of alpha particles and neutral beam ions by following the trajectories of a large number of particles in the total electric and magnetic fields, sum of the equilibrium plus the perturbation. The equilibrium is a simple analytical solution of the Grad-Shafranov equation with ITER like parameters and q_axis less than 1. To calculate the perturbed fields we use the experimental information regarding the space and time dependence of the displacement eigenfunctions corresponding to the modes condsidered and ideal MHD. A redistribution parameter is introduced to quantify the displacement of the particles from their initial flux surface. The effect of the (1,1), (2,2) and (2,1) modes is studied for different particle energies and mode frequency and amplitude. The results show that, for energies below 1 MeV, the redistribution can have a strong dependence on the particle energy and mode frequency.
        Speaker: Dr Ricardo Farengo (Comision Nacional de Energia Atomica)
        summary
      • 08:30
        Self Organization of High Beta_p Plasma Equilibrium with an Inboard Poloidal Null Sustained by Fully Non-Inductive Current Drive in QUEST 4h
        There is a considerable interest in operating tokamak at high value of beta_p , but maximum attainable beta_p is limited by a equilibrium limit with appearance of a null point at the inboard side. Such configuration is realized transiently earlier by Electron Cyclotron Waves, but in QUEST such equilibrium is stably produced in steady state and its equilibrium properties are investigated. In QUEST, successful production of high beta_p plasma (beta_p > 1) and its long pulse sustainment by fully non-inductive (NI) current drive with the help of a modest power (< 100 kW) ECW is demonstrated. High beta_p plasma is formed by confining energetic electrons produced by multiple resonant EC interaction in a high magnetic mirror configuration and high Bz/Bt > 0.1. We found that (i) high beta_p plasma is naturally self organized to form a stable natural Inboard Poloidal field Null (IPN) equilibrium, (ii) a critical beta_p, which defines the transition boundary from Inboard Limiter (IL) to IPN equilibrium and (iii) a new feature of plasma self organization to enhance its negative triangular shape to sustain high beta_p. With high beta_p formation, plasma naturally self organizes itself to reduce the elongation as first observed in TFTR, which is also observed in the present case. However, we found a new self organization feature, where plasma shape adjusts itself to becomes more negatively triangular. This new feature overcompensates the diminution of beta_p due to the reduction in elongation. A simple analytic solution of Grad-Shafranov equation is applied to investigate such aspect. The model supports this facet which, predicts higher beta_p at larger negative triangularity. The boundary flux surfaces generated through the model agree well with the measurements. The model is also in agreement with the critical beta_p for IL to IPN transition, which is very well matched with the measurements. This result shows a relatively simple method to produce and sustain high beta_p plasma close to the equilibrium limit in a stable configuration exploiting its self organization property.
        Speaker: Mr kishore mishra (Kyushu University)
        Poster
        summary
      • 08:30
        Simulation Study of a New Kind of Energetic Particle Driven Geodesic Acoustic Mode 4h
        A new kind of energetic particle driven geodesic acoustic mode (EGAM), which has weak bulk plasma temperature dependence of frequency, has been found in the Large Helical Device (LHD) experiments. In this work, the new kind of EGAM is investigated with a hybrid simulation code for energetic particles and magnetohydrodynamics (MHD). It is demonstrated that the new EGAM in the simulation results has weak bulk plasma temperature dependence of frequency, which is in contrast to the traditional EGAM whose frequency is proportional to the square root of bulk plasma temperature. Three conditions are found to be important for the transition from the traditional EGAM to the new EGAM: 1) energetic particle pressure substantially higher than the bulk plasma pressure, 2) charge exchange time (tau_cx) sufficiently shorter than the slowing down time (tau_s) to create a bump-on-tail type distribution, and 3) bulk plasma density is low enough. The energetic-particle distribution function is characterized by the tau_s = 8 s and tau_cx < 1 s. The energetic ion inertia term is added into the MHD momentum equation to simulate with energetic particle density comparable to the bulk plasma density. In addition, a Gaussian-type pitch angle distribution is assumed for the energetic ions. Linear growth properties of the new EGAM are investigated. It is found that the new EGAM frequency increases as the central value of the Gaussian pitch angle distribution decreases, where smaller pitch angle variable corresponds to higher parallel velocity and higher transit frequency. This indicates that the frequency of new EGAM is significantly affected by the energetic particle transit frequency, and the new EGAM is a kind of energetic particle mode (EPM) whose frequency is determined by the energetic particles. Furthermore, the frequency depends on energetic particle beta value (beta_h) and tau_cx. Growth rate of new EGAM increases as beta_h increases similarly with other energetic particle driven instabilities, but the frequency increases as beta_h increases. For higher beta_h, the effect of energetic particles is enhanced to make the frequency closer to the energetic-particle transit frequency. Moreover, shorter tau_cx causes higher growth rate and frequency, because more particles exist in the high-energy region of phase space.
        Speaker: Dr Hao WANG (National Institute for Fusion Science)
        summary
      • 08:30
        Solid Tungsten Divertor-III for ASDEX Upgrade and Contributions to ITER 4h
        AUG became a full tungsten experiment in 2007. At this time all plasma facing components have been coated with tungsten. To overcome the disadvantages of the coating - i.e. delamination of thick coatings, fast erosion of thin coatings in particular in the high heat load regime – we started to prepare a new outer divertor with solid tungsten at the outer strike line in 2010. The Div-III design was verified by extensive FEM calculations and high heat load testing of the target and its clamping structure in the test facility GLADIS. The Div-III concept was approved early in 2012 and the new divertor Div-III was installed in 2013. The redesign of the outer divertor geometry was a chance to increase the pumping efficiency in the lower divertor by increasing the gap between divertor and vessel. This increases the conductance between roof baffle and cryo-pump that is located behind the outer divertor. We expect that this results in a lower collisionality in the outer scrape-off layer and consequently in a better overlap between AUG, JET and ITER SOL parameters. To keep the option for operation with high SOL densities, a by-pass valve was placed into the cryo-pump allowing to operate AUG with full or 1/3 of the pumping speed. Safe divertor operation and heat removal becomes more and more significant for future fusion devices. This requires to develop ‘tools’ for divertor heat load control and to optimize divertor technology and geometry. Whereas the heat load receiving capability of target concepts can be tested in high heat load test facilities such as GLADIS in Garching, the target behaviour under plasma conditions has to be investigated in a fusion experiment. Here, the new divertor manipulator, DIM-II, offers a bunch of possibilities. DIM-II allows to retract a two target wide part of the divertor into a target exchange box without venting AUG. Different ‘front ends’ can be installed and exposed to the plasma. At present, front ends for probe exposition, gas puffing, electrical probes and actively cooled prototype targets are under construction. The installation of solid tungsten, the control of the pumping speed and the flexibility for divertor modifications on a weekly base is a unique feature of AUG and offers together with the extended set of diagnostics the possibility to investigate dedicated questions for a future divertor design.
        Speaker: Dr Albrecht Herrmann (IPP-Garching)
        summary
      • 08:30
        Studies of Magnetic Perturbations in High-Confinement Mode Plasmas in ASDEX Upgrade 4h
        ASDEX Upgrade is equipped with two rows of in-vessel saddle coils for magnetic perturbations with toroidal mode numbers up to n=4. A reliable ELM mitigation regime has been found in which large type-I ELMs are suppressed, and replaced by a small form of ELMs with significantly reduced energy loss from the plasma and heat load to the divertor. This regime is accessible at high pedestal plasma density, typically n_{e,ped} > 65% n_GW, and high pedestal collisionality, nu*>1.2. The access conditions to this regime are studied in more detail in experiments with gas fueling ramps and coil current ramps. The effect of magnetic perturbations on H-mode plasmas with low pedestal collisionality, $\nu^* < 0.5$, is studied in discharges without gas puff fueling. Accumulation of tungsten impurities, which can occur if eroded material from the fully tungsten-cladded wall penetrates into the plasma and is transported radially inward there, is avoided by strong central wave heating and large wall clearance. With neutral beam injection in opposite direction to the plasma current (counter-injection), complete ELM suppression is obtained during brief periods, reminiscent of Quiescent H-mode (QH) plasmas. However, the Edge Harmonic Oscillation (EHO) characteristic for QH plasmas is not observed. With co-injection, a systematic study has been undertaken to vary the conditions for penetration of the perturbation field into the plasma core: 1. Field-alignment of the magnetic perturbation (resonant or non-resonant perturbation), 2. Variation of plasma rotation and hence, perpendicular electron fluid flow velocity.For conditions with minimal field shielding, i.e. non-resonant field and vanishing perpendicular electron flow, significant rotation braking is caused by JXB torque from the m/n=1/1 sawtooth pre-cursor, as demonstrated by a perturbation coil current modulation experiment. Under the same conditions, JXB torque from rotating neo-classical tearing modes is observed. In plasmas with small field shielding, a reduction of type-I ELM losses is often observed, along with a reduction of pedestal density (often dubbed "pump-out"). Further experimentation in near future aims to diagnose the field perturbation using rotating magnetic perturbations and to study parameter dependencies of perturbation effects in these scenarios.
        Speaker: Dr Wolfgang Suttrop (Max-Planck-Institut für Plasmaphysik)
        summary
      • 08:30
        Study of GAM Radial Structure and Properties in OH and ECRH Plasmas in the T-10 Tokamak 4h
        Zonal flows and their high-frequency counterpart, the Geodesic Acoustic Modes (GAMs) are considered as a possible mechanism of the plasma turbulence self-regulation. The paper presents the results of the systematic study of GAM properties in the T-10 tokamak with heavy ion beam probe (HIBP) in the core and with multipin Langmuir probes in the edge. It was shown that GAM has radially homogeneous structure of the global eigenmode. The radial distribution of GAM frequency f_GAM is almost uniform, in spite of the temperature dependence on the radius. However, f_GAM grows with the radially averaged electron temperature approximately as T_e^1/2. The GAM amplitude also shows a tendency to be almost constant over the whole observed radial area in the plasma. The typical amplitude of GAM potential oscillations is ~ 20-80 Volts on the background of steady state values of potential up to -1 kV. GAMs are more pronounced during ECRH, when they have the main peak at 22-27 kHz and the higher frequency satellite peak at 25-30 kHz. GAM characteristics and limits of GAM existence were investigated as functions of density, magnetic field, safety factor and ECRH power. It was found that GAMs are suppressed with the density increase. The phase shift between the oscillations of potential and density was about π/2. The poloidal mode number for GAM associated potential perturbation is estimated as m=0 for the whole observed radial interval. The constancy of the GAM frequency with radius is in agreement with recent theoretical findings, which predicts that global GAM (GGAM) can exist in typical tokamak discharges with positive magnetic shear and monotonic temperature profiles. Eigenfrequency of GGAM is constant over the radius in contrast with frequency of GAM continuum.
        Speaker: Dr Alexander Melnikov (NRC 'Kurchatov Institute')
        summary
      • 08:30
        Study of ITB Formation, Electron Heat and Density Flux Structure in New ECRH/ECCD Experiments at T-10 Tokamak 4h
        In the present report, we focus at the analysis of four transport processes in T-10. First, we analyze inward electron heat pulse propagation (HPP) created by switch-on of additional off-axis ECRH on a sawteeth-free background sustained by off-axis ECRH. The presence of slow and diffusive inward HPP with “dynamic” hi-e value close to power balance value shows that the so-called “heat pinch” is either absent or very small. Second, analysis of sawtooth density oscillations in the regimes with central ECRH and in the experiments with ECCD current drive to damp the sawteeth oscillations (PECRH <0.7 MW), shows that the electron pinch velocity value is close to the neoclassical one in the plasma centre. Under PECRH > ~ 1.5 MW, ne profiles become hollow (or fully flat within the errorbars) at r/a<0.5 and we observe sawteeth density oscillations with inverted phase (rise at r=+/- 4cm and decay at r=+/-12cm). The decay of ne in the centre between the crashes is explained by the presence of the outward electron convective velocity with V p ≈ 0.3 m/s at r/a= 0.35. Third, a set of experiments with programmed plasma motion allows us to analyze fine detail of Te profile with ECE measurements. No signs of clear ITB at the q=1 surface have been observed so far (PECRH up to 0.9 MW). In several shots, the existence of a narrow ITB with a 0.5 cm width and a doubled Te gradient can be suggested within the erorbars. Finally, a new type of ITB created by sawteeth oscillations almost damped by off-axis ECCD has been found. A sawteeth crash causes the rise of Te outside rs and heat pulse does not propagate outside during 15 ms. The value of χe becomes 2.5 times lower compared with the L-mode scaling. The experiments with various values of current generation and reflectometer measurements are under the way.
        Speaker: Dr Sergey Neudatchin (IFT, NRC Kurchatov Institute)
        summary
      • 08:30
        The Auxiliary Heating and Current Drive Systems on The Tokamak T-15 Upgrade 4h
        The auxiliary Heating and Current Drive systems of the T-15upgrade tokamak are presented. The NBI system will consist of three hydrogen NB injectors 6 MW total power at pulse duration up to 30 s. The ECRH and CD system will consist of 7 gyrotrons 6 MW total RF launched power at pulse duration up to 30 s. The LHH and CD system will be able to launch 4 MW RF power with 30 s duration.
        Speaker: Dr Igor Roy (Institute of Tokamak Physics, National Research Centre Kurchatov Institute)
        summary
      • 08:30
        The First Lower Hybrid Current Drive Experiments in the Spherical Tokamak Globus-M 4h
        The development of quasi-stationary methods of non-inductive current drive in plasmas of spherical tokamaks is extremely important for their using in thermonuclear devices. Specific properties of spherical tokamaks provide the possibilities the current drive by the waves of intermediate frequency range slowed down in poloidal direction. This approach, developed theoretically in the Ioffe Physical-technical institute and described in [1], is used now in LHCD experiments on the low aspect ratio tokamak Globus-M (R=0.36 m, a0=0.24 m, B0=0.4 T, Ip=0.2MA, elongation – 1.8, triangularity – (0.4-0.5), n0 = (3-5)•1019m-3, nb=1•1017m-3, Te0=400-800 eV). The grill consists of 10 waveguides with inner cross-section 90x10mm. The waveguides are oriented so that the electric RF fields on the fundamental waveguide mode TE10 were co-directed with poloidal direction in the tokamak. Numerical simulation confirms the quite good efficiency of current drive. The first experimental results are presented for input power level of 100 kW (2.45 GHz).The value of generated current is estimated by the drop of loop voltage as 20-40 kA. [1] E.Z. GUSAKOV, V.V.DYACHENKO et al. PPCF, 52, (2010), 075018
        Speaker: Mr Valeriy Dyachenko (A.F. Ioffe Physico-Technical Institute)
        summary
      • 08:30
        The Isotope Effect in the RFX-Mod Experiment 4h
        The isotope effect, namely the dependence of plasma confinement on the mass Mi of majority ion, is a well known property of tokamak configuration. Increasing Mi leads to an improvement of energy, particle and momentum confinement in all regimes of tokamak plasmas. Besides, Mi influences also many MHD roperties, e.g. increasing the period of plasma instabilities. However in stellarators the confinement properties are independent on Mi. Despite a strong research effort an explanation of the isotope effect in tokamaks is still lacking. During the past year the Reversed Field Pinch (RFP) device RFX-mod started to operate using Deuterium (D), besides Hydrogen (H), as filling gas. In this paper we present first results onthe comparison among Hydrogen and Deuterium plasmas of RFX-mod, offering the opportunity of studying the isotope effect physics from a new perspective First analyses of Deuterium plasmas show clearly the presence of an isotope effect also in RFP configuration. The plasma properties change with Mi in a way which reminds what happens in tokamaks. The electron temperature in D plasmas is about 20% higher than in H ones. This increase is essentially due to the steepening of Te gradient in the external region of plasma (r/a > 0.7), while gradients in the plasma core does not undergo a significant modification. Discharges with similar plasma parameters are characterized by influxes of majority ion 30% lower in D plasmas than in H ones. Interestingly no significant difference is seen in the impurity influxes. The mass Mi influences also the MHD properties of plasmas. At high current (Ip > 1 MA) the plasma is in the Quasi Single Helicity (QSH) state where a single MHD instability, the dominant one, overcomes the others, the so-called 'secondary' modes. The QSH phases are transiently interrupted by burst of MHD activity, the Dynamo Relaxation Events (DRE). The duration of QSH phases increases by a factor ~1.5 changing the main gas from H to D, resulting in a longer time interval among DREs responsible of QSH collapse. The energy of secondary modes during QSH is about 20% lower in D. Since the amplitude of dominant mode does not exhibit a significant variation, QSH are purer in D than in H. Furthermore the comparison among D and H plasmas shows that the current profile in D plasmas is more peaked than in H.
        Speaker: Dr Rita Lorenzini (Consorzio RFX, Corso Stati Uniti 4, 35127, Padova, Italy)
        summary
      • 08:30
        Transport Theory for Energetic Alpha Particles and Superbananas in Tokamak Fusion Reactors with Broken Symmetry 4h
        Error fields or magnetohydrodynamic (MHD) activities break toroidal symmetry in real tokamak fusion reactors, e.g., ITER. It has been shown in a comprehensive theory for neoclassical toroidal plasma viscosity for tokamaks that broken symmetry enhances particle, momentum, and energy transport. The enhanced energy transport increases with increasing energy. Because fusion-born alpha particles have energy significantly higher than that of fuel ions, the enhanced energy loss for alpha particles can be an issue for reactors if their energy transport loss rate is faster than the slowing down rate. In that case, the fusion energy gain factor Q will be significantly impacted. To quantify the tolerable magnitude for the error fields or MHD activities, transport theory for energetic alpha particles is developed. The theory is a generalization of the theory for neoclassical toroidal plasma viscosity. The superbanana plateau and superbanana regimes are the most relevant regimes for fusion-born alpha particles. The transport theory for energetic alpha particles developed is extended to the limit where the slowing down operator dominates and to allow for the finite value of the radial electric field.
        Speaker: Dr Ker-Chung Shaing (University of Wisconsin)
        summary
      • 08:30
        Turbulent Electromagnetic Filaments in Toroidal Plasma Edge 4h
        Filament or blob structures have been observed in all magnetic configurations with very similar features despite the difference in the magnetic geometry: theory and experiments suggest they exhibit a radial convective motion across the SOL, and the interest in blob dynamics is further motivated by their interaction with first wall and divertor. Despite their different generation mechanism, turbulent structures and Edge Localized Mode (ELM) filaments share some common physical features, as the localization in the cross-field plane and the associated parallel current, with a convective radial velocity component somehow related to their dimension. The electromagnetic effects on filament structures deserve particular interest, among the others for the implication they could have for ELM, related for instance to their dynamics in the transition region between closed and open field lines or to the possibility, at high beta regimes, of causing line bending which could enhance the interaction of blobs with the first wall. Electromagnetic features of turbulent filaments, emerging from turbulent background, will be shown in four different magnetic configurations: the stellarator TJ-II, the Reversed Field Pinch RFX-mod, a device that can be operated also as a ohmic tokamak, and the Simple Magnetized Torus TORPEX. In all cases, direct measurements of both field-aligned current density and vorticity were performed inside the filament. Despite the great specific differences, the inter-machine comparison revealed a clear dependence of the filament vorticity upon the local time-averaged ExB flow shear. Furthermore the wide range of local beta that was explored through the four mentioned configurations allows concluding that this parameter plays a fundamental role in the appearance of the electromagnetic features of filaments, suggesting an underlying common physics. The RFX-mod experiment versatility is exploited also from the point of view of the active control of the edge magnetic topology focusing on the filament interaction with local magnetic island. High frequency fluctuations, characterizing electrostatic and magnetic filament features, have been observed to be affected by the island proximity. This observation hints at the challenging possibility of active control of filaments and their related transport by modulating the local magnetic topology.
        Speaker: Dr Monica Spolaore (Consorzio RFX, Padova, Italy)
        summary
      • 08:30
        W Impurity Poloidal Assymmetries Observed at ASDEX Upgrade Using Soft-X-Ray Tomography Reconstruction 4h
        Many tokamaks are nowadays equipped with metallic walls. The positive effect of such modification compared to the previous carbon walls is a strong reduction of the tokamak wall erosion and tritium retention. But the drawback is a production and potential accumulation of heavy impurities in the plasma core which can cause high radiation losses and even trigger radiative collapses often leading to disruptions. Poloidal inhomogeneities of impurity distribution have a significant impact on radial impurity transport [1]. Understanding of the mechanisms leading to inhomogeneous impurity distribution is thus useful knowledge for the control of heavy impurity transport and hopefully accumulation avoidance. Poloidal asymmetries due to centrifugal effects or/and other sources of an equilibrium poloidal electric field in the core plasma, as those generated by minority Ion Cyclotron Resonance Heating (ICRH), are usually negligible in experimental impurity transport analysis on low-medium Z impurities. Such assumptions are not always valid for medium-high Z impurities. Poloidal asymmetries such as those generated indirectly by minority ICRH heating or Neutral Beam Injection (NBI) have been observed and analyzed, using SXR tomographic reconstructions, during recent ASDEX Upgrade experimental campaigns. Trace injections of tungsten have been triggered by Laser Blow Off (LBO) ablation in different scenarios with fixed plasma current. Scans in NBI and ICRH power (H-minority heating scheme) and deposition location (change in frequency and/or toroidal field) have been performed in order to study their effects on Low Field Side (LFS) - High Field Side (HFS) asymmetries. Analysis of the obtained results is presented, focusing in particular on the effects of each actuator. Centrifugal effects have been recently implemented in the code GKW [2] for the description of turbulent impurity transport. The Hinton Wong neoclassical theory in the presence of rotation, implemented in the NEO code, [3], can describe the impact of rotational effect on neoclassical transport. These new theoretical tools with the RF asymmetries now being implemented allow complete transport modeling of poloidal asymmetries. [1] F.J. Casson et al, Phys. Plasmas 17, 102305 (2010) [2] C. Angioni et al, Phys. Plasmas 19, 122311 (2012) [3] E. Belli et al, Plasma Phys. Cont. Fusion 54 015015 (2012)
        Speaker: Dr Didier Mazon (CEA Cadarache)
        summary
    • 10:15 10:45
      Coffee Break 30m
    • 10:45 12:30
      3D Physics: EX/1 Blue 1-5

      Blue 1-5

      Hotel Park Inn Pribaltiyskaya

      Saint Petersburg, Russian Federation
      Convener: Mr Francois Waelbroeck (USA)
      • 10:45
        Recent Advances in the Understanding and Optimization of RMP ELM Suppression for ITER 20m
        Recent experiments with applied Resonant Magnetic Perturbations (RMPs) in low-collisionality ITER Similar Shape (ISS) plasmas on DIII-D have advanced the understanding of and increased confidence in obtaining ELM suppression in the ITER standard operating regime. ELM suppression is obtained with a reduced coil set (5–11 coils) on DIII-D, demonstrating the effectiveness of mixed harmonics (n=1,2,3) with a partial coil set and mitigating the risk of reduced coil availability on ITER. Recent advances in linear two-fluid MHD simulations indicate that resonant field penetration and amplification at the top of the pedestal is ubiquitous in these ISS plasmas, together with resonant field screening and kink amplification in the steep pressure gradient region. Measurements with the X-ray imaging camera reveal new information on the plasma response to 3D fields. There is good agreement between X-ray imaging and M3D-C1 simulation in the steep pressure gradient region of the pedestal, validating theoretical predictions of resonant screening and a dominant edge-kink response. While direct imaging of islands in ELM suppressed plasmas remains elusive, measurements with the newly upgraded magnetic sensors are suggestive of partially screened fields at the top of the pedestal, consistent with M3D-C1 simulations. Indirect evidence of island formation and resonant field penetration is also provided by the observed flattening of the electron pressure profile at the top of the pedestal and concomitant shrinkage of the pedestal width when the RMP is applied. In addition, the flutter model of electron transport also predicts an electron thermal diffusivity “hill” at the top of the pedestal that is comparable to experimental values when the resonant field amplification at the top of the pedestal is included in the calculation. Optimization of the pedestal pressure is an important issue for ELM suppression in ITER given that a reduction in the pedestal pressure is commonly observed in ISS plasmas with applied RMPs. Recent experiments demonstrate that the pedestal pressure can be maintained at the level before the onset of the RMP if the effect of density pumpout is counteracted with density feedback. This work was supported by the US Department of Energy under DE-AC02-09CH11466, DE-FG02-92ER54139, DE-FC02-04ER54698, DE-SC0007880, DE-FG02-07ER54917, and DE-AC05-00OR22725
        Speaker: Mr Mickey R. Wade (USA)
        Slides
      • 11:05
        Effect of Resonant Magnetic Perturbations on Low Collisionality Discharges in MAST and a Comparison with ASDEX Upgrade 20m
        The application of Resonant Magnetic Perturbations (RMPs) is foreseen as a mechanism to ameliorate the effects of ELMs on the ITER divertor. Various aspects of RMP operation crucial to ITER have been demonstrated on MAST such as mitigating the first ELM after the L-H transition, sustaining ELM mitigation during both the current ramp-up and in the event of failure of a sub-set of the in-vessel coils and applying a slowly rotating n=3 RMP, which sustains ELM mitigation while rotating the pattern of the strike point splitting. Although ELM suppression has not been observed on MAST, ELM mitigation has been achieved using RMPs with toroidal mode number of n=2, 3, 4 and 6 over a wide region of operational space, with considerable overlap with the regions where suppression of type-I ELMs is observed in other machines. The effect that the choice of toroidal mode number on the effectiveness of the mitigation has been investigated and shows that n=3 or 4 is optimal. The ELM mitigation phase is typically associated with a drop in plasma density and overall stored energy. By carefully adjusting the refuelling, either by gas or pellet fuelling, to counteract the drop in density it has been possible to produce plasmas with mitigated ELMs, reduced peak divertor heat flux and with minimal degradation in pedestal height and confined energy. Above a threshold value in the applied perturbation field (brres) there is a linear increase in normalised ELM frequency (fELM) with brres. Experimentally it has been found that both the lobes produced near the X-point and the mid-plane corrugations also increase linearly with the size of brres. These deformations to the plasma boundary have been replicated by modelling, which shows that they can strongly influence the peeling-ballooning stability boundary and hence lead to an increase in fELM. Mitigation of type I ELMs has also been achieved on ASDEX Upgrade at mid-low collisionalities, using RMPs with n=1 and 2. In a large number of cases an increase of fELM with brres is also observed. However, unlike in MAST, there are some cases where this is not the case. This presentation will compare and contrast the results from the two devices with an aim of increasing our understanding and ability to extrapolate to future devices. This work was part-funded by the RCUK Energy Programme and by Horizon 2020 programme.
        Speaker: Mr Andrew Kirk (UK)
        Slides
        summary
      • 11:25
        Comparative Studies of Edge Magnetic Islands and Stochastic Layers in DIII-D and LHD 20m
        Joint experiments on the DIII-D tokamak and the LHD stellarator/heliotron have resulted in the discovered of spontaneous heat transport bifurcations across the O-point of an applied m/n=2/1 magnetic islands in DIII-D and enhanced particle transport relative to heat transport in edge m/n=1/1 LHD isalnds. The DIII-D results suggest that the heat transport bifurcations are due to islands transitioning from smooth flux surfaces to partially stochastic layers. Alternatively, measurements of the particle and heat transport inside edge static magnetic islands in LHD plasmas show an enhancement of the particle flux relative to the heat flux. The DIII-D results suggest that externally applied static 3D magnetic fields can produce a dynamic evolution of the magnetic topology in the plasma due to a nonlinear toroidal coupling of resonant modes on various rational surfaces while the LHD results show that edge magnetic islands preferentially increase the particle flux relative to the heat flux for reasons that have yet to be clarified. Static magnetic islands and stochastic layers have been observed in low-β L-mode plasmas but not in diverted H-mode plasmas yet. In order to understand the physics of non-axisymmetrically perturbed high-β fusion H-mode plasmas, such as the mechanisms involved in edge localized mode (ELM) suppression with resonant magnetic perturbation (RMP) fields, it is necessary to determine if islands and stochastic layers exists and whether they are static or evolve in time due to the plasma response. Measurements in DIII-D show spontaneous transitions of the magnetic field on rational surfaces due to the plasma response. For example, flat spots in the T_e profile associated with m/n = 2/1, 3/1 and 4/1 islands are seen to appear and disappear as the discharge evolves suggesting either a time varying screening of the field by the plasma or a nonlinear coupling of the 2/1, 3/1 and 4/1 islands. In this contribution we discuss the measurements made in DIII-D along with transport results due to pellets injected into static islands in LHD and their implications for understaning the plasma response to 3D fields in H-mode plasmas. This work was supported in part by the US Department of Energy under DE-FC02-04ER54698, DE-AC05-00OR22725, DE-FG03-97ER54415, and the NIFS budget code NIFS11ULHH021.
        Speaker: Mr Todd E. Evans (USA)
        Slides
        summary
      • 11:45
        Physical Characteristics of Neoclassical Toroidal Viscosity in Tokamaks for Rotation Control and the Evaluation of Plasma Response 20m
        Favorable use of low magnitude (deltaB/B ~ O(10^-3)) three-dimensional (3D) magnetic fields in tokamaks includes mitigation of ELMs and Alfvénic modes, and alteration of the plasma rotation profile to strongly affect the stability of NTMs and RWMs. However, in ITER, these fields can significantly reduce the fusion gain, Q, by increasing alpha particle transport. These effects have been theoretically addressed using neoclassical toroidal viscosity (NTV) theory [K.C. Shaing and C.T. Hsu, Nucl. Fusion 54 (2014) 033012]. NTV magnitude and profile that determines the critical 3D applied field level for Q reduction, or for rotation feedback control, depends on the field spectrum, plasma collisionality, and plasma response to the field. The present work focuses on these critical questions with new analysis of results from NSTX and KSTAR. Experimental angular momentum alteration is directly compared to theoretical NTV torque density profiles, T_NTV, created by a range of applied 3D field spectra and plasma parameters in NSTX including configurations with dominant n = 2 and n = 3 field components. Large radial variations of T_NTV are found in ideal MHD models when the flux surface displacement is derived using an assumption of a fully penetrated deltaB. In contrast, experimentally measured T_NTV does not show strong torque localization. NSTX experiments yield a computed displacement ~ 0.3cm, smaller than the ion banana width, and averaging T_NTV over the banana width more closely matches the measured dL/dt profile. Results from a model-based rotation controller designed using NTV from applied 3D fields as an actuator for instability control are shown. A favorable observation for rotation control, clearly illustrated by KSTAR experiments, is the lack of hysteresis of the rotation when altered by non-resonant NTV. These experiments also show the theoretical scaling of T_NTV with deltaB^2 and ion temperature ~ T_i^2.5. Due to this strong dependence of the T_NTV profile on deltaB, the T_NTV measurements significantly constrain the allowable field amplification. Plasma response models being tested against experiment include the fully-penetrated deltaB model, and various physics models in the M3D-C1 resistive MHD code. Analysis shows that the M3D-C1 single-fluid model produces a flux surface-averaged |deltaB| consistent with the measured T_NTV.
        Speaker: Mr Steven A. Sabbagh (USA)
        Slides
        summary
      • 12:05
        Successful ELM Suppressions in a Wide Range of q95 Using Low n RMPs in KSTAR and its Understanding as a Secondary Effect of RMP 20m
        As the most plausible technique to control the edge localized modes (ELMs) of high confinement mode (h-mode) plasmas, which is critical for ITER and beyond, non-axisymmetric resonant magnetic perturbations (RMPs) have been actively investigated in KSTAR. Since the first success of ELM suppression using n=1 RMPs in 2011 [1], our effort has been devoted to extend the operation regime of ELM control for both external magnetic configurations and plasma parameters. As results, it shows a possibility that the ELM-suppression can be achieved in a wide range of q95 (=3.5~6.5) if the RMP field is configured properly. Several unique features of RMP-driven ELM-suppression are found, by which the physics mechanisms of ELM suppression and mitigation can be explicitly distinguished. These are a long time delay (a secondary effect) and an improved confinement (transport bifurcation). Furthermore, a persistent, rapidly repeating bursty event, localized in the plasma edge, is observed and suggested as a key player in the underlying physics mechanism of RMP-driven ELM-suppression. [1] Y.M. Jeon et al., Phys. Rev. Lett. 109, 035004 (2012)
        Speaker: Mr YoungMu Jeon (Republic of Korea)
        Slides
        summary
    • 12:30 14:00
      Lunch Break 1h 30m
    • 14:00 16:10
      Overview 4: Magnetic Fusion: OV/4 Blue 1-5

      Blue 1-5

      Hotel Park Inn Pribaltiyskaya

      Saint Petersburg, Russian Federation
      Convener: Mr Myeun Kwon (Korea, Republic of)
      • 14:00
        Overview of HL-2A Recent Experiments 25m
        Since the last IAEA FEC, experiments on HL-2A tokamak have been dedicated to address the physics on L-H transition, energetic-particles (EPs) and shear Alfvén waves (SAWs), ELM mitigation, disruption mitigation, edge impurity transport and other MHD related activities. In particular, significant progresses have been made in the following areas: (i) For the first time in experiments, it was found that the phase between normalized radial electric field and the envelope of density fluctuations reverses during the intermediate phase (I-phase) in comparison to the usual predator-prey regime as the plasma approaches H-mode during the L-I-H confinement transition; (ii) The frequency up- and down-sweeping reverse shear Alfvén eigenmodes (RSAEs) were observed in NBI plasmas with qmin~1. By using kinetic AE code (KAEC) simulation, it has been confirmed that the down-sweeping modes are kinetic RSAEs, and the up-sweeping modes are RSAEs that exist in the ideal or kinetic MHD limit; (iii) The transition and interaction among low-frequency MHD modes have been observed during NBI, which suggests profound interaction existing among fishbone mode, long-live mode (LLM) and tearing mode (TM); (iv) With SMBI a runaway electron plateau was observed at a rather low toroidal magnetic field. In addition, progresses have also been made in the analysis of the loss of energetic-ions, the ELM mitigation induced by supersonic molecular beam injection (SMBI), and the onset of neoclassical tearing mode (NTM) during nonlocal effect with SMBI, etc. All these experiments benefited from several newly installed diagnostics, such as Motional Stark Effect (MSE), Charge Exchange Recombination Spectroscopy (CXRS) and a scintillator-based lost fast-ion probe (SLIP), and the upgrade of ECRH heating power to 5 MW.
        Speaker: Mr Min Xu (China)
        Slides
        summary
      • 14:25
        The Science Program of the TCV Tokamak: Exploring Fusion Reactor and Power Plant Concepts 25m
        TCV is acquiring a new 1 MW neutral beam and 2 MW additional third-harmonic ECRH to expand its operational range. Its existing shaping and ECRH launching versatility was amply exploited in an eclectic 2013 campaign. A new high-confinement mode (IN-mode) was found with an edge barrier in density but not in temperature. Density limits close to the Greenwald value were reached – at reduced confinement - by sawtooth regularization with ECRH. The edge gradients were found to be regime-dependent and to govern the scaling of confinement with current. A new theory predicting a toroidal rotation component at the plasma edge, driven by inhomogeneous transport and geodesic curvature, was tested with promising results. The L-H threshold power was measured to be 15-20% higher in both H and He than D, to increase with the length of the outer separatrix, and to be independent of the current ramp rate. Core turbulence was found to decrease from positive to negative edge triangularity deep into the core, consistent with global confinement increase. The geodesic-acoustic mode was studied with multiple diagnostics, and its axisymmetry was confirmed by a full toroidal mapping of its magnetic component. The heat flux profile decay length and heat load profile on the wall were documented as functions of plasma shape in limited plasmas. In the snowflake (SF) divertor configuration we have documented the heat flux profiles on all four strike points. SF simulations with the 3D EMC3-Eirene code, including the physics of the secondary separatrix, underestimate the flux to the secondary strike points, possibly resulting from steady-state ExB drifts. With neon injection, radiation in a SF was 15% higher than in a conventional divertor. The novel triple-X and X-divertor configurations were achieved transiently in TCV. A new sub-ms real-time equilibrium reconstruction code was used in ECRH control of NTMs and in a prototype shape controller. The detection of visible light from the plasma boundary was also successfully used in a position-control algorithm. A new bang-bang controller improved stability against vertical displacements. The RAPTOR real-time transport simulator was applied to current density profile control experiments with ECCD. Shot-by-shot internal inductance optimization was demonstrated by iterative learning control of the current reference trace.
        Speaker: Mr Stefano Coda (Switzerland)
        Slides
        summary
      • 14:50
        Overview of Recent Physics Results from NSTX 25m
        NSTX is currently being upgraded to operate at twice the toroidal field and plasma current (up to 1 T and 2 MA), with a second, more tangentially aimed neutral beam for current and rotation control, allowing for pulse lengths up to 5 s. Recent NSTX physics analyses have addressed topics that will allow NSTX-U to achieve the research goals critical to a Fusion Nuclear Science Facility. These goals include accessing low collisionality and high beta, producing stable, 100% non-inductive operation and assessing Plasma Material Interface (PMI) solutions to handle the high heat loads expected in the next-step devices. Including rotation and kinetic resonances, which depend on collisionality, is necessary for predicting experimental stability thresholds of fast growing Ideal Wall and Resistive Wall Modes. Non-linear gyrokinetic simulations have been performed to study transport of heat, particles and momentum in the core plasma, and its dependence on collisionality and profile shapes. These studies include coupling between low- and high-k turbulence, the effect of rotation and non-local transport. PMI studies have focused on the effect of ELMs and 3D fields on plasma detachment and heat flux handling. DEGAS-2 has been used to study the dependence of gas penetration on SOL temperatures and densities for the MGI system being implemented on the Upgrade. Studies of lithium evaporation on graphite surfaces indicate that lithium increases oxygen surface concentrations on graphite, and deuterium-oxygen affinity, which increases deuterium pumping and reduces recycling. Source studies showed that the low lithium level observed in the core of lithium-coated wall NSTX plasmas was due to both high retention of lithium in the divertor as well as large neoclassical diffusivity. Noninductive operation and current profile control in NSTX-U will be facilitated by Coaxial Helicity Injection as well as RF and NB heating. CHI studies using NIMROD indicate that the reconnection process is consistent with the 2D Sweet-Parker theory. Full wave AORSA simulations show that RF power losses in the SOL increase significantly for both NSTX and NSTX-U when the launched waves propagate in the SOL. TAE avalanches can affect NB driven current through energy loss and redistribution of fast ions. Upgrade construction is moving on schedule with first operation of NSTX-U planned for Autumn 2014.
        Speaker: Mr Steven Sabbagh (USA)
        Slides
      • 15:15
        3D Effects of Edge Magnetic Field Configuration on Divertor/SOL Transport and Optimization Possibilities for a Future Reactor 25m
        This paper overviews recent progress on the experimental identification and physics interpretation of 3D effects of magnetic field geometry/topology on divertor transport in helical devices and tolamaks with RMP. The 3D effects are elucidated as a consequence of competition between transports parallel (//) and perpendicular to magnetic field, in open field lines cut by divertor plates or in magnetic islands. The competition process has strong impacts on divertor functions, density regime, impurity screening, and detachment stability. Based on experiments and numerical simulations, key parameters (indicated with [ ] below) governing the transport process are discussed suggesting demanding issues to be addressed for divertor optimization in future reactors. The divertor density regime, which is known for strong up- and down-stream coupling, high-recycling regime in 2D axi-symmetric configurations, is affected by the 3D configuration. In W7-AS, LHD, TEXTOR-DED and HSX, the dependency is weakened. This is due to enhanced loss of //-momentum or of //-conduction energy. The dependency is functions of magnetic geometry parameters, [field line connection length], [poloidal wave length of RMP] and [Br/Bt]. Impurity screening is observed in Tore Supra, LHD, TEXTOR-DED with edge stochastization, and in W7-AS/X, TJ-II with island divertor. The enhanced outward particle flux due to [Br] provides the screening via friction force exerted on impurity. It is also found that suppression of ion thermal force, in the case of small [Br/Bt]~1.e-4, is responsible for the screening. The systematic study in TEXTOR-DED and LHD has shown that [a thicker stochastic region] provides better screening effects. In W7-AS and LHD, the larger [edge island width] leads to detachment stabilization. This is due to capturing of radiation with the islands and to the decoupling of edge and core plasmas in terms of core fueling of plasma/impurity. In TEXTOR-DED, [rotating RMP] fields result in density limit extension, avoiding MARFE onset. This is caused by spreading of recycling region, preventing edge cooling localization by recycling neutral/impurity. Systematic understandings of the 3D effects of edge magnetic field based on the key parameters shown above will offer new perspectives on divertor optimization for future reactors, which are not available in 2D axi-symmetric configuration.
        Speaker: Mr Masahiro Kobayashi (Japan)
        Slides
        summary
      • 15:40
        Transport, Stability and Plasma Control Studies in the TJ-II Stellarator 25m
        Recent improvements in TJ-II plasma diagnostics and operation have led to a better understanding of transport, stability and plasma control in fusion plasmas. Impurity transport: Observations of asymmetries in impurity parallel flows in TJ-II ion-root plasmas have been interpreted as an indication of the compressible variation of the impurity flow field and hence of in-surface impurity density asymmetries. In addition, first-time observations of electrostatic potential variations within the same magnetic flux surfaces are presented which are reproduced by neoclassical Montecarlo calculations. The dependence of impurity confinement time has been also studied as a function of charge and mass. Momentum transport and isotope physics: TJ-II has provided evidence that three-dimensional magnetic structures convey significant impact on plasma confinement and L-H transitions. Recent observations on the temporal ordering of the limit cycle oscillations at the L-I-H transition show the leading role of the plasma turbulence. Comparative studies in tokamaks and stellarators have provided direct experimental evidence of the importance of multi-scale physics for unravelling the physics of the isotope effect on transport. Power exhaust physics: Novel solutions for plasma facing components based on the use of liquid metals like Li and alloys have been developed on TJ-II. The TJ-II programme on liquid metals addresses fundamental issues like the self-screening effect of liquid lithium driven by evaporation to protect plasma-facing components against heat loads and tritium inventory control, using Li-liquid limiters (LLL) recently installed. Plasma stability studies: Experiments with magnetic well scan on TJ-II suggest that stability calculations, as those presently used in the optimization criteria of stellarators, might miss some stabilization mechanisms. Fast particle control: The TJ-II results show that, upon moderate off-axis ECH power application, the continuous character of the Alfvén eigenmode (AEs) changes significantly and starts displaying frequency chirping. This result shows that ECH can be a tool for AE control that, if confirmed, could become ITER and reactor-relevant.
        Speaker: Mr Joaquín Sánchez (Spain)
        Slides
        summary
    • 14:00 18:45
      Poster 2: P2 Green 8-9

      Green 8-9

      Hotel Park Inn Pribaltiyskaya

      Saint Petersburg, Russian Federation
      • 14:00
        (N)TM Onset by Central EC Power Deposition in FTU and TCV Tokamaks 4h 45m
        The onset of both the conventional and neoclassical tearing modes (N)TMs remains an important issue for the fusion plasma operations. The understanding of the (N)TM onset driven by on-axis EC action, far from the mode locations, is a field still not well understood for the MHD instability control. Comparison of the responses from different devices with comparable size and operation parameters can give clear information about the main mechanisms leading to the mode destabilization. In L-mode the effects of central electron cyclotron heating (ECH) and current drive (ECCD) on the presence of (N)TMs have been investigated in TCV without explicit triggers as sawteeth and in FTU with the presence of latent MHD activity. In TCV two possible concomitant driving mechanisms for these instabilities, due to the on-axis EC power, have been associated to the change of plasma current density profile and of mode stability parameter upstream of the resonant location q=m/n and to the change in sign of the local difference between the toroidal plasma and the (N)TM velocity due to the EC torque, allowing the destabilizing action of the ion polarization current. In FTU the former mechanism has been related to both the mode onset and the amplification of a mode that was present in a marginally stable state. A scaling for the onset / amplification of (N)TM will be given and discussed taking into account geometrical and operational parameters. Investigation of the plasma current density evolution will be done using the current diffusion equation in transport codes as JETTO and ASTRA in order to calculate the stability parameter changes and compare this dynamics in FTU and TCV.
        Speaker: Dr Silvana Nowak (CNR)
        Poster
        summary
      • 14:00
        A Reduced Model of ELM Mitigation by SMBI and Pellet Injection 4h 45m
        The importance and urgency of ELM control for ITER urge to develop tractable reduced model to understand ELM mitigation experiments because first principle models of ELMs have not been developed due to terrific complexities of ELM-related physics. In this work, we present such an effort particularly focused on ELM mitigation experiment by supersonic molecular beam injection (SMBI) and pellet injection (PI). We employs cellular automata model (CA) including key physical elements for transport (i.e. turbulent transport, its suppression by diamagnetic shear flow, and MHD limit) in the H-mode pedestal and expand it to include the SMBI and PI. A finding is that extended CA model can capture many essential features of the experiments, and both SMBI and pellet injection can be seen as different regions of a continuous spectrum of experiments with varying amounts and penetration depths of injected material. Shallow and small injection (SMBI) can mitigate large ELMs by triggering more frequent, yet smaller scale ejection events. With larger and deeper deposition of material, injection forces the formation of pedestal pressure profiles which trigger large ELMs. Therefore, repetitive deep injection emerges as ELM pacing.
        Speaker: Dr Tongnyeol Rhee (NFRI, Daejeon, South Korea)
        Poster
        summary
      • 14:00
        Achieving Steady-State Conditions in High-Beta Hybrid Scenario in DIII-D 4h 45m
        The natural attributes of the hybrid scenario, especially the anomalously broad current profile that suppresses sawteeth by maintaining the safety factor minimum (qmin) above unity, allows steady-state conditions with zero surface loop voltage to be achieved in 1 MA discharges in DIII-D with efficient central current drive and simultaneous high beta and high confinement. Steady-state hybrid plasmas can achieve βN=3.6 for the full duration of the NB pulse (>1 τ_R) without exciting the deleterious m/n=2/1 tearing mode, corresponding to βT up to 3.4%. The experimental βN exceeds the DCON-calculated no-wall n=1 stability limit by 20%. With central current drive, the surface loop voltage is driven down to zero for >1 τ_R when the poloidal beta is increased above 1.9 by raising the EC power to 3.05 MW and reducing IP from 1.1 MA to 1.0 MA. High-β hybrids can have slightly more than 50% bootstrap current fraction despite qmin≈1; the other half of the plasma current is driven efficiently using central EC and NB current drive. Inside of the q=3/2 surface the current profile is strongly overdriven, and time dependent TRANSP modeling shows that qmin should drop to ~0.8 by the end of the discharge. The fact that qmin remains above unity and sawteeth are suppressed shows that the hybrid scenario maintains an anomalously broad current profile even in the presence of strong central current drive. The thermal energy confinement time is excellent, with confinement factors of up to H98y2=1.6 even during strong central EC heating. The experimental density and temperature profiles are well reproduced by the TGLF transport model, and measured changes in electron thermal transport, due to shape-induced pedestal changes or electron heating, parallel the predicted changes in the level of high-k turbulence. A zero-dimensional physics model demonstrates that attractive scenarios with Qfus=3.5–3.8 exist for steady-state operation in ITER and FNSF using central EC current drive, with higher Qfus possible using higher efficiency NB current drive. Therefore, high-β hybrid plasmas with central current drive should be considered as an alternative method for achieving the fusion performance goals in steady-state scenarios in ITER and FNSF. This work was supported in part by the US Department of Energy under DE-FC02-04ER54698, DE-FG02-04ER54761, DE-AC52-07N27344, and DE-FG02-08ER54984
        Speaker: Dr C. Craig Petty (General Atomics)
        summary
      • 14:00
        Advancing the Physics Basis of Quiescent H-Mode Through Exploration of ITER Relevant High Density Operation 4h 45m
        Recent experiments on DIII-D have overcome a long-standing limitation in accessing quiescent H-mode (QH-mode), a high confinement state of the plasma that does not exhibit the explosive instabilities associated with edge localized modes (ELMs). In the past, QH-mode was associated with low density operation, but has now been extended to high normalized densities compatible with operation envisioned for ITER. Through the use of strong shaping, QH-mode plasmas have been maintained at high densities, both absolute (bar{n}_e>7x10^{19}m^{-3}) and normalized Greenwald fraction bar{n}_e/n_G>0.7 . In these plasmas, the pedestal can evolve to very high pressures and current as the density is increased, becoming comparable to some of the highest performance transient pedestals seen on DIII-D. Calculations of the pedestal height and width from the EPED model are quantitatively consistent with the experimental observed evolution with density. Such comparisons of the dependence of the maximum density threshold for QH-mode with plasma shape help validate the underlying theoretical peeling-ballooning models describing ELM stability. High density QH-mode operation with strong shaping has allowed stable access to a previously predicted regime of very high pedestal dubbed “Super H-mode”. In general, QH-mode is found to achieve ELM-stable operation while maintaining adequate impurity exhaust, due to the enhanced impurity transport from an edge harmonic oscillation, thought to be a saturated kink-peeling mode driven by rotation shear. In addition, the impurity confinement time is not affected by rotation, even though the measured ExB shear is observed to increase at low toroidal rotation, resulting in reduced turbulence and increased energy confinement. Together with the simultaneous achievement of high beta, high confinement and low q95 for many energy confinement times, these results suggest QH-mode as a potentially attractive operating scenario for ITER’s Q=10 mission. This work was supported by the US Department of Energy under DE-AC02-09CH11466 and DE-FC02-04ER54698.
        Speaker: Mr A. M. Garofalo (USA)
      • 14:00
        Applying the Radiating Divertor Approach to Innovative Tokamak Divertor Concepts 4h 45m
        Results are reported and interpretation made of recent experiments on DIII-D that assess the effectiveness of three innovative tokamak divertor concepts under radiating divertor (RD) conditions: (1) high performance standard double-null divertor (DND) plasmas, (2) high performance double-null “snowflake” (SF-DN) plasmas, and (3) single-null H-mode plasmas with different parallel connection lengths between their X-points and outer divertor targets (L||-XPT). In general, all three concepts are attractive, with reduced heat flux and good H-mode confinement. Significant reductions in peak divertor heat flux (q⊥,P) of more than 50% and 85% at the outer and inner targets, respectively, were observed in DND plasmas under neon/deuterium-based RD conditions, and high performance metrics were maintained, e.g., βN ≅ 3.0 and H98(Y,2) ≅ 1.35. Under these RD conditions, <20% of the input power (≈10-13 MW) was radiated in the core, while >40% outside the main plasma. Impurity injection from poloidal locations other than the private flux region opposite the Bx∇B drift direction produced high levels of fuel dilution. High performance SF-DN plasmas mirrored the DND results under similar RD conditions. While the heat flux profiles at the inner target of the SF-DN and DND plasmas behaved similarly under comparable RD conditions, q⊥,P at its outer divertor target of the SF-DN cases was generally about a factor of two lower. Impurity build up in the main plasma, however, was 15%-20% higher in the SF-DN, due in part to difficulty in pumping the broad density profile under the outer divertor leg of the SF-DN. Plasmas with longer L||-XPT had lower q⊥,P than those with the shorter L||-XPT. SOLPS modeling has indicated that cross-field transport between the X-point and the divertor target resulted in broadened heat flux profiles and reduced q⊥,P. Under similar RD conditions, the longer L||,XPT cases maintained lower q⊥,P by at least 50%. Partial detachment at the outer divertor under RD conditions occurred at lower bar_ne in the longer L||,XPT cases. This study represents a first systematic step in examining three potential solutions to the excessive power loading expected in future generation high-powered tokamaks. This work was supported in part by the US DOE under DE-FC02-04ER54698, DE-AC52-07NA27344, DE-AC05-00OR22725, DE-FG02-07ER54917, DE-FG02-04ER54541, and DE-AC04-94AL85000.
        Speaker: Dr Thomas W. Petrie (General Atomics)
        summary
      • 14:00
        Avoidance of Tearing Mode Locking and Disruption with Electro-Magnetic Torque Introduced by Feedback-Based Mode Rotation Control in DIII-D and RFX-Mod 4h 45m
        Disruptions caused by tearing modes (TM) are considered to be one of the most critical roadblocks to achieving safe steady state operation of tokamak fusion reactors. Here, a new scheme to avoid such disruptions is proposed by utilizing the electro-magnetic (EM) torque produced with 3D coils that are available in many present tokamaks. In this scheme, the EM torque to the modes is created by a toroidal phase shift between the externally-applied field and the excited TM fields, compensating the mode momentum loss due to the interaction with the resistive wall and error field. Fine control of torque balance is provided by a feedback scheme. Experimentally, we have explored this in DIII-D, a non-circular divertor configuration with 3D coils inside the wall, and made comparisons to the performance in RFX-mod (operated as a tokamak), a circular limiter device with active feedback coils outside the wall using a slightly different, independently developed technique. In high beta poloidal discharges in DIII-D, by applying sufficiently high gain, a large amplitude m/n=2/1 TM propagating initially with the plasma rotation was successfully slowed down in a controlled manner to very low frequency, i.e. of the order of the inverse of resistive wall time τ_w, and sustained over several seconds. Upon termination of feedback, the mode became locked leading to disruption. The controllability of torque balance was illustrated by rotating the mode forward or backward with varying feedback phase shift. In RFX-mod the controllability of mode rotation has been demonstrated at moderate plasma density (n/n_G<0.5 where n_G is the Greenwald limit) and low q(a) [q(a)<2.5], using pure real gains with no added phase shift. Here the key element for sustaining a mode at slow rotation frequency, ~ 1/τ_w, is the minimization of coil-sideband pollution. Theoretical models developed independently for both devices showed that experiment observations are consistent with common understanding. The achievements in both devices showed that this approach is robust, suggesting that the application to ITER would expand the horizon of its operational regime. The 3D coils currently under consideration for ELM suppression would be well suited to this purpose. This work was supported by the US Department of Energy under DE-AC02-09CH11466, DE-FC02-04ER54609, DE-FG02-08ER85195, and DE-FG02-04ER54761.
        Speaker: Dr Lionello Marrelli (Italy)
      • 14:00
        Can Gyrokinetics Really Describe Transport in L-Mode Core Plasmas? 4h 45m
        The common view in fusion theory is that nonlinear gyrokinetics constitutes a reliable first-principles approach to describe turbulent transport in MCF devices. Surprisingly, however, two recent findings challenged this notion. First, the experimental ion heat fluxes in the outer core of certain DIII-D L-mode discharges were underpredicted by GK simulations by almost an order of magnitude. This finding has been dubbed the „shortfall problem“ and has triggered extensive theoretical efforts on an international level. Second, a careful analysis of some L-mode discharges in the JET tokamak revealed a significant reduction of ion temperature profile stiffness in the presence of strong NBI [1]. This was first attributed to a combination of high toroidal flow shear and low magnetic shear. However, nonlinear GK simulations failed to confirm this suspicion, overpredicting the observed fluxes by up to an order of magnitude. This finding could be called the „excess problem“ and is as severe as the shortfall problem described above. The main goal of the present contribution is to revisit both of these problems and substantiate or refute them. At stake is the plasma theory community’s confidence to devise a predictive transport capability for devices like ITER or DEMO on the basis of nonlinear GK. Via careful studies with the GENE code (using about 30 million CPUh), both of these challenges could be met successfully. While the transport levels in outer-core L-mode discharges of DIII-D, C-Mod, and ASDEX Upgrade [2] can be reproduced within the experimental error bars, the observed ion temperature stiffness reduction in JET can be explained in terms of nonlinear electromagnetic effects in the presence of fast ions [3]. Thus, a number of ideas about possible elements missing in the present theoretical description or even a possible breakdown of GK are identified as premature. Meanwhile, these studies highlight the fact that the search for adequate minimal models of turbulent plasma transport under various experimental circumstances is highly non-trivial. [1] Mantica, PRL 107, 135004 (2011) [2] Told, PoP 20, 122312 (2013) [3] Citrin, PRL 111, 155001 (2013) This work was supported by EURATOM and carried out within the framework of the European Fusion Development Agreement. The views and opinions expressed herein do not necessarily reflect those of the European Commission.
        Speaker: Prof. Frank Jenko (IPP Garching)
        summary
      • 14:00
        Cherenkov Emission Provides Detailed Picture of Non-Thermal Electron Dynamics in the Presence of Magnetic Islands 4h 45m
        In the 2013 experimental campaign a novel optical diagnostic system, based on the Cherenkov effect, was installed on FTU, in collaboration with the NCBJ group (IPPLM Association) to study the dynamics of non-thermal electrons in the presence of magnetic islands. Data from the Cherenkov probe was correlated with data from several other diagnostics, including ECE, neutron and gamma detector, Mirnov coils, soft X-ray cameras, demonstrating that the modulation of the Cherenkov signal is due to the rotation of the magnetic island. An aspect of interest under investigation is the internal structure of the peaks of the Cherenkov signal, with sub-peak full-width at half maximum of the order of 10 microseconds, and approximately 20 microseconds sub-peak separation. This level of detail opens up new possibilities for the investigation of fast electrons dynamics in the presence of high-amplitude magnetic islands. The study will present the analysis of the experimental data focusing on the identification of the mechanisms and evidence of electron acceleration that can be extracted from the correlation with the magnetic island position and geometry during tearing mode instabilities. This work is in the initial phase; future work is planned to enable energy discrimination of the incoming electrons (for example through the implementation of a multi-channel probe), as well as an increase of the spatial information (by installing a second probe).
        Speaker: Dr Federica Causa (ENEA C. R. Frascati)
        summary
      • 14:00
        Compatibility of Internal Transport Barrier with Steady-State Operation in the High Bootstrap Fraction Regime on DIII-D 4h 45m
        A high bootstrap current fraction plasma regime is desirable for steady-state tokamak operation because it reduces the demands on external non-inductive current drive. Typically, this regime is characterized by high βN and an internal transport barrier (ITB), leading to concerns about stability limits and profile control with reduced external input (power). Recent DIII-D research has increased confidence in the potential of the high bootstrap fraction approach for applicability to a steady-state fusion reactor. Fully noninductive plasmas have been sustained for long durations with large-radius ITBs, bootstrap fraction≥80%, βN≥3, βT~1.5%, and with the ITBs and good confinement maintained even with low net torque from neutral beam injection (NBI). Building on earlier DIII-D work [1], the new experiments utilized an approach to fully noninductive operation based on removing the current drive by transformer induction. The plasmas exhibit excellent energy confinement quality, with H98y2~1.5. Similar confinement was obtained after reducing the NBI torque from ~6 Nm to <3 Nm. The excellent confinement is associated with the formation of an ITB at large minor radius in all channels (ne, Te, Ti, rotation). The very broad bootstrap current profile is fairly well-aligned with the total current profile, explaining why the minimum safety factor is high and constant or slowly increasing, and the ITB is maintained at large minor radius for ~4 s, more than three times the current profile relaxation time, τCR estimated to be ~1 s. A further important result, providing evidence of dynamical stability, is that the ITB is maintained at large minor radius despite edge localized mode (ELM) perturbations, which become particularly large at the highest obtained βN~3.5. Stability analysis shows that this βN value is close to the ideal wall MHD stability limit, because a large outer gap was used to reduce wall heating by prompt fast ion losses. However, detailed analysis shows that fast ion losses are anomalously high only during the βN and density ramp-up phase. Future experiments will test an optimized outer gap waveform and methods of ELM control to enable a further increase of βN and thus of the plasma current. This work was supported in part by the US DOE under DE-FC02-04ER54698, DE-AC52-07NA27344, and DE-AC02-09CH11466. [1] P.A Politzer, et al., Nucl. Fusion 45 (2005) 417.
        Speaker: Mr Andrea M. Garofalo (General Atomics)
      • 14:00
        Development of Fully Noninductive Scenario at High Bootstrap Current Fraction for Steady State Tokamak Operation on DIII-D and EAST 4h 45m
        A main goal for EAST is to investigate an approach to fully noninductive long-pulse operation based on high bootstrap fraction under fusion relevant conditions. Building on results of previous DIII-D experiments , first joint experiments on DIII-D have developed a fully noninductive plasma scenario with EAST-similar plasma cross section shape, plasma current formation consistent with the superconductive coils on EAST, and values of plasma current, toroidal field, and heating power consistent with the new EAST capabilities. These EAST demonstration discharges on DIII-D have achieved and sustained fully noninductive conditions at ƒBS≥80%, beta_N≥3.0, beta_T~2.0%, and ƒGreenwald≥90%. Data supports that an ITB observed at large minor radius (rho~0.7) can be consistent with steady-state operation. Excellent energy confinement and high normalized pressure (beta_N>3) were maintained even with the low NBI torque (~3 Nm) expected on EAST. ELM dynamics appear as a limiting instability toward stationary sustainment of even higher performance. These fully noninductive high bootstrap discharges need to pursue integration of ELM control on both of DIII-D and EAST. The experimental conditions on DIII-D have been used to simulate possible advanced steady-state scenarios for EAST, and the first experimental tests of these scenarios on EAST will be presented. Simulations using PTRANSP with the CDBM transport model show that such DIII-D scenarios are accessible on EAST for 0.5MA steady-state plasma at βN~2.4 and with bootstrap current fraction of 65% by utilizing half of the total EAST H&CD capabilities planned for 2014. A weak core magnetic shear similar to the DIII-D scenario and with ITB footprint at rho~0.6 is achieved in the simulations. The scenario demonstrated on DIII-D and predicted for EAST could be extended to durations much longer than the current relaxation time and even the wall equilibration time using the expected EAST capabilities in the upcoming campaign, prior to the IAEA conference. Success of this endeavor will be a significant progress toward the goal of fusion energy. This work was supported in part by the US Department of Energy under DE-FC02-04ER54698, DE-AC52-07N27344, and DE-AC02-09CH11466, and the National Magnetic Confinement Fusion Science Program of China under No. 2011GB101000.
        Speaker: Prof. Xianzu Gong (Insititute of Plasma Physics, Chinese Academy Sciences)
      • 14:00
        Electron Temperature Critical Gradient and Transport Stiffness 4h 45m
        In a continuing effort to validate turbulent transport models, the electron heat flux has been probed as a function of electron temperature gradient on the DIII-D tokamak. In the scan of gradient, a critical electron temperature gradient has been found in the electron heat fluxes and stiffness at various radii in L-mode plasmas. The TGLF reduced turbulent transport model [G.M. Staebler, J.E. Kinsey, and R.E. Waltz, Phys. Plasmas 14, 055909 (2007)] and full gyrokinetic GYRO model [J. Candy and R.E. Waltz, J. Comput. Phys. 186, 545 (2003)] obtain the observed critical gradients and stiffnesses, but they do not predict the absolute level of transport at all radii. Here the stiffness is defined as the ratio of the heat pulse diffusivity (obtained by modulating 1 of the electron cyclotron heating gyrotrons) to the power balance diffusivity. For a regime where the heat flux is linearly proportional to the temperature gradient (with no offset), the stiffness is 1. There can be a critical gradient above which the flux is no longer proportional to the gradient; the stiffness will then jump above 1. Consistent with a critical gradient paradigm, the inferred stiffness at each radial location starts around 1 and jumps up above 1 at a critical gradient. The value of the critical gradient is observed to increase with radius in the plasma. The TGLF and GYRO predicted fluxes and stiffnesses exhibit a similar critical gradient; however there are conditions, such as when the experimentally inferred fluxes are large compared to gyrobohm fluxes, under which both codes underpredict the baseline level of transport. In addition to inferring the power balance and heat pulse diffusivities, the electron temperature fluctuations were measured at ρ=0.6 using the Correlated Electron Cyclotron Emission diagnostic. Although the GYRO code does reasonably well at predicting the fluxes at this radius, it substantially underpredicts these electron temperature fluctuations, while still exhibiting a critical gradient similar to the measured fluctuations. This is a quandary because, in the past, predicted fluxes and fluctuations either both agree or disagree with experiment. The nature of the experiment-code discrepancies is prompting reevaluation of the range of wavenumbers used in GYRO simulations. This work was supported by the US Department of Energy under DE-FC02-04ER54698.
        Speaker: Dr Sterling Smith (General Atomics)
      • 14:00
        ELM Pacing with Periodic Plasma Column Displacements 4h 45m
        Although ELM control by RMP's has been successfully demonstrated on a number of tokamaks, whether this method will be successful in all anticipated ITER operational scenarios is not clear. Thus, there is a recognized need to investigate other approaches to reducing transient heat and particles loads due to ELM's on PFC's in ITER. A low-cost technique that may not require dedicated equipment is the ELM-pacing with periodic vertical displacement (or ``jogging'') of the plasma. Since its first demonstration on the TCV tokamak, vertical jogging has been shown to be an effective mitigation scheme on ASDEX, JET, and other devices. More recently, KSTAR has demonstrated ELM control and mitigation using a number of different techniques, including vertical jogs of the plasma. However, a general understanding of the physics consistent with all experimental observations is still missing. On TCV, ELM triggering by vertical jogs is attributed to increased edge currents induced by the plasma displacement away from the X-point. On ASDEX, however, ELM's are triggered when the plasma is moving towards the X-point, with a decrease in the edge currents. The peeling-ballooning stability boundary sensitively depends on the parallel current and pressure profiles at the edge, both of which are affected by plasma displacements. In particular, the current is modified both inductively and through induced changes in the pressure gradient. The inductive component seems to play a dominant role in the TCV results, consistent with the fast time-scale of the “kicks.” Results from other tokamaks are in qualitative agreement with the assumption that the changes in the pressure gradient, either directly (at the ballooning boundary) or indirectly through its effect on the bootstrap current (at the peeling boundary), play a more significant role in triggering ELM's. A more quantitative comparison with experimental results and possible implications of this method for ELM-mitigation in ITER await detailed stability calculations.
        Speaker: Dr Ahmet Y. Aydemir (National Fusion Research Institute, Daejeon, Korea)
        summary
      • 14:00
        Expanding the Physics Basis of the Baseline Q=10 Scenario toward ITER Conditions 4h 45m
        Results obtained recently in DIII-D provide critical information for ITER baseline scenario operation. Much of the physics basis for ITER baseline scenario operation has been obtained in plasmas with significant fueling and applied torque from neutral beam injection (NBI). DIII-D has unique capabilities to extend this physics basis toward ITER conditions by applying neutral beam injection (NBI) with combinations of co- and counter-injection to reduce torque input, applying electron cyclotron heating (ECH) to reduce fueling and torque and to equilibrate the electron and ion temperatures, and exploring the effects of steady-state and transient divertor heat flux reduction with radiative divertor operation. All of these tools have been applied to plasmas with a boundary shape close to that of ITER to minimize systematic effects in projection of the results to ITER. The existence of stationary plasmas at nearly zero applied torque in DIII-D with sufficient normalized pressure and confinement for Q=10 in ITER at 15 MA is a key validation of the baseline scenario. Sustained operation with normalized parameters sufficient for Q=10 operation in ITER (βN=1.9, H_98=1.05, I_N=1.41) has been achieved with nearly zero external torque input from NBI (TNB=0.3 Nm) for more than four resistive relaxation times (4τR). Similar conditions with PEC>PNB and TNB=0.5 Nm have been sustained for >3τR. Confinement at low torque is reduced relative to the standard co-NBI, but there is sufficient confinement margin in DIII-D so that the reduction brings the plasmas to H98~1. Application of ECH does not reduce the confinement quality of the plasmas relative to those with NBI only as long as plasmas at the same applied torque are compared. Radiative divertor operation is successful in reducing steady-state and transient heat flux to the divertor in DIII-D at low q95 and reduced torque without enhanced accumulation of the seed impurity used for radiation. The studies of the flux usage indicate sufficient flux should be available in ITER to meet the >300 s operational requirement. However, the operational difficulties encountered with tearing mode stability at low applied torque suggest that a more diverse set of plasmas should be considered for the Q=10 mission, due to the sensitivity of ITER to disruptions. This work was supported by the US Department of Energy under DE-FC02-04ER54698.
        Speaker: Dr Timothy C. Luce (General Atomics)
        Poster
        summary
      • 14:00
        Experimental Simulation of Burn Control Using DIII-D In-Vessel Coils 4h 45m
        A new approach has been developed to control fusion power by applying a non-axisymmetric magnetic field (n=3) using the DIII-D in-vessel coils (I-coils) to modify the energy confinement time. This has potential advantages for a power plant due to the reduced power requirements relative to auxiliary heating and that it may enable the control of the plasma response more rapidly than with fueling or impurity influxes due to recycling of the fuel gas and impurities. In the relatively low collisionality DIII-D discharges, the application of non-axisymmetric magnetic fields results in a decrease in confinement time and density pump-out. The stored energy, which is used as a surrogate for fusion power, was controlled by the application of non-axisymmetric fields. The regulation of the stored energy by means of I-coil feedback yields comparable to or more stationary conditions than by the conventional approach of varying the neutral beam power. Transient increases in neutral beam power were used to simulate alpha-heating excursions. The feedback loop largely compensated the increased heating power by increasing the I-coil current, which reduced the energy confinement time. The accompanying increased particle transport in the pedestal was compensated by means of feedback control of the density at the top of the pedestal using the Thomson scattering system and fueling by means of the gas system. TRANSP was used to examine fast ion and profile effects in the interpretation of these results. These experiments demonstrated that it is possible to control the stored energy, which is a proxy for fusion power, by means of applying non-axisymmetric magnetic fields. The work was supported by US Department of Energy under DE-AC02-09CH11466, DE-FC02-99ER54512 and DE-FC02-04ER54698.
        Speaker: Dr Egemen Kolemen (USA)
        summary
      • 14:00
        Experiments and Modelling on FTU Tokamak for EC Assisted Plasma Start-up Studies in ITER-like Configuration 4h 45m
        The intrinsic limited toroidal electric field (0.3 V/m for ITER) in devices with superconducting poloidal coils (ITER, JT-60SA, DEMO) requires an additional heating, like Electron Cyclotron (EC) waves, to initiate plasma current and to sustain it during the burn-through phase. The circular full metallic FTU tokamak, equipped with an ECRH system (140 GHz, 0.5 s, up to 1.6 MW), has contributed in the past to a wide documented study on the possible configuration and perspective of EC assisted plasma breakdown, performing experiments focused on low electric field start-up with perpendicular injection of EC power. Afterwards a new experimental and modelling activity, addressing the study of assisted plasma start-up in a configuration close to ITER one (magnetic field, wave oblique injection and polarization) realized on FTU, has been initiated and presented. These new experiments have been supported by a 0-D code, BKD0, developed to model the FTU plasma start-up and linked to a beam tracing code to computing, in a consistent way, EC absorption. The FTU results demonstrate the role of polarization conversion with oblique injection at the inner wall reflection, confirmed by a faster plasma current ramp up (from 3.5 to 8.6 MA/s), when the waves reflects on the inner vessel surface. This effect is related to the higher temperature reached as a result of the better power absorption of extraordinary polarized waves generated at reflection, guaranteeing a wider operational window in term of filling pressure and toroidal electric field. Dedicated experiments showed also the capability of EC power to sustain plasma start-up even in presence of strong vertical magnetic field (10mT), with a null outside the vacuum vessel. These results assume more and more importance considering that the first plasma in ITER will be likely obtained at half field (2.5 T), where the influence of stray field is doubled. The 0-D BKD0 code, developed and applied to FTU data to reproduce the ITER relevant configurations at half and full toroidal magnetic field, has been used to determine operational window of sustained breakdown as a function of toroidal electric field and filling pressure. Experimental results are in agreement with the BKD0 simulations, supporting the use of the code to predict start-up also in future tokamaks, like ITER and JT60SA.
        Speaker: Dr GUSTAVO GRANUCCI (ISTITUTO DI FISICA DEL PLASMA - CNR)
        summary
      • 14:00
        Experiments on Magneto-Hydrodynamics Instabilities with ECH/ECCD in FTU Using a Minimal Real-Time Control System 4h 45m
        Experiments of magneto-hydrodynamics (MHD) instabilities control using injection of Electron Cyclotron Waves (ECW) are being performed in the FTU tokamak at toroidal field of 5.3T, plasma current of 0.5MA, line averaged density of 0.6 10^20 m^-3. The control system is based on only three real time key items: an equilibrium estimator (EQUIFAST) based on a statistical regression, a MHD instability marker (SVDH) using a 3d array of pick-up coils and the fast ECW launcher. One beam of 0.33MW, 140GHz, max pulse duration of 0.5s has been used for heating in the Ordinary Mode polarization (OM1) with perpendicular toroidal injection and nearly on axis EC resonance. The EC absorption volume has been controlled by the poloidal steering of the launcher. The MHD activity (m,n=2,1 or 3,2 modes) has been deliberately induced either by Neon impurity injection or by a density ramp hitting the density limit. No diagnostics providing the radial localization of the instabilities have been used. This is given a-posteriori through the evaluation of the effectiveness of the stabilization. When the ECW power is switched on, the instability amplitude shows a marked sensitivity to the position of the absorption volume with an increase or decrease of its growth rate. A significant reduction of the MHD amplitude has been obtained during the ECW injection phase. However, the continued cooling by Neon recycling that originates the instabilities does not allow their complete suppression at least at this ECW power level. The MHD control loop has been modified for the density limit experiments. The automatic search of the steering angle producing the fastest instability reduction has been introduced, based on the evaluation of the time derivative of the MHD amplitude. Once such angle is reached the controller holds the position until the SVDH signal crosses the switching off threshold. This control criterion has led to the suppression of the instabilities, even if the ECW injection is in some cases accompanied by a density pump-out reducing the density below the onset threshold. These experiments are the first attempt of feedback control of the ECW launcher in FTU for MHD control purposes. The control tools used are essential and based on a minimal set of diagnostics. Such experimental condition mimics the situation of a fusion reactor where reduced diagnostics capabilities are expected.
        Speaker: Dr Carlo Sozzi (Istituto di Fisica del Plasma CNR, Associazione EURATOM-ENEA)
        Poster
        summary
      • 14:00
        Extreme Low-Edge Safety Factor Tokamak Scenarios via Active Control of Three-Dimensional Magnetic Field on RFX and DIII-D 4h 45m
        High current, stable tokamak plasmas with edge safety factor below or around 2 are attractive for magnetic fusion due to favourable high fusion gain and higher confinement. But they have long been considered inaccessible in modern devices owing to the unforgiving MHD instabilities. Even in tokamaks with a resistive wall, the onset of an n=1 resistive wall mode leads to a disruptive limit at edge safety factor q_edge =2 (for limiter plasmas) and q_95≈ 2 (diverted plasmas). This paper reports how for the first time two very different tokamaks, a large MA-class shaped device like DIII-D and a high aspect ratio circular experiment like RFX-mod, have robustly overcome the edge safety factor = 2 limit by active control of plasma stability and demonstrate that operation below 2 is possible for hundreds of resistive wall times. In addition these experiments reveal a new tool to control sawtooth frequency and amplitude, a result that has the potential of extending the tokamak operating space even further by avoiding deleterious giant sawteeth. The application of 3D fields with a strong n=1 component that couples to the m=1,n=1 internal kink is found to significantly reduce the sawtooth amplitude and increase their frequency, demonstrating the benefits of this helical state. Experimental results have been compared with theory and numerical models. The experimental stability limit has been identified in DIII-D limiter plasmas as q_edge=2.08 ± 0.011, slightly higher than the external kink mode limit q_edge=2.0 predicted by ideal MHD analysis (DCON). In both devices the approach to the stability limit is characterized by the onset of an n=1, m=2 mode and growth rate of the order of τwall, consistent with ideal MHD predictions. RFX-mod shows that at q_edge<2.0 the growth rate of the uncontrolled modes decreases as q_edge is lowered towards 1.5, i.e. indicating stability improvement as q_edge=1.5 is approached from above. This is consistent with external kink mode stability expected from ideal MHD theory at q_edge<2.0: a peak in the external kink growth rate is expected near, but below q_edge=2.0, followed by a decrease in the growth rate. Sawtooth destabilization via applied magnetic perturbation is explained with the nonlinear MHD PIXIE3D code as a purely nonlinear effect, and not simply a modification of the (1,1) kink linear stability.
        Speaker: Prof. Piero Martin (Consorzio RFX)
        summary
      • 14:00
        Fishbone Modes in Plasmas with Dual Neutral Beam Injection Heating 4h 45m
        Neutral beam injection (NBI) is one of the important methods to heat plasmas in current tokamaks. However, fishbone instability induced by fast ions during NBI experiment is the main source for resulting in fast ion loss. Generally speaking, the density gradient of fast ions is the primary driving force to destabilize the fishbone modes. Therefore, it is possible to reduce the instability by eliminating the density gradient of the fast ions, employing dual neutral beam injection (DNBI) scheme in tokamak plasmas. The DNBI refers to two NBI lines heating plasma with one at the magnetic axis and the other (called off-axis NBI) at another radial position. With such tangential DNBI, a radial density profile of fast ions can be formed from superposition of dual Gaussian distributions. The dispersion equation for the fishbone instability is numerically solved for a density profile of fast ions of DNBI. A slowing down distribution with Gaussian pitch angle profile is used for each NBI. The dependences of the real frequency and growth rate of the fishbone modes on the parameters such as beta of hot ions (ratio of fast ions pressure/magnetic pressure), Delta (the distance between the axis and deposition position one of the off-axis NBI) and chi (ration of the on-axis NBI intensity and the off-axis one) are investigated in detail. The results show that the density distribution of fast ions from DNBI can bring about a stable window in the radial direction where the fishbone mode cannot be excited by fast ions. The width of the stable window increases linearly with radius increasing of magnetic flux surface of safety factor q=1. Besides, the width of the stable window increases with decreasing of density profile index of fast ions and keeps constant for large enough density profile index. The growth rates of fishbone modes dramatically decrease with the ratio of DNBI intensity and the critical beta values of fast ions increase with increasing of the ratio. The fishbone instabilities can be avoided with DNBI and may be an effective method to prevent fast ion loss resulted from fishbone modes.
        Speaker: Dr Hongda He (Southwestern institute of physics)
        summary
      • 14:00
        Fluid Simulation of Particle and Heat Fluxes during Burst of ELMs on EAST and DIIID 4h 45m
        In this paper we report the simulations of the evolution of the particle and heat flux during the burst of ELMs in realistic discharges on DIIID and EAST tokamaks. A set of six-field two-fluid equations based on the Braginskii equations with non-ideal physics effects is found to simulate pedestal collapse under the BOUT++ framework [1]. In general studies with shifted-circular geometry, the analysis of radial transport coefficients indicates that the ELM size is mainly determined by the energy loss at the crash phase. The typical values for transport coefficients in the saturation phase after ELM crashes are Dr∼200m2/s, χir∼χer ∼ 40m2/s. The DIIID ELMy H-mode discharge #144382 is a lower single-null, small ELM crash event detected with multiple fast acquisition data chords in the pedestal, scrape-off layer (SOL) and divertor [2]. The measured density, temperatures and electric field profiles inside the separatrix are used in our simulation. The ELMs of this discharge is destabilized by the resistive-ballooning modes according to the linear analyzes. In order to consider the kinetic modification of parallel transport in SOL, the sheath limit of the flux limited expression are applied for the parallel thermal conduction. The energy loss during our simulation is around 18kJ, which is close to the experiment measured value 17kJ. The collapse width of the electron density profile is the same as the measurements. The peak amplitude of heat flux distributions on divertor targets in our simulation is 700W/cm-2 at 0.28ms after ELM crash, compared to the measured value 500W/cm-2. The radial heat flux distributions indicate that this ELM is convective dominant. For EAST ELMy H-mode discharge #38300 [3], which is close to double null geometry, the measured profiles of density and temperatures inside the separatrix are used in our simulation. This discharge is ideal peeling-ballooning unstable. The power loss of the simulation is around 0.7MW, which is the typical value of EAST discharges with LHCD. The dependency of the direction of toroidal magnetic field on the asymmetric distribution of particle fluxes on upper and lower divertor targets will be reported in this paper. [1] Nuclear Fusion, 53, 073009 (2013). [2] M.E.Fenstermacher, et al. 40th EPS, P4.104. [3] PPCF., 55, 125008 (2013).
        Speaker: Dr Tianyang XIa (Institute of Plasma Physics, Chinese Academy of Sciences)
        summary
      • 14:00
        Frequency and Damping Rate of the Geodesic Acoustic Mode in Collisional Plasmas 4h 45m
        The frequency and damping rate are two most fundamental properties of the GAM. The collisional effect could be important in the plasma edge. In our work [1], where a number conservation Krook collisional operator was used in the gyrokinetic model, it was found that the damping rate of the GAM is non-monotonic as the collision rate increases. At low ion collision rate the damping rate increases linearly with the collision rate; while as the ion collision rate is higher than v_ti/R, the damping rate decays with an increasing collision rate. At the same time, as the collision rate increases, the GAM frequency decreases. However, it is noted that the number-conserving Krook collision operator is rather approximate. It is of interest to investigate the eigen-frequency of the GAM using more accurate operators and thereby find which properties of a collision operator are important for the dynamics of the GAM. In this work four different ion collision operators, including (a) a Krook operator with number conservation only, (b) a Krook operator with number and energy conservation, (c) a Lorentz operator which conserves number and energy automatically and (d) a Lorentz operator with an energy-dependent collision rate, are employed in a drift-kinetic model to investigate the collisional effect on the GAM frequency and damping rate. Comparison between different collision operators is performed as well. For operator (a), the result is the same as previously. For operator (b), the damping rate is only one ninth of that from (a). For operator (c), the damping rate approximates to that from (a) at low collisionality but give a lower damping rate than that from (b) and (a) at high collisionality. The result from operator (d) is close to that from (c). Due to finite collisional damping, the GAM frequency decreases. At very high collisionality, the GAM frequency approaches to v_ti/R for operator (a) but to sqrt(5/3)v_ti/R for other three operators. The result shows that both density and energy conservation of the collision operator are important for determining the GAM frequency and damping rate. The absence of energy conservation induces the overestimation of collisional effect at high collisionality. Work supported by NSFC, under Grant Nos. 10990214, 11075092, 11261140327 and 11325524, MOST of China, under Grant No. 2013GB112001. [1] Gao, Phys. Plasmas 20, 032501(2013)
        Speaker: Prof. Zhe Gao (Tsinghua University)
        summary
      • 14:00
        Full-f Neoclassical Simulations toward a Predictive Model for H-Mode Pedestal Ion Energy, Particle and Momentum Transport 4h 45m
        Optimization and control of the H-mode pedestal and scrape-off layer (SOL) for burning plasma devices such as ITER requires a predictive model for the transport of particles, energy and momentum from the top of the pedestal to the first wall. The transport simulation code XGC0 leverages high-performance computing to rigorously compute the full-f multi-species (D+, C6+, e-) flux-driven neoclassical transport with self-consistent neutral recycling and model-based anomalous transport in the H-mode pedestal and SOL. Net particle transport and electron thermal transport is anomalous, while ion thermal and momentum transport is predominately neoclassical in the steep-gradient region of the pedestal. The separate transport mechanisms resolves the decoupling of energy and particle transport often observed in the evolution of the H-mode barrier or in regimes with enhanced particle transport from an edge mode, such as QH-mode or I-mode. It is shown that the radial electric field (Er) in the pedestal is the root solution that balances ion orbit loss of high-energy counter-Ip ions against a pinch of colder ions. Neoclassical effects lead to non-Maxwellian ion energy distributions that manifest as intrinsic co-Ip parallel flows, Ti anisotropy (Tθ ≠ Tϕ) and non-monotonic Zeff profiles, especially at low collisionality. These effects are quantitatively demonstrated through comparisons of XGC0 simulations to a low-collisionality QH-mode pedestal and a zero-torque intrinsically rotating ECH-heated discharge on DIII-D. EPH-mode on NSTX is a stationary ELM-free regime that has a double barrier in the ion thermal transport and a single barrier in the particle transport. XGC0 simulations indicate that the enhanced thermal ion confinement occurs when the parallel rotational shear length is on the order of ion orbit, improving the confinement of tail ions and reinforcing the rotational shear. The neoclassical origin of Er results in a predictable connection between the E×B flow shear and magnetic geometry that can be leveraged to control the requirements for the L-H transition. For example, experiment and simulation demonstrate that a lower edge Ti is required to enter H-mode as the X-point increases to a low-triangularity shape on NSTX. This work was supported by the US Department of Energy under DE-AC02-09CH11466, DE-FG02-07ER54917, DE-FC02-04ER54698, and DE-AC05-00OR22725.
        Speaker: Dr Sterling Smith (USA)
        summary
      • 14:00
        Gyrokinetic Simulation of Microturbulence in EAST Tokamak and DIII-D Tokamak 4h 45m
        The new capabilities in the gyrokinetic simulation code GTC enable it to simulate the turbulent transport in real tokamak experiments. We apply these capabilities to simulate one ITG turbulence case for DIII-D tokamak and one TEM turbulence case for EAST tokamak with real experimental profiles and equilibrium magnetic field. For DIII-D case, the radial heat diffusivity profile simulated by GTC is highly consistent with that by GYRO. For the EAST case, we find that the collisional effect is very important in successfully explaining the low mode frequency and large wavelength for the electron coherent mode (ECM) observed in the EAST pedestal.
        Speaker: Mr Zhiyong Qiu (Institute for Fusion Theory and Simulation, Zhejiang University, Hangzhou, China)
        summary
      • 14:00
        Heat Transport and Enhancement Confinement Regimes in Tokamak as a Result of Plasma Selforganization 4h 45m
        Based on hypotheses about self-organization and pressure profile shape conservation in tokamak plasma, the turbulent heat transport processes are analyzed. The mechanism of internal transport barriers formation in regions without low number rational magnetic surfaces is suggested. The stronger pressure profile distortions from the self-consistent profile bring to the lower mode number excitation, increasing the heat flux. It is shown that the specific feature of the energy turbulent transport by low modes is the possibility of the internal transport barriers formation. The nontraditional explanation for H mode and regimes with the improved confinement (“advanced tokamak”) is suggested. Work is supported by ROSATOM contract № H.4x.44.90.13.1101.
        Speaker: Mrs Ksenia Razumova (NRC Kurchatov Institute)
        summary
      • 14:00
        High Internal Inductance for Steady-State Operation in ITER and a Reactor 4h 45m
        Increased confinement and ideal stability limits at relatively high values of the internal inductance (l_i) have enabled an attractive scenario for steady-state tokamak operation to be demonstrated in DIII-D. The potential of the scenario was shown in high elongation and triangularity double-null divertor discharges in which β_N>4.5 was achieved at l_i≈1.3. This high value of β_N just reached the ideal n=1 kink stability limit calculated without the effect of a stabilizing vacuum vessel wall, with the ideal-wall limit still higher at β_N>5.5. Confinement is above the H-mode level with H_98≈1.8. This type of discharge is a candidate for a reactor that could either operate stably at β_N≈4 without the requirement for a nearby conducting wall or n≥1 active stabilization coils, or at β_N≈5 with wall stabilization. With the high β_N and relatively high q_95=7, the discharge in the experiment is overdriven with bootstrap current fraction f_BS≈0.8, noninductive current fraction f_NI>1 and negative surface voltage. For ITER, operation at l_i≈1 is a promising option. Improved core confinement at high li could compensate for reduced H-mode pedestal confinement if a low pedestal height results from pedestal physics and/or ELM-stabilization using 3D fields. At l_i≈1, f_BS would be ≈0.5 with the remainder from external current driven efficiently near the axis. This scenario has been tested in the ITER shape in DIII-D at q_95=4.8, so far reaching f_NI=0.7 and f_BS=0.4 at β_N≈3.4 with performance appropriate for the ITER Q=5 mission, H_89 β_N/q_95^2 >0.3. High l_i discharges thus far take advantage of inductively driven current density near the axis as a partial substitute for externally-driven current. Studies with the FASTRAN transport code using the TGLF energy transport model explored how increased current drive power for DIII-D, 9 MW electron cyclotron current drive (ECCD) and 13 MW off-axis beam power, could be applied to maintain a stationary, fully noninductive high l_i discharge. Solutions are found at β_N=4, l_i=1.07, and f_BS=0.5 calculated stable without a conducting wall with ECCD and neutral beam current drive near the axis and at β_N=5 calculated to be stable with the vacuum vessel wall. This work was supported by the US Department of Energy under DE-FC02-04ER54698, DE-AC52-07NA27344, DE-AC05-00OR22725, DE-AC02-09CH11466, and DE-FG02-04E54761.
        Speaker: Dr John Ferron (General Atomics)
        summary
      • 14:00
        Impact of NBI-Injected Fast Ions in the Stabilization of the Resistive Wall Mode in High-β_N Plasmas 4h 45m
        Modeling results, obtained with the full kinetic MARS-K code [1] for a set of DIII-D experimental equilibria, predict that the absence of fast beam generated ions in ITER will lead to a plasma response ~40% higher than in the present NBI-sustained H-mode plasmas. It has been postulated that the presence of fast ions may have a stabilizing effect on the RWM that would account for its observed stability, specifically in the DIII-D tokamak (the fast-ion generating Neutral Beam Injection system is the main heating system DIII-D plasmas). These dependencies may extrapolate unfavourably to machines with significantly smaller fractions of fast ions such as ITER. Elevated plasma response values will likely cause the potential onset of a resistive wall mode (RWM) instability. If the RWM is destabilized, operation above the no-wall limit will likely require a set of coils and a feedback system capable of detecting and stabilizing the external kink instability on the wall-time scale (~2.5 ms in DIII-D). Active MHD spectroscopy can be used to indicate the approach to pressure driven MHD limits. This technique can define when the RWM feedback is needed and ultimately help to avoid disruptions. The MHD spectroscopy system reacts differently in the case of current driven instabilities. The plasma response amplitude in these cases remains low up to ~93% of the limit, showing an abrupt increase only in the last ~5% of the current ramp, making it much less effective as a warning system. However, new modelling shows that the mode structure of the current driven RWM is very close to that of the pressure driven case, as measured in DIII-D plasmas with q_{edge}~2 and modelled with the MARS-F and MARS-K codes. These equilibria, which show a hard disruptive limit when qedge crosses 2, were used to develop an RWM feedback system that allowed to cross the q=2 limit and sustain the plasma at q~1.9 for ~400 wall times [2]. [1] Y. Liu, Phys. Plasmas 15, 092505 (2008) [2] J.M. Hanson, et al, Bulletin of the American Physical Society, Division of Plasma Physics 58, TI2.00003 (2013). Work supported in part by the US DOE under DE-FG02-04ER54761, DE-AC02-09CH11466 and DE-FC02-04ER54698.
        Speaker: Dr Francesca Turco (Columbia University)
        summary
      • 14:00
        Influence of Boundary Conditions on Turbulent Transport and Plasma Energy Confinement Time Evolution in Tokamaks with Additional Heating: Simulations for T-10 and T-15 Tokamaks 4h 45m
        Temporal evolution of anomalous transport and global plasma energy confinement time is studied in simulations of plasma turbulence in tokamaks with additional plasma heating. The simulations have shown that external boundary conditions with the specially chosen power dependence of heat fluxes on the local time-dependent values of plasma density and temperatures at the boundary of plasma core with SOL region can provide the evolution of the plasma confinement time to the known steady state plasma confinement scalings. Such boundary conditions can be interpreted as the corresponding power scaling for an effective plasma confinement time in SOL. The first set of simulation runs were performed for conditions of experimental shots at T-10 tokamak using CONTRA-C code (cylindrical geometry). The second set of simulations were performed both for T-10 experimental shots and for expected T-15 conditions using transport code ASTRA with special turbulent block CONTRA-A (toroidal geometry with non-circular plasma cross-section). Both codes are based on adiabatically reduced MHD-like equations of turbulent convection.
        Speaker: Dr Vladimir Pastukhov (NRC Kurchatov Institute)
        summary
      • 14:00
        Measurement of Radiated Power Asymmetry during Disruption Mitigation on the DIII-D Tokamak 4h 45m
        Experiments have been undertaken on the DIII-D tokamak to examine the magnitude and causes of radiated power asymmetries during disruption mitigation. In order to mitigate the most deleterious effects of disruptions, massive quantities of radiating impurities can be injected into the pre-disruptive plasma to pre-emptively radiate away the stored thermal and magnetic energy. However, toroidal and poloidal asymmetries in the radiation pattern could still result in localized melting of ITER’s Be first wall. Measurements of the toroidal asymmetry in radiated power during disruption mitigation by massive gas injection (MGI) on the DIII-D tokamak indicate that the asymmetry during the thermal quench (TQ) and current quench (CQ) is largely insensitive to the number or location of injection sites [1]. Moreover, the observed absolute values of asymmetry during the TQ & CQ are well below those expected to be problematic for ITER. Infra-red imaging of the MGI valve location and surrounding wall indicates no highly localized, preferential heating of the injector location relative to the surrounding sector of wall, providing confidence that localized melting of the injector site in ITER is unlikely. Modification of the observed magnitude of the toroidal asymmetry during the TQ by application of a large n=1 error field supports recent modeling results that indicate large n=1 MHD during the TQ is the root cause of the radiation asymmetry [1,2]. Further work examines the poloidal radiation asymmetries resulting from massive impurity injection and the effect of spatially distributed impurity injectors upon those asymmetries. In addition, the radiation asymmetries observed during MGI are compared to those observed during shattered pellet injection [3]. This work was supported in part by the US Department of Energy under DE-FC02-04ER54698, DE-AC05-00OR22725, DE-FG02-07ER54917, DE-AC52-07NA27344, and DE-AC05-00OR23100. [1] N. Commaux, et al., submitted to Phys. Plasmas [2] V.A. Izzo, Phys. Plasmas 20, 056107 (2013). [3] N. Commaux, et al., Nucl. Fusion 50, 112001 (2010).
        Speaker: Dr Nicholas Eidietis (General Atomics)
      • 14:00
        MHD Instability Excited by Interplay between Resistive Wall Mode and Stable MHD Modes in Rotating Tokamak Plasmas 4h 45m
        A mechanism exciting magnetohydrodinamic (MHD) instabilities in rotating tokamak plasmas is found numerically for the first time. This mechanism is the interplay between a resistive wall mode (RWM) and a stable MHD mode. When a plasma has a discrete stable MHD eigenmode, the RWM can be destabilized when the plasma rotation frequency is close to the real frequency of the stable eigenmode. In a cylindrical plasma, such a destabilizing mechanism can be observed as the result of the interplay between RWM and a stable external kink mode. In a tokamak plasma, it is found that not only an external kink mode but also Alfven eigenmodes can be the counterpart of this interplay. It is numerically demonstrated that this mechanism can overcome the continuum damping leading to the destabilization of RWM in a realistic tokamak plasma. These results indicate that understanding of the stable MHD modes is important for robust stabilization of RWM. The destabilization can be avoided by optimization of the safety factor profile. This optimization is indispensable in the design of steady state high beta tokamaks such as JT-60SA, DEMO and future tokamak reactors.
        Speaker: Dr NOBUYUKI AIBA (Japan Atomic Energy Agency)
        summary
      • 14:00
        Momentum and Particle Transport Across the ITG-TEM Turbulence Regimes in DIII-D H-Mode Plasmas 4h 45m
        Turbulent particle transport and momentum transport have been shown to be closely connected theoretically as well as experimentally. In DIII-D H-mode plasmas we study the changes in particle and momentum transport across the linear ITG-TEM stability regime as well as studying the changes in particle transport as a function of the rotation profile. Counter to previous experimental [1], we do not find a correlation between normalized rotational shear (u’) and density peaking [max(a/Ln)] and only a weak dependence of core Te/Ti on density peaking. The density profiles do not become more peaked when we vary u’ from, by changing the momentum injection with the neutral beams. However, there is a correlation between the radius where, the maximum peak in the density scale length occurs and the radius at which the plasma transitions from ITG to TEM. We find that the intermediate density fluctuations increase in the core of the co- and balanced injected experiments and decrease outside mid-radius when compared with the counter-injected experiment. Together with a strong reduction in ExB shear outside mid-radius, this should result in a large increase in transport for the counter injected case. However, experimental perturbative D and v transport coefficients show that the increase in D for the counter-injected experiment is fully countered with an even stronger increase in v. In order to further study the role of turbulence in determining particle and momentum transport, we compare experiments in which we vary the heating from dominantly NBI-heated to ECH while keeping the injected momentum constant. We find that the ECH plasmas have a slightly more peaked density profile, but the overall density is strongly reduced. The rotation profiles for the ECH plasmas are more flat, but no rotation reversal is observed even though the plasmas transition from ITG to TEM close to mid-radius. The NBI heated plasmas all exhibit a rotational “well” in the pedestal area, which is not observed in the ECH heated plasmas and results in an overall lower core rotation. When adding a gas puff in the ECH heated plasma, we observe the return of the rotational well and the overall reduction of the core rotation at higher density. This work supported by the US DOE under DE-SC0007880, DE-FG02-08ER54984, and DE-FC02-04ER54698. [1] C. Angioni et al. Nucl. Fusion 51 (2011) 023006
        Speaker: Dr Saskia Mordijck (College of William and Mary)
        summary
      • 14:00
        Nonlinear Particle Simulation of Radio Frequency Waves in Fusion Plasmas 4h 45m
        Nonlinear global particle in cell simulation model in toroidal geometry has been developed for the first time to provide a first principle tool to study the radio frequency (RF) nonlinear interactions with plasmas. In this model, ions are considered as fully kinetic ion (FKi) particles using the Vlasov equation and electrons are treated as guiding centers using the drift kinetic (DKe) equation. FKi/DKe is suitable for the intermediate frequency range, between electron and ion cyclotron frequencies. This model has been successfully implemented in the gyrokinetic toroidal code (GTC) using realistic toroidal geometry with real electron-to-ion mass ratio. To verify this simulation model, we first use an artificial antenna to verify the linear mode structure and frequencies of electrostatic normal modes including ion plasma oscillation, ion Bernstein wave, lower hybrid wave, and electromagnetic modes and fast wave and slow wave in the cylindrical geometry. We then verify the linear propagation of lower hybrid waves in cylindrical and toroidal geometry. Because of the poloidal symmetry in the cylindrical geometry, the wave packet forms a standing wave in the radial direction. However, in the toroidal geometry, the waves propagate as two counter propagating waves in the poloidal direction due to the poloidal asymmetry of the magnetic field. The wave packet propagates faster in high field side compare to the low field side. This feature has been verified by the Wentzel–Kramers–Brillouin (WKB) solution. The nonlinear GTC simulation of the lower hybrid wave shows that the amplitude of the electrostatic potential is oscillatory due to the trapping of resonant electrons by the electric field of the lower hybrid wave. The nonlinear bounce frequencies have been verified with the analytic results. For comparison, in linear simulation the lower hybrid wave decays exponentially due to linear Landau damping.
        Speaker: Dr Animesh Kuley (University of California Irvine)
        summary
      • 14:00
        Nonlocal Transport from Edge to Core in Tokamak Plasmas 4h 45m
        A nonlocal response of plasma to edge density sources, which has been sometimes observed in toroidally magnetic confinement plasmas, is found in global fluid simulations. In the 4-field reduced MHD model, a toroidally-elongated particle source is applied in the edge region after saturation of the resistive ballooning mode turbulence is attained. The nonlocal transport appears at the location far from the edge source. The detailed process of the nonlocal transport is revealed for the first time. Both nonlinear and toroidal couplings between axisymmetric Fourier modes are responsible for the nonlocal transport. Especially (m,n)=(1,0) and (-1,0) modes play an essential role to produce the nonlocal transport, where m and n are poloidal and toroidal mode numbers, respectively. In the RMHD simulation only the resistive ballooning modes are unstable at the edge and no turbulence exists in the core region where the nonlocal response appears. The (1,0) pressure perturbation is an ingredient of the geodesic acoustic mode (GAM) oscillation of zonal flows (ZFs) and can be driven by drift wave turbulence such as ion temperature gradient (ITG) driven turbulence in toroidal plasmas. Therefore, it is interesting to investigate what happens in the nonlocal transport observed in the RMHD simulation if the ITG turbulence exists in the core. In order to study effects of edge density source on the ITG-ZF/GAM behavior, the global ITG turbulence code has been modified, in which the edge density source is implemented as a sink or cold pulse in the temperature equation. It is found in electrostatic ITG turbulence simulations that strong GAM oscillation is excited in an inner region than the sink location when the sink is imposed to simulate the cold pulse propagation. After investigating the above simulations in detail, we will perform simulations including both density source and temperature sink by extending the codes further.
        Speaker: Dr Naoaki Miyato (Japan Atomic Energy Agency)
        summary
      • 14:00
        Off-Axis Current Drive with High Harmonic Fast Waves for DIII-D 4h 45m
        Modeling shows that fast waves at very high ion cyclotron harmonics (also called “whistlers” or “helicons”) can drive current efficiently in the mid-radius region of a high beta tokamak plasma, as is required to sustain steady-state high performance discharges in a DEMO-like configuration. DIII-D has developed discharges with high electron beta and high electron temperature so that full first-pass damping of the waves is expected to take place off-axis. We show that in a specific existing high-beta DIII-D target discharge, 0.5 GHz fast waves at launched n_|| ~ 3–4 would drive a noninductive current of 60 kA/MW at ρ=0.55, where the electron density is ~5x10^19 m^-3 and the electron temperature is ~3 keV. With complete first-pass absorption, loss processes (mode conversion, far-field sheath formation, etc.) associated with weak single-pass damping are minimized. The calculated current drive efficiency is 2 to 4 times higher than that of off-axis neutral beams or electron cyclotron current drive using the present DIII–D systems. Strong, radially localized absorption on electrons can be obtained only for local values of βe exceeding 1.8%. At lower values, the waves propagate to smaller minor radius before being absorbed. Varying the launched value of n_|| shows that the driven current hardly changes in either magnitude or in radial location in the range of 2.8
        Speaker: Dr Robert Pinsker (General Atomics)
        summary
      • 14:00
        On the Measurement of the Threshold Electric Field for Runaway Electron Generation in FTU 4h 45m
        The determination of the threshold density value to be achieved by means of massive gas injection for runaway electron (RE) suppression in ITER relies on the relativistic collisional theory of RE generation which predicts that, below a critical electric field (E_R), no runaway electrons can be generated. No account of additional loss mechanisms, that may reduce the critical density, is usually made in the above treatment. Past experiments by means of electron-cyclotron-resonance heating in the flat-top phase of FTU discharges had shown that runaway suppression is found to occur at electric fields substantially larger than those predicted by the relativistic collisional theory of runaway generation. This was found to be consistent with an increase of E_R due to the electron synchrotron radiation losses, which lead to a new electric field threshold (E_R_rad). The main aim of this work is to verify whether the earlier finding that synchrotron radiation losses play a role in runaway generation is also confirmed in FTU discharges with no additional heating. Recent experiments in FTU have systematically investigated the conditions for RE generation in stationary and reproducible conditions (flat top of ohmic discharges) for a wide range of plasma parameter values: toroidal magnetic field (Bt=3-7.2 T), plasma current (Ip=0.35-0.5 MA) and Zeff (1.5-15). The threshold electric field for RE generation was measured using feedforward gas programming to obtain a decreasing electron density until RE are generated. The results indicate that the measured threshold electric field is larger by a factor ∼5-10 than expected according to the purely collisional theory and is very close to the new threshold calculated including synchrotron radiation losses (∼2E_R_rad).
        Speaker: Mr Basilio Esposito (ENEA)
        summary
      • 14:00
        Peaked Density Profiles Due to Neon Injection on FTU 4h 45m
        The density profile peaking produced by Ne-gas puffing was studied in in different L-mode plasma scenarios during recent experimental campaigns on FTU. In fact an stable radiative edge seeded with light impurities has beneficial effects and provokes density peaking without any undesirable central impurity accumulation [1,2]; on the other hand, a too large amount of impurities can lead to a disruptive MHD activity. In order to maintain the positive effect of the edge radiation it is important to fix the conditions of a strong increase of particle confinement while minimizing the amount of impurities needed. On FTU the Ne injection causes a spontaneous increase of the line average density by a factor 2 in the absence of Deuterium gas puffing. A Ne doped discharge is compared with a complementary one that reaches the same density by D gas puffing in the absence of Neon injection. The comparison shows a more peaked density for the Ne doped discharge. A qualitative estimate from UV spectroscopy measurements indicates that the density behavior cannot be attributed simply to the stripped electrons from the puffed Ne, but a modification of particle transport should be invoked in order to explain the spontaneous rise and the higher peaking. The recent experiments were devoted to characterize the plasma response to Ne injection at different densities and plasma currents. The principal results are: i) if density before the Ne puff is increased, the same injected impurity amount induces an abrupt density increase that rapidly ends in a disruption; ii) as the plasma current rises a less steep increase of the density is observed, and more D gas is necessary to obtain the same MHD activity that leads to the disruption. Finally, the observed density peaking is analyzed in terms of electron diffusion coefficients D and pinch velocity U. In the framework of a simple particle transport model [3], the presence of an inward pinch is confirmed. A micro-stability analysis will be performed to investigate the role of the ion and electron gradient driven modes on particle transport. [1] G. Telesca et al. 2000 Nucl. Fusion 40 1845 [2] A. Messiaen et al. 1996 Phys. Rev. Lett. 77 2487 [3] V. Zanza et al 1996 Nucl. Fusion 36 825
        Speaker: Mrs Cristina Mazzotta (ENEA)
        summary
      • 14:00
        Physics-Model-Based Control of the Plasma State Dynamics for the Development and Sustainment of Advanced Scenarios in DIII-D 4h 45m
        DIII-D experiment results are presented to demonstrate the potential of integrated physics-model-based q-profile and βN control for the systematic development and sustainment of advanced scenarios. Both simulations and experiments demonstrate improved control performance relative to unoptimized preprogrammed control, by utilizing a combined feedforward+feedback scheme. At the core of the control scheme is a nonlinear, physics-based, control-oriented model of the plasma dynamics valid for H-mode scenarios. A partial differential equation model of the q-profile dynamics is developed by combining 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 (both auxiliary and bootstrap). The plasma internal energy (related to βN) dynamics are modeled by a volume-averaged energy balance equation. 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 a target state (q-profile and βN) at a predefined time during the discharge, such that the achieved state is stationary in time, subject to the plasma dynamics (described by the physics-based models) and plasma state and actuator constraints. Secondly, integrated feedback controllers are designed to track a target q-profile and βN evolution with the goal of rejecting the effects of external plasma disturbances and adding robustness to the control scheme. The feedback controllers are synthesized by embedding both static and dynamic physics-based plasma response models into the control design and to be robust to uncertainties in the electron density, electron temperature and plasma resistivity profiles. The algorithms use the heating/current-drive system and total plasma current as actuators to control the plasma dynamics. Finally, experimental and simulation results are presented to demonstrate the capabilities of the optimized actuator trajectories and feedback controllers to control the plasma dynamics in DIII-D advanced scenarios. Work supported by US DOE (DE-SC0001334, DE-SC0010661 and DE-FC02-04ER54698).
        Speaker: Mr Justin E. Barton (Lehigh University)
      • 14:00
        Progress on Transport Modeling by Trapped Ion Resonance Driven Turbulence 4h 45m
        Predictive modeling of turbulent transport is essential to the success of ITER and DEMO. Due to the collisionless nature of fusion plasmas, turbulence with strong wave-particle interaction – such as collisionless trapped electron or ion modes (CTEM/CTIM) or energetic particle (EP) modes – can develop in fusion plasmas. However, transport caused by these turbulence cannot be described by the conventional quasilinear transport analysis. This is since the precession resonance allows a long correlation time between resonant trapped particles and fluctuations to produce a group of correlated resonant trapped particles, called granulations. In the presence of granulations, transport is not determined by quasilinear diffusion, but by Lenard-Balescu flux with Fokker-Planck drag. In this paper, we report progress on the modeling of turbulent transport caused by trapped ion granulations. In the first part of the work, we present application of transport caused by trapped ion granulations to the problem of toroidal momentum transport. In this part, we show that residual stress, which is a part of momentum flux that is not proportional to the velocity or velocity shear, can arise from momentum flux carried by trapped ion granulations. We discuss that this process can be viewed as a conversion of poloidal and toroidal momentum via trapped ion granulations. As an application to tokamak phenoemenology, we consider plasmas at L-H transiton, where poloidal ExB flows are accelerated. This acceleration can be converted to toroidal acceeration via the mechanism presented here. For typical parameters at the top of the pedestal of medium size tokamaks, we quantitatively find that this process can accelerate toroidal flows up to thermal Mach number ~0.1. In the second part of the work, we discuss the feedback mechanism from macroscopic flows to microscopic turbulence with trapped ion granulations. We discuss extension of Biglari-Diamond-Terry mechanism of turbulence suppression by shear flows to the problem of turbulence with degree of freedom in velocity space, in order to account for the effect on trapped ion granulations. We show that energy dependent trapped ion precession enters decorrelation process, in addition to conventional ExB scattering and shearing.
        Speaker: Dr Yusuke Kosuga (Institute for Advanced Study, Kyushu University)
        summary
      • 14:00
        Reduction of Net Erosion of High-Z PFC Materials in DIII-D Divertor Due to Re-Deposition and Low-Z Coating 4h 45m
        We report a substantial reduction of net compared to gross erosion of a tungsten PFC surface observed in DIII-D divertor in good agreement with modeling, and suppression of molybdenum erosion by a local gas injection. A sample featuring a 1 mm and a 1 cm diameter 15-24 nm thick W films deposited on a Si substrate over a carbon inter-layer was exposed in the lower divertor of DIII-D using the Divertor Material Evaluation System (DiMES). The exposure was performed in lower single null L-mode deuterium plasma discharges near the attached outer strike point (OSP) for a total of ~16 s. The plasma density ne = 1.2x10^{19} m^{-3} and electron temperature T_e = 32-35 eV near OSP were measured by the divertor Langmuir probes. Net erosion was determined by comparing Rutherford backscattering (RBS) measurements of the W layer thickness on the 1 cm spot before and after the exposure, and gross erosion was estimated from similar measurements on the 1 mm spot. The measured net and gross erosion rates were 0.14 and 0.48 nm/s, respectively, giving net/gross erosion ratio of 0.29. REDEP/WBC modeling of this experiment yielded a very close ratio of 0.33. A second exposure of a sample with similar W coatings on a Mo inter-layer performed in similar geometry at similar T_e but ~x3 higher ne yielded erosion rates about twice higher and net/gross erosion ratio of 0.38. Modeling of the second experiment is in progress. In another experiment, two Mo-coated Si samples 1 cm in diameter were exposed near attached OSP, first in L-mode for ~14 s, second in H-mode for ~7 s, with ^13CH_4 gas injected ~12 cm upstream of the samples. Suppression of Mo erosion was evidenced in situ by the disappearance of MoI line radiation at 386.3 and 390.2 nm once the gas injection was turned on. Post-mortem RBS analysis found the erosion of Mo near the center of the samples being below the measurement resolution of 0.3 nm, corresponding to a rate of 0.02 nm/s. Compared to the previously measured erosion rates in L-mode of 0.4–0.7 nm/s this constitutes a reduction of more than x20. Carbon deposition was measured on both samples, corresponding to a rate of ~20 nm/s in L-mode and ~4 nm/s in H-mode. The ratio of ^{13}C/^{12}C carbon in the deposits was about 1.6 on both samples, indicating that the deposition was largely from the gas injection. This work was supported by the US DOE under DE-FG02-07ER54917.
        Speaker: Mr Dmitry L. Rudakov (University of California San Diego)
      • 14:00
        Runaway Electron Control in FTU 4h 45m
        Runaway electrons (RE) are highly energetic electrons that might gain energy up to 20-30MeV (FTU). Runway electron beam can be harmful for plasma facing components: its low pitch angle allows the deposition of a high amount of energy on small areas yielding serious and deep damages of the vessel structure. For Tokamakas such as ITER, RE beams current should be around 11-13 MA and an impact with the vessel would irreparably damage the machine. We have proposed a new tool in the FTU plasma control system (PCS) for position and Ip ramp-down control of disruption-generated runaway electrons and first experimental results are discussed. The RE hybrid control algorithm switches among three phases: 1) in the Ip pre-quench phase the currents in the poloidal coils used for the radial control of the plasma are optimized by a dedicated algorithm called Current Allocator; 2) when the Ip quench is detected (on-line) the system performs actions to improve the radial control in case of formation of a RE plateau; 3) at the onset of the RE plateau, which is detected on-line by dedicated algorithms, the Ip is ramped down and the RE beam position is controlled, by means of real-time (RT) diagnostic signals (magnetics, neutrons), in such a way to minimize the interactions with the plasma facing components and safely shut down the discharge. This algorithm was tested in dedicated low density plasma discharges in which a significant RE population is generated during the Ip flat top and Ne gas is injected to induce a disruption: the rapid variation of the resistivity and the increased loop voltage at the disruption accelerate the pre-existing RE population and lead, in some cases, to the formation of a RE current plateau. The effectiveness of the RE hybrid control algorithm in this phase will be discussed as well as the possible improvements including code portability to other devices.
        Speaker: Dr daniele carnevale (Universita' Roma Tor Vergata, Dipartimento di Ing. Civile ed Ing. Informatica)
        summary
      • 14:00
        Study of Nonlinear Fast Particle Transport and Losses in the Presence of Alfvén Waves 4h 45m
        A nonlinear hybrid model is used to study energetic particle transport and losses in realistic TOKAMAK – particularly ASDEX Upgrade – multi-mode scenarios. The model consists of the vacuum-extended version of the drift kinetic HAGIS code. As crucial new elements of a realistic scenario, the perturbation structures, frequencies and damping rates are taken as obtained from the gyrokinetic eigenvalue solver LIGKA. In the view of ITER, where in certain scenarios, a “sea” of small-amplitude perturbations is likely, realistic multi-mode simulations will be carried out in the near-stability regime. This requires the use of the newly implemented non-local damping via accounting for a parallel electric field. The crucial question is, if the interaction between the “sea” of perturbations with the EPs will drive linearly stable or weakly unstable modes such that particle losses occur in a domino effect. Which of the modes are driven unstable in the simulation and their amplitudes can then be compared with the dominant modes measured in present-day experiments such as ASDEX Upgrade. Moving further above the stability threshold, it is not only important to account for the correct damping mechanisms. Also, the radial wave structure is very sensitive to the EP distribution function and is expected to evolve, as the distribution function changes. Although the nonlinear wave-particle interaction is calculated self-consistently within the HAGIS-LIGKA model, at the present status, other nonlinearities such as the evolution of wave structure and damping rate are not included yet. Before extending the model in this direction, the expected effect of the radial wave structure evolution is investigated, both with a numerical approach as well as with the help of experimental observation. Concerning the numerical study, HAGIS-LIGKA results are compared to those of a different hybrid code, HMGC, which already contains wave structure evolution. For that comparison, a new phase space diagnostic technique is developed for both codes, the so called Hamiltonian Mapping Technique. From the experimental side, next to Alfvénic modes, fishbones offer a good opportunity to model frequency and amplitude evolution according to experimental observation and compare the occurring transport in phase space.
        Speaker: Dr Mirjam Schneller (Max-Planck-Institut für Plasmaphysik)
        Poster
        summary
      • 14:00
        Suppression of Type-I ELMs with Incomplete I-Coil Set on DIII-D 4h 45m
        Recent experiments on DIII-D have demonstrated the ability to suppress edge localized modes (ELMs) using edge-resonant magnetic perturbations (RMPs) produced by an incomplete I-coil set in ITER similar shape plasmas with low pedestal electron collisionality. Robust ELM suppression has been reproducibly obtained on DIII-D using a wide range of toroidal RMP modes during experiments in which various non-axisymmetric coil loops were turned off pseudo-randomly. RMP ELM suppression was achieved on DIII-D with 11, 10, 9, 7, and 5 out of 12 I-coils. In these experiments, using fine I-coil current steps we determined the I-coil current amplitude threshold for RMP ELM suppression in each I-coil configuration. The suppression current threshold showed almost no dependence on the number of the active I-coils between the 11 and 7 coil configurations. These results provide confidence that the ITER ELM coils will likely be able to meet the ELM suppression criterion in case of multiple coil failures. The experimental results confirmed the previous modeling work predictions [1,2] that ITER ELM coils would be able to meet the ITER coil design criterion even with 19 of 27 loops by adjusting the coil currents within the allowed range of current amplitudes. In the DIII-D experiments, while the dominant n=3 harmonic was reduced due to an overall decrease in the amount of the perturbation, the sidebands assisted in maintaining necessary value of the island overlap region resulting in sufficient level of stochastization of the plasma pedestal region that is believed to be needed for RMP ELM suppression. This was also confirmed in the linear two-fluid plasma response modeling with M3D-C1. The effect of non-purity of the perturbation spectrum on ELM suppression may lead to new ELM suppression strategies and better understanding of the suppression mechanisms. Sustained RMP ELM suppression with only 7 of 12 I-coils was demonstrated on DIII-D. As the ELM suppression was achieved at constant I-coil current, high pedestal electron toroidal rotation and constant pedestal electron density were maintained for the duration of the ELM suppression phase, as well as good plasma confinement. This work was supported in part by the US DOE under DE-FG02-07ER54917 and DE-FC02-04ER54698. [1] D.M. Orlov, et al., Fusion Eng. Design 87 (2012) 1536 [2] T.E. Evans, et al., Nucl. Fusion 53 (2013) 093029
        Speaker: Dr Dmitri Orlov (University of California San Diego)
        summary
      • 14:00
        The Combining Effect of the Inductive Electric Field and the Lower Hybrid Waves on the Impurity Ions Toroidal Rotation in the Lower Hybrid Current Drive Tokamak Plasmas 4h 45m
        Plasma rotation in tokamaks driven by the lower-hybrid-waves (LHW) was firstly reported by the Alcator C-Mod team and also observed in the EAST tokamaks. The LHW injection can induce both co- and counter-current directed changes in toroidal rotation. The direct momentum absorption of the LHW induces the impurity ions to rotate in the counter-current direction. The inductive electric field decreases due to the drop of the loop voltage during the lower hybrid current drive. The inductive electric field in tokamaks has considerable effect on the impurity ions rotation and causes the impurity ions to rotate in the counter-current direction. The rotation of the impurity ions is usually measured in the experiments. The resulting rotation velocity of the impurity ions should be determined by the combining effect of the inductive electric field and the LHW. For the higher current case, the effect of the inductive electric field is negligible. For the lower current case, the effect of the inductive electric field can not be neglected.
        Speaker: Chengkang Pan (China)
        Poster
      • 14:00
        The Single Dominant Mode Picture of Non-Axisymmetric Field Sensitivity and its Implications for ITER Geometric Tolerances 4h 45m
        Experiments at DIII-D have demonstrated that several key 3D field sensitivities are directly related to their coupling to the least-stable kink mode of the plasma, and concomitantly that the plasma is remarkably insensitive to fields which have no net coupling to this single dominant (kink) mode. Specifically, plasma rotation and error field (EF) penetration thresholds are nearly unchanged despite application of large amplitude probing fields with no kink coupling. The plasma sensitivity to 3D fields which have no kink coupling is of critical importance as this sets the true geometric tolerance of the tokamak – so long as it is equipped with at least a single row of EF correction coils (EFCCs) and its 3D field sources are well characterized, thus allowing the kink-coupling of the intrinsic EF to be nulled by the EFCCs. The observed weak sensitivity to the no kink coupling field challenges the stringent tolerance requirements currently enforced, as a strong performance recovery when using EFCCs is expected though it is not presently taken into account. The validity of the single dominant mode picture is determined experimentally by contrasting the plasma sensitivity to large-amplitude probing fields that have varying levels of coupling to the kink mode. Sensitivity to rotation braking is contrasted in both H- and L-mode plasmas, where for both scenarios, braking by n=1 probing fields is reduced by nearly a factor of ten when the kink mode coupling is nulled. EF penetration is also contrasted with both H-mode and Ohmic plasmas. In both cases the penetration threshold is nearly unchanged (vs a no-field baseline) when the probing field has no kink coupling, despite its large amplitude. The maintenance of the edge rotation due to the neoclassical toroidal viscosity (NTV) with n=2 fields is also largest when coupling to the kink is maximized. A validated single dominant mode picture can also be applied to predicting optimal EFCC currents for any plasma scenario regardless of 3D field source. This is achieved by nulling the kink mode coupling of the intrinsic EF. Recent work shows that an exhaustive database of over 20 experimentally determined n=1 optimal EFCC currents is consistent with nulling the n=1 kink coupling of each plasma. This work was supported by the US DOE under DE-AC05-06OR23100, DE-FC02-04ER54698, DE-AC02-09CH11466, and DE-FG02-04ER54761.
        Speaker: Dr Carlos Paz-Soldan (Oak Ridge Institute for Science Education)
      • 14:00
        Thermal Loads on FTU Actively Cooled Liquid Lithium Limiter 4h 45m
        Power load on the divertor is one of the main problems to be solved for steady state operation on the future reactors and liquid metals (Li, Ga, Sn) could be a viable solution for the target materials. Since 2006 experiments by using a Capillary Porous System (CPS) Liquid Lithium Limiter (LLL) were successfully performed on FTU indicating a good capability of the system to sustain power loads. In order to prevent the overheating of the liquid Li surface and the consequent strong Li evaporation for T > 500 °C, an advanced version of LLL has been realized and installed on FTU by using the same vertical bottom port of the previous limiter. This new system, named Cooled Lithium Limiter (CLL), has been optimized to demonstrate the lithium limiter capability to sustain thermal loads as high as 10 MW/m^2 with up to 5s of plasma discharge duration. CLL includes an actively cooled system with water circulation at high pressure and is characterized by: 1) the small thickness of the Li CPS meshes (of W) placed in direct contact with the cooling tube (of Mo) and 2) the double role of water circulation at the temperature of about 200 °C, for heating lithium up to the melting point and for the heat removal during the plasma discharges. The heat load on the CLL is evaluated by different means: a fast infrared camera observing the whole limiter, the temperature measurements of inlet and outlet water as detected by the thermocouples and the measurements of the electron temperature and density by Langmuir probes placed on the CLL. Thermal loads analysis is performed by applying ANSYS code that has been adapted to the real CPS geometry and to the active cooling conditions of the new limiter. As first step, CLL has been tested in the FTU scrape-off layer to identify the best plasma conditions for a good uniformity of the thermal load by using the intensity of the Li and D atom emission at 670.8 nm and 656.3 nm to control Li production. The first experiments analyzed so far and simulated by ANSYS code, point out that heat loads as high as 2-3 MW/m^2 for 1.5s have been withstood without problems. Analysis is in progress for plasma discharges with CLL inserted deeper in the FTU scrape-off layer and close to the LCMS. In this paper the heat load measurements and their analysis will be reported and discussed to provide a clear understanding of CLL behavior under plasma discharges.
        Speaker: Dr Giuseppe Mazzitelli (ENEA - Unità Tecnica Fusione)
      • 14:00
        Toroidal Rotation Produced by Disruptions and ELMs 4h 45m
        In several experiments, including JET [1], Alcator C-Mod [2], and NSTX [3], it was observed that disruptions were accompanied by toroidal rotation. There is a concern that there may be a resonance between rotating toroidal perturbations and the resonant frequencies of the ITER vacuum vessel, causing enhanced damage. We present MHD simulations with M3D [4] of ITER, JET, and DIII-D, as well as theory, demonstrating that asymmetric vertical displacement event (AVDE) disruptions and ELMs can produce toroidal rotation. Net toroidal rotation requires three conditions [5]. (1) The poloidal magnetic field penetrates the wall, which is a condition that the plasma can transmit torque to the wall. (2) Rotation requires vertical asymmetry, which can be produced by a VDE. Simulations and theory indicate that the magnitude of the rotation is a stong function of VDE displacement. (3) Rotation requires MHD turbulence. In disruption simulations, the thermal quench and rotation generation occur at the same time, and are caused by toroidally varying MHD perturbations. The rotation persists into the current quench. This work was supported in part by the U.S.D.O.E., ITER and F4E. [1] S. N. Gerasimov, et. al, Proc. of EPS 37th Conference on Plasma Physics, Dublin, Ireland (2010). [2] R. S. Granetz, et. al, Nucl. Fusion 36, 545 (1996). [3] S. P. Gerhardt, et. al, Nucl. Fusion 52 063005 (2012). [4] W. Park, E. Belova, G. Y. Fu, X. Tang, H. R. Strauss, L. E. Sugiyama, Phys. Plasmas 6 1796 (1999). [5] H. Strauss, L. Sugiyama, R. Paccagnella, J. Breslau, S. Jardin, Nucl. Fusion, in press (2014).
        Speaker: Dr Henry R. Strauss (HRS Fusion)
        summary
    • 16:10 16:40
      Coffee Break 30m
    • 16:40 18:45
      Overview 5: Magnetic Fusion: OV/5 Blue 1-5

      Blue 1-5

      Hotel Park Inn Pribaltiyskaya

      Saint Petersburg, Russian Federation
      Convener: Mr Predhiman Kaw (India)
      • 16:40
        Overview of Gyrokinetic Studies on Electromagnetic Turbulence 25m
        Recent results on electromagnetic turbulence from gyrokinetic studies in different magnetic configurations are overviewed, showing the characteristics of electromagnetic turbulence and transport in situations where it is both expected and unexpected, and showing how it is affected by equilibrium magnetic field scale lengths. Ballooning parity ion temperature gradient (ITG) turbulence is found to produce magnetic stochasticity and electron thermal transport through nonlinear excitation of linearly stable tearing parity modes. The process is governed by nonlinear three-wave coupling between the ITG mode, the zonal flow, and the damped tearing parity mode. A significant electron thermal flux scales as beta squared, consistent with magnetic flutter. Above a critical beta known as the nonzonal transition, the magnetic fluctuations disable zonal flows by allowing electron streaming that effectively shorts zonal potential between flux surfaces. This leads to a regime of very high transport levels. A consideration of the residual flow in the presence of magnetic flutter confirms the disabling effect on zonal flows. Tearing parity microtearing modes become unstable in the magnetic geometry of spherical tokamaks and the RFP. They yield a growth rate in NSTX that requires finite collisionality, large beta, and is favored by increasing magnetic shear and decreasing safety factor. In the RFP, a new branch of microtearing with finite growth rate at vanishing collisionality is shown from analytic theory to require the electron grad-B/curvature drift resonance. However, when experimental MST RFP discharges are modeled gyrokinetically, the turbulence is remarkably electrostatic, showing trapped electron mode turbulence, large zonal flows, and a large Dimits shift. Analysis of the effect of the RFP’s shorter equilibrium magnetic field scale lengths shows that it increases the gradient thresholds for instability of trapped electron modes, ITG and microtearing. The stronger magnetic shear increases the beta threshold for kinetic ballooning mode (KBM) instability. This in turn increases the thresholds for magnetic activity, including the nonzonal transition.
        Speaker: Mr Paul Terry (USA)
        Slides
        summary
      • 17:05
        Overview of the RFX-Mod Contribution to the International Fusion Science Program 25m
        The RFX-mod device can be operated both as a Reversed Field Pinch (RFP), where advanced regimes featuring helical shape develop, and as a Tokamak. Due to its flexibility, RFX-mod is contributing to the solution of key issues in the roadmap to ITER and DEMO, including 3D nonlinear MHD modeling, MHD instability control, internal transport barriers, edge transport and turbulence, isotopic effect, high density limit. In RFP configuration, in the last two years advancements in the understanding of the self-organized helical states, featuring strong electron transport barriers, have been achieved; the role of microtearing modes in driving the residual transport at the barrier has been investigated experimentally and by gyrokinetic simulations. First experiments with deuterium as filling gas showed increased temperature and confinement time. New results on fast ion confinement and on the isotope effect on edge transport and turbulence are reported. RFX-mod contributed to the general issue of the high density limit physics, showing that in the RFP the limit is related to a toroidal particle accumulation due to the onset of a convective cell. The experimental program was accompanied by substantial progress in the theoretical activity: 3D nonlinear visco-resistive MHD and non-local transport modelling have been advanced; resistive wall and fast particle modes have been studied by a toroidal MHD kinetic hybrid stability code. In Tokamak configuration, q(a) regimes down to q(a)=1.2 have been pionereed, with (2,1) Tearing Mode (TM) mitigated and (2,1) RWM stabilized: the control of such modes can be obtained both by poloidal and radial sensors with proper control algorithm. Progress has been made in the avoidance of disruptions due to the locking of the (2,1) TM. External 3D fields have been applied to study a variety of physical issues: effect of magnetic perturbations on sawthooth control, plasma flow, runaway electron decorrelation. Probes combining electrostatic and magnetic measurements have been inserted to characterize turbulence and flow pattern at the edge.
        Speaker: Ms Maria Ester Puiatti (Italy)
      • 17:30
        Overview of Results from the MST Reversed Field Pinch Experiment 25m
        This overview of results from the MST reversed field pinch program summarizes physics important for the advancement of the RFP as well as for improved understanding of toroidal magnetic confinement in general. Topics include energetic particle effects, 3D helical equilibria, beta and density limit studies, microturbulence, ion heating, and magnetic self-organization physics. With neutral beam injection, several bursty energetic particle (EP) modes are observed. The profiles of the magnetic and density fluctuations associated with these EP-modes are measured using an FIR interferometer-polarimeter. Equilibrium reconstructions of the quasi-single-helicity 3D helical state are provided by the V3FIT code that now incorporates several of MST’s advanced diagnostics. A predator-prey theoretical model based on sheared flow and/or magnetic field has been developed that captures key QSH dynamics. Upgraded pellet injection permits study of density and beta limits over MST’s full range of operation, and an MST-record line-average density of 0.9E20 / m^3 (n/n_G =1.4) has been obtained. Plasma beta exhibits saturation at beta_tot ≤ 20% for a wide range of density, 0.2 < n/n_G < 1.6. Gyrokinetic simulations (GENE) based on experimental toroidal equilibrium reconstructions predict unstable trapped electron modes. Nonlinear simulations show that the “Dimits shift” is large and persists at finite beta. Experimentally, small-scale density fluctuations are detected in improved confinement plasmas. Impurity ion temperature measurements reveal a charge-to-mass-ratio dependence in the rapid heating that occurs during a sawtooth crash. Also, a toroidal asymmetry in the ion temperature is measured, correlated with 3D magnetic structure associated with tearing modes. Magnetic self-organization studies include measurements and modeling of the dynamo emf in standard RFP operation as well as with an applied ac inductive electric field to investigate the dynamics of oscillating field current drive (OFCD). Extended MHD computation for standard RFP conditions using NIMROD predicts dynamical coupling of current and plasma flow relaxation. The dynamo emf has also been measured when OFCD is applied, strengthening the understanding of and possibility for steady-state current sustainment using inductive current drive.
        Speaker: Dr Brett Chapman (USA)
        Slides
        summary
      • 17:55
        Overview of the FTU Results 25m
        Since the 2012 IAEA-FEC Conference, FTU operations have been largely devoted to runaway electrons (RE) generation and control, to the exploitation of the 140 GHz EC system and to liquid metal limiter elements. Experiments on RE have shown that the measured threshold electric field is larger than predicted by collisional theory and can be justified considering synchrotron radiation losses. A new RE control algorithm was developed and tested in presence of a RE current plateau, allowing to minimize the interactions with plasma-facing components and safely shut down the discharges. The experimental sessions with 140 GHz EC system have been mainly devoted to experiments on real time control of MHD instabilities using the new EC launcher with fast steering capability. Experiments with EC power modulation have confirmed the possibility to lock the sawtooth period to the EC period, with EC injection inside the q=1 surface, while experiments with central EC injection have shown the onset of 3/2 and 2/1 modes. EC assisted breakdown experiments have been focussed on ITER start-up issues, exploring the polarization conversion at reflection from inner wall and the capability to assure plasma start-up even in presence of a large stray magnetic field. A new actively Cooled Lithium Limiter (CLL) has been installed and tested. The CLL was inserted close to the last closed magnetic surface, without any damage to the limiter surface, and first elongated FTU plasmas with EC additional heating were obtained with the new CLL. Reciprocating Langmuir probes were used to measure the heat flux e-folding length in the scrape-off layer, with the plasma kept to lay on the internal limiter to resemble the ITER start-up phase. Density peaking and controlled MHD activity driven by Neon injection were investigated at different plasma parameters, and a full real-time algorithm for disruption prediction, based on MHD activity signals from Mirnov coils, was developed exploiting a large database of disruptions. New diagnostics were successfully installed and tested, as a gamma camera for RE studies and a diamond probe to detect Cherenkov radiation produced by fast electrons. Laser Induced Breakdown Spectroscopy measurements were performed under vacuum, so demonstrating the possibility to provide useful information on the fuel retention in present and future tokamaks, such as ITER.
        Speaker: Mr Gianluca Pucella (Italy)
        Slides
        summary
      • 18:20
        First Experiments in SST-1 25m
        Steady State Superconducting Tokamak (SST-1) has been commissioned after the successful experimental and engineering validations of its critical sub-systems. During the `engineering validation phase’ of SST-1; the cryostat was demonstrated to be leak tight to superconducting magnets system operations in all operational scenarios, the 80 K thermal shield was demonstrated to be uniformly cooled without regions of `thermal run away and hot spots’, the superconducting Toroidal Field (TF) magnets were demonstrated to be cooled to their nominal operational conditions and charged up to 1.5 T of field at the major radius, the assembled SST-1 machine shell was demonstrated to be a graded, stress-strain optimized and distributed thermo-mechanical device and the integrated vacuum vessel was demonstrated to be UHV compatible etc. Subsequently, `field error components’ in SST-1 were measured to be acceptable towards plasma discharges. A successful break-down in SST-1 was obtained in SST-1 in June 2013 assisted with electron cyclotron pre-ionization in second harmonic mode, thus marking the `First Plasma’ in SST-1 and arrival of SST-1 into the league of contemporary steady state devices as well. Subsequent to the first plasma, both physical experiments and boosting of engineering parameters in SST-1 have begun. A successful plasma start-up with E ~ 0.4 V/m, plasma current in excess of 50 kA for 100 ms assisted with ECH pre-ionization in second harmonic at a field of 0.75 T have been achieved. Lengthening the plasma pulse duration with LHCD, plasma current boosting up with ECH assisted pre-ionization in fundamental mode at 1.5 T apart from advance plasma physics experiments are presently being attempted in SST-1. In parallel, SST-1 has demonstrated in unique fashion pure cold gas cooling based nominal operations of its vapour cooled TF current leads up to 4650 A corresponding to 1.5 T of field in the plasma major radius. SST-1 has also achieved the distinction of being the only superconducting Tokamak in the world where the cable-in-conduit-conductor (CICC) based TF magnets are operated with helium cooling in Two-Phase mode during the plasma discharges up to 2.0 T of field at the plasma major radius.
        Speaker: Mr Subrata Pradhan (India)
        Slides
        summary
    • 19:30 22:00
      Nuclear Fusion Board Meeting Blue 1-5

      Blue 1-5

      Hotel Park Inn Pribaltiyskaya

      Saint Petersburg, Russian Federation
    • 08:30 10:15
      Heating & Disruption: FIP/2 Blue 1-5

      Blue 1-5

      Hotel Park Inn Pribaltiyskaya

      Saint Petersburg, Russian Federation
      Convener: Mr Alexander Litvak (Russian Federation)
      • 08:30
        Disruption Mitigation System Developments and Design for ITER 20m
        Disruptions present a challenge for ITER to withstand the intense heat flux, the large forces from halo currents, and the potential first wall damage from multi-MeV runaway electrons. Injecting large quantities of material into the plasma when a disruption is detected will reduce the plasma energy and increase its resistivity and electron density to mitigate these effects and thus a system with this capability is needed for maintaining successful operation of ITER. A disruption mitigation system is under design for ITER to inject sufficient material deeply into the plasma for a rapid shutdown and runaway electron collisional suppression. Here we present progress on the development and design of both a shattered pellet injector that produces large solid cryogenic pellets to provide reliable deep penetration of material [1] and a fast opening high flow rate gas valve for massive gas injection. The shattered pellet injector utilizes a multi-barrel pipe-gun type device that forms large cryogenic pellets in-situ in the barrels. The pellets are accelerated by a high pressure gas burst and are shattered when they impinge on a bend guide tube in the port plug shield block that is optimized to produce a spray of solid fragments mixed with gas and liquid at speeds approaching the sound speed of the propellant gas. A prototype injector has been fabricated and tested with deuterium pellets of 16 mm size for thermal mitigation and is being upgraded to test and characterize 25 mm size D2 and neon pellets for runaway electron suppression. A fast opening high flow rate gas valve for massive gas injection has been designed for use in the ITER environment, which requires a novel eddy current flyer plate design and large diameter tritium compatible seat material as compared to earlier DMV designs by Juelich [2]. Modeling of the gas flows from the valves through guiding tubes gives a response time for the MGI design to be less than desired unless the valves are mounted within port plugs. Implications of the design with respect to response time and reliability are discussed. [1] N. Commaux, et al., Nucl. Fusion 50 (2010) 112001. [2] S.A. Bozhenkov, et al., Rev. Sci. Instrum. 78 (2007) 033503.
        Speaker: Mr So Maruyama (ITER Organization)
        Slides
      • 08:50
        Prototype Development of the ITER EC System with 170GHz Gyrotron & Development of Dual Frequency Gyrotron and Launcher for the JT-60SA ECH/ECCD System & Development of Over 1 MW and Multi-Frequency Gyrotrons for Fusion 20m
        A To study the operational performance of ITER EC heating and current drive system (H&CD), a mock-up of the ITER mm wave system has been assembled using the high power long gyrotron test stand in JAEA. The prototype system is composed of the primary parts of the EC H&CD system, including: 170GHz gyrotron, power supply, transmission line (TL) and mock-up of equatorial launcher (EL) and control system. The gyrotron power was transmitted via the precise aligned TL (40m) with 7 miter bends to the EL achieving a 91% of HE11 mode purity. The experiments were realized using a mock-up of the conceptual EC control system based on the ITER Plant Control Design Handbook (PCDH). The system has achieved CW 5 kHz power switching, which demonstrates the compatibility for MHD control of ITER plasma. The modulation was achieved using a novel configuration of the electron beam acceleration power supply. In the experiment, stable 5 kHz of power modulation was demonstrated with minimized spurious frequency excitation at the ramp-up phase of each pulse., which satisfied the ITER criteria. The JAEA test stand is a flexible system with its center piece a frequency-step-tunable gyrotron at 170GHz/137GHz/104GHz. The output beam is radiated to the identical direction from the output window for each frequency, consequently the power was transmitted to the end of the TL at these three frequencies. B The development of a gyrotron and launcher operated at two frequencies, 110 GHz and 138 GHz, has made a significant progress toward electron cyclotron heating (ECH) and current drive (ECCD) in JT-60SA. High-power, long-pulse gyrotrons are required for the JT-60SA ECH/ECCD system which has the total injection power of 7 MW and the pulse duration of 100 s using 9 gyrotrons. The wave frequency in the original specification is 110 GHz, which is effective for off-axis ECH/ECCD to sustain a high-beta plasma at the toroidal field of 1.7 T. On the other hand, the higher frequency waves at 130 ~ 140 GHz enables ECH/ECCD in the core plasma region at the maximum toroidal field of 2.3 T in JT-60SA. However, a dual frequency gyrotron that can generate the target output power and pulse length (1 MW for 100 s) was not developed since it requires high oscillation efficiency to obtain high power and low diffraction loss to achieve long pulse, simultaneously. In 2011, we stared to develop a new dual frequency gyrotron (110 GHz, 138 GHz) equipped with a triode type electron gun to obtain high oscillation efficiency. High-power, long-pulse operations of the dual frequency gyrotron have been carried out since the last IAEA FEC. Developments of an ECH launcher with high reliability based on a linear-motion antenna concept and a polarizer with optimized groove depth, width, and period for dual frequency operation are also in progress. Main results are as follows: (i) Oscillations of 1 MW for 10 s were successful at both frequencies for the first time in the world as a dual-frequency gyrotron by optimizing electron pitch factor using the triode electron gun; (ii) Low diffraction loss and cavity Ohmic loss enabling 1 MW for 100 s and 1.5 - 2 MW for several seconds were experimentally confirmed, and a 100 MJ oscillation was achieved (0.51 MW, 198 s, 110 GHz), so far; (iii) An oscillation at 82 GHz was also successful as an additional frequency showing the possibility of the use of fundamental harmonic waves; (iv) Launcher optics design toward dual-frequency operations showed little difference in the poloidal beam width for these frequencies; (v) Prototype tests of a wide-band twister polarizer at both low power (< 1 mW) and high power (~0.25 MW, 3 s) showed promising results. C EC (Electron Cyclotron) scheme is quite promising tool for heating and current drive (H&CD) and plasma control for present and future devices up to Demo and Commercial reactors. Development of gyrotron is a key to open this promising door. Multi-MW and multi-frequency technologies are major issues to challenge for robust and cost effective reactor heating system. In University of Tsukuba, gyrotrons of wide range of frequencies from 14 GHz to 300 GHz have been developed for this purpose in collaboration with JAEA, NIFS and TETD. Over-1 MW dual frequency gyrotron of new frequency range (14 – 35 GHz), where the reduction of diffraction loss and cathode optimization are quite important, has been developed for EC/EBW H&CD for GAMMA 10/PDX, QUEST, Heliotron J and NSTX-U. Output power of 1.25 MW at 28 GHz and estimated oscillation power of 1.2 MW at 35.45 GHz from the same tube have been achieved with the cathode angle improvement. This is the first demonstration of the over 1 MW dual-frequency operations in lower frequency, which contributes to the technology of wide band multi-frequency/multi-MW tube for Demo EC/EBW H&CD. The output power of 600 kW for 2 s at 28 GHz is also demonstrated. It is applied to the QUEST and has resulted higher EC-driven current than ever. Further, in the joint program of NIFS and Tsukuba for LHD ECH gyrotrons, a new frequency of 154 GHz has been successfully developed with a TE_28,8 cavity, which delivered 1.16 MW for 1 s, 0.3 MW in CW (30 minutes), and the total power of 4.4 MW to LHD plasma with other three 77 GHz tubes, which extended the LHD plasma to high T_e region. All these gyrotron performances are new records in each frequency range.
        Speaker: Mr Yasuhisa Oda (Japan)
        Slides
      • 09:10
        ICRF Actuator Development at Alcator C-Mod 20m
        Future fusion reactors will present more severe constraints on ion cyclotron range of frequency (ICRF) heating and current drive actuators than ITER. Reliably coupling power to the plasma despite load variations is critical. In addition, ICRF interaction with the edge plasma, particularly impurity contamination and enhanced localized heat loads, is challenging. We report on progress developing an ICRF actuator with favorable scaling towards reactors. Using a field aligned (FA) antenna, we have found that the FA antenna loading is similar to TA antennas but the FA antenna reflection coefficient has significantly reduced variation, thus it is inherently load tolerant. We speculate the variation in reflection coefficient is a result of slow wave coupling of neighboring straps and field alignment significantly reduces this coupling. The underlying physics of RF plasma edge interaction is thought to be linked to RF electric fields parallel to the magnetic field, E||. One source of RF E|| is from the antenna itself and can minimize integrated E|| through geometry. Experiments comparing a field aligned (FA) and a toriodally aligned (TA) antenna have demonstrated that FA antenna has significantly reduced impurity contamination compared to TA antennas. The impurity sources measured at the antenna are nearly eliminated for the FA antenna. This is an important milestone since this is the first demonstration that an ICRF antenna can be made with reactor compatible materials. Furthermore, the heat flux to the FA antenna is reduced to a level similar to that observed for identical discharges heated by the TA antenna and the FA antenna is not powered. The estimated energy deposited is 0.4% of the total injected energy and marks the first time an ICRF antenna has achieved the target level for the ITER design, 0.625% of 20 MW. One path to increase antenna power density is to use materials with high strength and high melting temperature. Furthermore copper will be restricted to thin coatings in a reactor due to material swelling and poor strength at high temperature. We have found that the higher strength materials have higher breakdown voltage compared to copper. Highly polished molybdenum and tungsten breakdown field is 40% higher than copper. The latest results and analysis will be presented.
        Speaker: Mr Stephen Wukitch (USA)
        Slides
      • 09:30
        Progress Status of the Activities in EU for the Development of the ITER Neutral Beam Injector and Test Facility 20m
        The development of the Neutral beam system for ITER has been progressing well thanks to the start of the operations of the ELISE (Extraction from a Large Ion Source Experiment) at the Max Planck Institute for Plasma Physics in Garching, Germany, and to the big effort devoted to the establishment of the ITER Neutral Beam (NB) Test Facility in Padua, Italy. This paper presents the main experimental results of ELISE, the status of the manufacturing activities for the components of the NB test facility and the progress made in the design of the mechanical components and in that of auxiliary and power supplies systems.
        Speaker: Mr Antonio Masiello (European Commission)
        Slides
        summary
      • 09:50
        Development of DC Ultra-High Voltage Insulation Technology for ITER NBI & Progress in Long Pulse Production of Powerful Negative Ion Beams for JT-60SA and ITER 20m
        A In the ITER NBI for plasma heating and current drive, a 1 MeV, 40 A deuterium negative ion (D-) beam is designed to be accelerated for 3600 s. The beam energy and the pulse duration of the D- beam are 2-5.5 times higher and 360 times longer than those in the negative-ion-based NBIs on LHD and JT-60U, respectively. Thus, to realize higher voltage and longer pulse duration, the generation, transmission and insulation of DC ultra-high voltage are critical issues for the ITER NBI. In addition, the high-current busbar, cooling water and gas pipes at -200 kV~ -1 MV potential are simultaneously transmitted through the HV bushing from the gas-insulated transmission line in the PS to the beam source (BS) to minimize the installation space. Those are significant differences from that in existing N-NB systems. Especially, a DC 1 MV insulating transformer for feeding an electric power from the ground to 1 MV potential is one of most challenging components. For stable power transmission through the HV bushing, 1 MV vacuum insulation and a stiffness to withstand the maximum pressure difference of 0.9 MPa in a limited space are required. Japan Atomic Energy Agency (JAEA) is in charge of the procurement of these high voltage parts of the 1 MV PS and the HV bushing. As for the insulating transformer, a DC long pulse insulation structure and a composite bushing for the isolation of the high-voltage to the air have been newly developed. The mockup transformer successfully demonstrated a stable insulation of DC -1.2 MV for 3600 s. The HV bushing serves as the terminal of the HV transmission line. It is made in five 200 kV stages and a two-stage mockup has been developed, and stable voltage holding at 480 kV for 3600 s was demonstrated. These R&D results fulfill the ITER requirement, which allows the realization of the PS and HV bushing for the ITER NBI. B The long pulse generation of the powerful negative ion beams of 500 keV, 22 A (130 A/m^2) and 1MeV, 40 A (200 A/m^2) is the essential challenge to realize the negative-ion-based neutral beam injectors (NBIs) for JT-60SA and ITER, where 10 MW D0 beam for 100 s and a 16.5 MW for 3600 s are designed, respectively. In Japan Atomic Energy Agency (JAEA), after the achievements of the beam current density and energy required for JT-60SA and ITER with a short pulse duration, the target of R&D is focused on the extension of the pulse duration in JT-60 negative ion source and the MeV accelerator. Significant progress in the extension of pulse duration of the powerful negative ion beams has been made to realize the neutral beams injectors for JT-60SA and ITER. The pulse duration and the current density of the JT-60 negative ion source has been successfully improved from 30 s at 80 A/m^2 in the previous operation to 100 s at 120-130 A/m^2, which satisfy the rated values for JT-60SA. This progress has been achieved by controlling the negative ion production via the surface temperature of the plasma grid. The pulse duration of the MeV class negative ion beams for ITER has been also extended by more than an order of magnitude in the MeV accelerator. A long pulse acceleration of 8.7 s has been achieved at 880 keV, 130 A/m^2 by improving the cooling capability of the extraction grid where the aperture displacement for the beamlet steering is also modified, so there is no limitation to increase the power density and the pulse duration. This is the longest pulse duration of the MeV-class negative ion beams in the world.
        Speaker: Mr Hiroyuki Tobari (Japan)
        Slides
    • 08:30 12:30
      Poster 3: P3 Green 8-9

      Green 8-9

      Hotel Park Inn Pribaltiyskaya

      Saint Petersburg, Russian Federation
      • 08:30
        A Systematic Approach to the Linear-Stability Assessment of Alfvén Eigenmodes in the Presence of Fusion-Born Alpha Particles for ITER-like Equilibria 4h
        A systematic approach to assess the linear stability of Alfvén eigenmodes in the presence of fusion-born alpha particles is described. Because experimental results for ITER are not available yet, there is no guidance about which Alfvén eigenmodes will interact more intensively with the fast-particle population. Therefore, the number of modes that need to be considered in stability assessments becomes quite large and care must be exercised when choosing the numeric tools to work with, which must be fast and efficient. In the presented approach, the eigenmodes are first found after an intensive scan of a suitable frequency range, performed within reasonable bounds for the toroidal and poloidal mode numbers. Each solution found is then tested to find if its discretization over the radial grid in use is adequate. Finally, the interaction between the identified eigenmodes and the alpha-particle population is evaluated with the drift-kinetic code CASTOR-K, in order to assess their growth rates and hence their linear stability. The described approach enables one to single out the most unstable eigenmodes in a given scenario, which can then be handled with more specialized tools. This ability eases the task of evaluating alpha-particle interactions with Alfvén eigenmodes, either for ITER scenarios or for any other scenario planning.
        Speaker: Dr Paulo Rodrigues (Instituto de Plasmas e Fusao Nuclear)
        summary
      • 08:30
        Advancing Power Exhaust Studies from Present to Future Tokamak Devices 4h
        Power exhaust is a crucial issue for future fusion devices such as ITER and DEMO. A device like DEMO despite being of a similar geometrical size of ITER will need to accommodate an about 3 to 4 times higher thermal power, aggravating the issue of power exhaust. ASDEX Upgrade with its fully tungsten covered wall, high ratio of heating power to major radius and extensive edge and SOL diagnostics is well suited for studying most aspects of power exhaust. Limiting the total power flux to the divertor target plates is only possible in the detached regime. Despite its crucial importance for safe operation of future larger devices the understanding of the processes leading to divertor detachment is incomplete. The paper summarizes the efforts undertaken in gradually advancing the understanding of power exhaust in a variety of conditions: It presents how the H-mode density limit is controlled by a fuelling limit and an enhanced loss of power at the plasma edge. The power fall off length in the divertor is determined by the volumetric dissipation in the divertor connected to the recycling of neutrals and consequently to the divertor geometry. Experiments with nitrogen as seeding impurity for L-mode and H-mode are used for validating the SOLPS code package. In such studies the activation of drift terms in the numerical model is crucial for minimizing the differences to the experimental data. The movement of the radiation in the divertor under varying conditions is explained and maximum radiation is reached with stable radiation in the vicinity of the X-point. A phase of strong fluctuating radiation in the vicinity of the X-point on the high field side of the divertor is identified as a condition for strongest discrepancy between the numerical and experimental results. Studies on the snowflake as an alternative divertor geometry solution using the EMC3-EIRENE code are also presented.
        Speaker: Dr Marco Wischmeier (Max-Planck-Insitut für Plasmaphysik)
      • 08:30
        Alfvén Eigenmode Evolution in ITER Steady-State Scenario 4h
        Alfvén eigenmode instability analysis in ITER steady-state plasma scenarios with reversed magnetic shear was performed with the NOVA and TAEFL codes [1]. In our work for this scenario we explore the stability of Alfvén eigenmodes with the KINX code [2]. Both isotropic fusion alphas and beam ions contribute into the mode drive. Fast particle dynamics, linear growth rate, mode amplitude evolution and the wave nonlinear saturation level are computed with the VENUS+df [3] orbit following code. Anysotropic beam particle distribution is computed from realistic geometry of ITER NBI. Calculation results give the estimations of the Alfvén stability linear growth rates and nonlinear saturation level of the mode amplitude for ITER steady state scenario. [1] M.A. Van Zeeland et al, Nucl. Fusion 52(2012)094023. [2] L. Degtyarev et al, Comp. Phys. Comm., 103(1997)10. [3] W.A. Cooper et al, Plasma Phys. Contr. Fus. 53(2011)024001.
        Speaker: Dr Maxim Isaev (Kurchatov Institute)
      • 08:30
        Assessment of Operational Space for Long-Pulse Scenarios in ITER 4h
        Operational space (I_p-n) for long pulse scenarios (t_burn ~ 1000 s, Q > 5) foreseen in ITER was assessed by 1.5D core transport modelling with pedestal parameters predicted by the EPED1 code. The analyses include the majority of transport models (CDBM, GLF23, Bohm/GyroBohm (BgB), MMM7.1, MMM95, Weiland, Scaling-Based) presently used for interpretation of experiments and ITER predictions. The EPED1 code was modified to take into account boundary conditions predicted by SOLPS for ITER. In contrast with standard EPED1 assumptions the EPED1 with the SOLPS boundary conditions predicts no degradation of the pedestal pressure with density reduction. Reducing the plasma density to n_e ~ 5-6 10^19m^-3 leads to an increased plasma temperature (similar pedestal pressure) which reduces the loop voltage and increases the duration of the burn phase to t_burn~ 1000 s with Q > 5 for I_p > 13 MA at moderate normalised pressure, beta_N ~ 2 in ITER. These ITER plasmas require the same level of additional heating power as the reference Q = 10 inductive scenario at 15 MA (33 MW NBI and 17 - 20 MW EC heating and current drive power). However, unlike the ‘hybrid’ scenarios considered previously, these H-mode plasmas do not require specially shaped q profiles nor improved confinement in the core for the transport models considered in this study. Thus, these medium density H-mode plasma scenarios with I_p > 13 MA present an attractive alternative to hybrid scenarios to achieve ITER’s long pulse Q > 5 and deserve further analysis and experimental demonstration in present tokamaks.
        Speaker: Dr Alexei R. POLEVOI (ITER Organization)
        summary
      • 08:30
        Asymmetry Current in ICRF Heating ITER Plasmas 4h
        The possibilities of using transverse ICRF heating tokamak plasma minorities for toroidal current driving is investigated in this paper. Three ways of this heating utilization are proposed. Firstly, such heating gives possibility to drive the seed current near magnetic axis, secondly, it can be used for safety current profile adjusting (the negative shear producing) due to synchronous heating of two types of minorities – hydrogen and helium ions, and , third, it can be used for non-inductive asymmetry toroidal current drive in plasma cross-section. Unlike to the isotropic heating when the ratio between amount of trapped and untrapped particles is conserved, during the transverse heating almost all particles become trapped and precisely trapped particles drive toroidal current due to asymmetry ion motion in and against inductive current direction. For the asymmetry current calculation theoretical distribution function of the minority proposed by T.H.Stics and distribution function measured in JET tokamak were used. Fulfilled estimations show that the transverse ICRF heating hydrogen minority up to energy in several MeV that is possible to have the toroidal current in the mega-amperes range.
        Speaker: Mr Yury Gott (Kurchatov Institute)
        Poster
        summary
      • 08:30
        Attainment of High Electron Poloidal Beta in Axisymmetric State and Two Routes to Self-Organized Helical State in Low-Aspect-Ratio RFP 4h
        Improvement of plasma performance has been advanced in the low-aspect-ratio (low-A) reversed field pinch (RFP) RELAX, whose main objectives include exploring the low-A RFP configuration. In axisymmetric RFP states in deep-reversal region, it is found that central electron poloidal beta, beta_p (=p_e0/(B_pa^2/(2mu_0))), which almost equals the electron beta in the RFP, has reached to 5~10%. Feedback control using saddle coil array is applied to stabilize a single resistive wall mode (RWM). As a result, fine tuning of the equilibrium becomes effective in achieving the resultant beta_p of higher than 10%. The attained parameter region is close to where we expect sizable fraction of the bootstrap current which is characteristic to low-A RFP configuration. In shallow-reversal region characterized by relaxation to the quasi-single helicity (QSH) state, soft-X ray (SXR) computed tomography (CT) technique has revealed helically deformed m/n=1/4 structure of SXR emissivity profile. 3-D MHD simulation using the MIPS code has shown two possible routes to the self-organized helical state depending on the initial equilibrium; one through core-resonant tearing mode, and the other through internally non-resonant kink mode. The self-organized helical structure agrees well with experimental results.
        Speaker: Prof. Sadao Masamune (Kyoto Institute of Technology)
        summary
      • 08:30
        Basic Investigations of Turbulence and Interactions with Plasma and Suprathermal Ions in the TORPEX Device with Open and Closed Field Lines 4h
        TORPEX is a flexible device dedicated to investigating basic plasma physics phenomena of importance for fusion. It can feature a simple magnetized toroidal (SMT) configuration with a dominant toroidal magnetic field and a small vertical field component, or accommodate closed field-line configurations using a current-carrying conductor suspended in the center of the chamber. This produces a poloidal magnetic field with a rotational transform, which, combined with vertical field coils, allows magnetic configurations of increasing complexity and of more direct relevance to confined plasma experiments. Among these are simple plasmas limited by the vessel on the low field side, single or double-null X-points, and even advanced divertor configurations like snowflakes. Using an extensive set of diagnostics, systematic studies of plasma instabilities, their development into turbulence and meso-scale structures, and their effects on both thermal and suprathermal plasma components are performed. The impact of the experimental results obtained on TORPEX is enlarged by their systematic application to model validation, performed using rigorous methodologies for quantitative experiment-theory comparisons. In the past two years, we conducted investigations of supra-thermal ion-turbulence interaction, a basic issue for burning plasmas, on SMT plasmas. These investigations reveal that the transport of supra-thermal ions is generally non-diffusive and can be super- or sub-diffusive depending on two parameters: the suprathermal ion energy normalized to the electric temperature and the electric potential fluctuations normalized to the electron temperature. The orbit averaging mechanism predicted to reduce the effect of turbulence on the suprathermal ions in burning plasmas has been clearly identified, both for gyro- and drift-orbits. To better mimic the SOL-edge magnetic geometry in tokamak, we have installed a new system that creates twisted field line configurations. First experiments are devoted to the characterization of the background plasma and fluctuation features in the presence of quasi circular-shaped flux surfaces. Measurements of toroidal and poloidal wave numbers indicate field aligned modes. Further studies are under way to compare the experimental measurements with the simulation results and assess the main instability driving mechanism.
        Speaker: Dr Ivo Furno (EPFL- CRPP)
        summary
      • 08:30
        Comparative Studies of Edge Magnetic Islands and Stochastic Layers in DIII-D and LHD 4h
        Joint experiments on the DIII-D tokamak and the LHD stellarator/heliotron have resulted in the discovered of spontaneous heat transport bifurcations across the O-point of an applied m/n=2/1 magnetic islands in DIII-D and enhanced particle transport relative to heat transport in edge m/n=1/1 LHD isalnds. The DIII-D results suggest that the heat transport bifurcations are due to islands transitioning from smooth flux surfaces to partially stochastic layers. Alternatively, measurements of the particle and heat transport inside edge static magnetic islands in LHD plasmas show an enhancement of the particle flux relative to the heat flux. The DIII-D results suggest that externally applied static 3D magnetic fields can produce a dynamic evolution of the magnetic topology in the plasma due to a nonlinear toroidal coupling of resonant modes on various rational surfaces while the LHD results show that edge magnetic islands preferentially increase the particle flux relative to the heat flux for reasons that have yet to be clarified. Static magnetic islands and stochastic layers have been observed in low-β L-mode plasmas but not in diverted H-mode plasmas yet. In order to understand the physics of non-axisymmetrically perturbed high-β fusion H-mode plasmas, such as the mechanisms involved in edge localized mode (ELM) suppression with resonant magnetic perturbation (RMP) fields, it is necessary to determine if islands and stochastic layers exists and whether they are static or evolve in time due to the plasma response. Measurements in DIII-D show spontaneous transitions of the magnetic field on rational surfaces due to the plasma response. For example, flat spots in the T_e profile associated with m/n = 2/1, 3/1 and 4/1 islands are seen to appear and disappear as the discharge evolves suggesting either a time varying screening of the field by the plasma or a nonlinear coupling of the 2/1, 3/1 and 4/1 islands. In this contribution we discuss the measurements made in DIII-D along with transport results due to pellets injected into static islands in LHD and their implications for understaning the plasma response to 3D fields in H-mode plasmas. This work was supported in part by the US Department of Energy under DE-FC02-04ER54698, DE-AC05-00OR22725, DE-FG03-97ER54415, and the NIFS budget code NIFS11ULHH021.
        Speaker: Mr Todd E. Evans (USA)
      • 08:30
        Core Plasma Rotation Characteristics of RF-Heated H-Mode Discharges on EAST 4h
        Access to high-confinement plasmas on the EAST tokamak was readily achieved through LHCD, ICRF, or their combined application along with improved wall conditioning and wave-plasma coupling capabilities. Using a tangentially viewing X-ray crystal spectrometer, core plasma rotation profiles and their temporal evolutions were obtained. This paper presented typical plasma rotation behaviors for non-stationary and stationary H-mode discharges generated with concurrent LHCD and ICRF heating. A substantial increase of the co-current core rotation was observed at L-H transitions. For unsteady discharges with multiple L-H and H-L transitions, central rotation velocity varied as the plasma entered and left the H-mode phase. For stationary ELMy H-mode discharges, the rotation increases at an L-H transition and core plasma rotation profile remains very stable during the entire H-mode phase, although the occurrence of ELMs tended to slow down the core rotation. Changes of the steady-state core rotation at L-H transitions were found to be dependent on the plasma parameters for different ELM types. A linear relation between the rotation and stored energy, similar to the Rice scaling was obtained for both ELM-free and ELMy H-mode discharges; and for ELMy-free discharges the slope was by a factor of 1.75 steeper. This work was supported by National Magnetic Confinement Fusion Science Program of China (No. 2011GB101004, 2011GB107001 and 2013BG112004), National Science Foundation of China (No. 11175208, 11305212 and 11375235) and JSPS-NRF-NSFC A3 Foresight Program in the field of Plasma Physics (No. 11261140328).
        Speaker: Dr Bo Lyu (Institute of Plasma Physics, Chinese Academy of Sciences)
        summary
      • 08:30
        Design and First Applications of the ITER Integrated Modelling & Analysis Suite 4h
        The ITER Integrated Modelling & Analysis Suite (IMAS) will support both plasma operation and research activities on the ITER tokamak experiment. The IMAS will be accessible to all ITER Members as a key tool for the scientific exploitation of ITER. It will allow collective development of Integrated Modelling tools, by sharing data, code components and, ultimately, workflows based on coupling together various code components. Its design started in 2011 and a first prototype of the IMAS infrastructure has already been implemented at the ITER Organization (IO). The purpose of this paper is to describe the essential features of the IMAS design, the implemented prototype, as well as the first physics applications which have been developed under the IMAS infrastructure. The IMAS infrastructure is based on a standardized data model that covers experimental and simulated data with the same representation. The standard data model is device-generic and can be used to describe data from existing experiments. Since the data model will progressively cover a large number of areas (plasma, diagnostics, actuators, other tokamak subsystems, …) and will be developed by many contributors, a set of data model design rules and guidelines have been established to ensure consistency and homogeneity of the data model. Physics components, once interfaced to the data model, can be coupled into an Integrated Modelling workflow orchestrated by a workflow engine. A first implementation of all these infrastructure elements has been carried out and is described in this paper. First applications have been integrated under the prototype IMAS infrastructure to allow their performance to be tested and to demonstrate the expected functionalities of the infrastructure. Transport solvers with free boundary equilibrium capabilities have been integrated to the IMAS infrastructure, namely CORSICA and DINA. ITER pulse simulations have been carried out by coupling the Plasma Control System (PCS), which in ITER has a dedicated and distinct simulation platform based on the Simulink® software, to the physics solvers. This has required the development of an original co-simulation technique between the plasma and plant simulator (under IMAS) and the dedicated PCS simulator. First results of full tokamak simulations under the IMAS infrastructure will be described in the paper.
        Speaker: Dr Frédéric Imbeaux (CEA-IRFM)
        summary
      • 08:30
        EAST Snowflake Experiment: Scenario Development and Edge Simulations 4h
        Snow Flake Divertor (SFD) configuration has been proposed as one valid way to reduce the plasma-wall interaction. For technological reasons, SFD configuration is difficult to realize and control in real experiments, especially for a tokamak like EAST that does not have dedicated divertor coils designed to locally shape the magnetic field topology. For this reason, Quasi-SFD (QSF) static equilibrium configurations (where two poloidal field Bp=0 points are close enough to produce a large region with poloidal field close to zero in the divertor region) have been studied by using EFIT and FIXFREE equilibrium codes. The tokamak simulation code (TSC), a numerical model of the axisymmetric tokamak plasma and the associated control systems, has been then used to model the EAST QSF scenario (300kA/1.8T). During the simulation, iso-flux control scheme is used to control plasma shape, and the poloidal field (PF) coils current is limited to be smaller than 15% of the actual allowed technical limits. TSC outputs will be preliminary used to set the plasma control system (PCS) operating during the experiments. CREATE-NL tools has been used to linearize the configuration, in order to increase the QSF to higher plasma current, and a dedicated control algorithm will be developed to use magnetic topology as a control actuator of the local radiation in presence of impurities seeding. The analysis of vertical stability growth rates of QSF configurations with 2D and 3D models is ongoing. Preliminary results on the comparison of edge simulations between the standard divertor (SD) and SFD will be also presented. First QSF experiment will be performed during next EAST restart.
        Speaker: Dr Giuseppe Calabro (ENEA - Italy)
        summary
      • 08:30
        Effect of Multi-Pass Absorption of Electron Cyclotron Heating Wave on Initial Stage of Discharge in ITER-like Tokamak 4h
        A model is suggested for calculating the efficiency of multi-pass absorption of EC heating wave in tokamaks at initial stage of discharge, and the effect on the start-up in ITER-like tokamak is analyzed. The single-pass absorption of injected EC wave is evaluated with the scaling obtained using the OGRAY code. The model for subsequent multi-pass absorption, after first reflection of the EC wave from the wall of vacuum chamber, assumes isotropy/uniformity of the respective EC radiation intensity in plasma. The model modifies the CYNEQ code approach developed for the plasma-produced EC radiation transport at high EC harmonics and verified in the benchmarking with other codes. We consider the following case: (a) multiple reflection of injected EC wave (O-mode) from the wall; (b) polarization scrambling in wall reflections; (c) full single-pass absorption of the X-mode. Our parametric analysis for typical electron temperature and density at initial stage of discharge in ITER-like tokamak shows strong dependence of multi-pass absorption efficiency on the O-X conversion in wall reflections. The multi-pass absorption model is incorporated in the 1-D simulations of plasma start-up with the DINA code that enables us to extend the results of previous simulations with the single-pass absorption model.
        Speaker: Mr Pavel Minashin (NRC "Kurchatov Institute")
        Poster
        summary
      • 08:30
        Effect of Resonant Magnetic Perturbations on Low Collisionality Discharges in MAST and a Comparison with ASDEX Upgrade 4h
        The application of Resonant Magnetic Perturbations (RMPs) is foreseen as a mechanism to ameliorate the effects of ELMs on the ITER divertor. Various aspects of RMP operation crucial to ITER have been demonstrated on MAST such as mitigating the first ELM after the L-H transition, sustaining ELM mitigation during both the current ramp-up and in the event of failure of a sub-set of the in-vessel coils and applying a slowly rotating n=3 RMP, which sustains ELM mitigation while rotating the pattern of the strike point splitting. Although ELM suppression has not been observed on MAST, ELM mitigation has been achieved using RMPs with toroidal mode number of n=2, 3, 4 and 6 over a wide region of operational space, with considerable overlap with the regions where suppression of type-I ELMs is observed in other machines. The effect that the choice of toroidal mode number on the effectiveness of the mitigation has been investigated and shows that n=3 or 4 is optimal. The ELM mitigation phase is typically associated with a drop in plasma density and overall stored energy. By carefully adjusting the refuelling, either by gas or pellet fuelling, to counteract the drop in density it has been possible to produce plasmas with mitigated ELMs, reduced peak divertor heat flux and with minimal degradation in pedestal height and confined energy. Above a threshold value in the applied perturbation field (brres) there is a linear increase in normalised ELM frequency (fELM) with brres. Experimentally it has been found that both the lobes produced near the X-point and the mid-plane corrugations also increase linearly with the size of brres. These deformations to the plasma boundary have been replicated by modelling, which shows that they can strongly influence the peeling-ballooning stability boundary and hence lead to an increase in fELM. Mitigation of type I ELMs has also been achieved on ASDEX Upgrade at mid-low collisionalities, using RMPs with n=1 and 2. In a large number of cases an increase of fELM with brres is also observed. However, unlike in MAST, there are some cases where this is not the case. This presentation will compare and contrast the results from the two devices with an aim of increasing our understanding and ability to extrapolate to future devices. This work was part-funded by the RCUK Energy Programme and by Horizon 2020 programme.
        Speaker: Mr Andrew Kirk (UK)
      • 08:30
        ELM Mitigation by Lower Hybrid Waves in EAST 4h
        ELM mitigation has been observed on the Experimental Advanced Superconducting Tokamak (EAST) when lower hybrid waves (LHWs) are applied to H-mode plasmas sustained mainly with ion cyclotron resonant heating (ICRH). This has been demonstrated to be due to the formation of helical current filaments (HCFs) flowing along field lines in the scrape-off layer induced by LHWs. Because of the geometric effect of the LHW antenna, the perturbation fields induced by the HCFs are dominated by the n=1 components, where n is the toroidal mode number. In comparison to previous RMP ELM mitigation experiments, using a set of fixed in-vessel coils, ELM mitigation with LHWs on EAST is achieved with a wider range of q95. This is because the HCFs induced by the LHWs flow along the magnetic field lines in the SOL, thus the helicity of the HCFs always closely fits the pitch of the edge field lines whatever the value of the plasma edge safety factor. Splitting of the outer divertor strike point during LHWs has been observed similar to previous observations with RMPs. The change in edge magnetic topology has been qualitatively modelled by including the HCFs in a field line tracing code. The results show a strong modification of the plasma edge topology dependent on the edge safety factor as well as the amplitude of currents flowing in these filaments. This can qualitatively explain the experimental observations of SP splitting.
        Speaker: Prof. Yunfeng Liang (Forschungszentrum Juelich GmbH)
      • 08:30
        Evaluation of Fuelling Requirements and Transient Density Behavior in ITER Scenarios 4h
        ITER operation requires effective fuelling of the core-plasma for conditions in which neutral dynamics through the scrape-off layer is expected to affect significantly the efficiency of gas penetration. In order to assess fuelling requirements in transients as well as in stationary phases, integrated core-edge plasma modelling has been carried out for plasma conditions expected in the reference 15 MA Q = 10 scenario with emphasis on H-mode operation. The simulations follow the build-up of the H-mode scenario all the way through the L-mode phase, L-H transition, initial ELM free H-mode and the stationary ELMy H-mode phase with controlled ELMs. The JINTRAC suite of codes has been used for this analysis. As a first step the edge plasma has been modelled separately for a range of conditions with the EDGE2D/EIRENE code (included in the JINTRAC suite) and previous results obtained with SOLPS have been confirmed. Full plasma simulations have been carried out with JINTRAC in integrated mode with both the Bohm gyro Bohm core-transport model and GLF23 including two impurities species (Be and Ne). Simulations show that following the H-L transition for the 15 MA DT plasma fuelled by gas-fuelling only the plasma density increases initially on a very fast timescale. However, after this initial phase the pedestal density starts to evolve at a much slower rate due to the low value of particle transport diffusion in the pedestal (~ neoclassical) and decreasing edge transparency to neutrals and finally saturates at values of <ne>~(5-6)x1019 m-3, even for the largest gas-fuelling rates achievable in ITER. Following the L-H transition the pedestal temperature raise rapidly until the edge stability limit is reached and ELMs are triggered. In correspondence of each ELM the density rise drops down significantly. This evolution is similar for DT H-mode plasmas in ITER over a large range of conditions suggesting that, within the uncertainties of the particle transport model, <ne>~(5-6)x1019 m-3 is the highest plasma density achievable in H-mode with gas fuelling alone in ITER; thus allowing high <ne>/nGW operation by gas-fuelling only for the lowest plasma currents (5–7.5 MA). The achievement of high density H-modes for plasma currents above those requires pellet injection. This work was funded jointly by the RCUK and ITER Task Agreement C19TD51FE.
        Speaker: Dr Michele Romanelli (CCFE)
        summary
      • 08:30
        Examination of the Entry to Burn and Burn Control for the ITER 15 MA Baseline and Other Scenarios 4h
        ITER will reach the burning regime Pfusion/Pinput of ~10 by operating in an ELMy H-mode. Control of the flattop burn phase is a critical demonstration for ITER, showing the simultaneous regulation of the plasma core fuel density, fusion power gain, and consistent divertor operation, under several constraints and perturbations. The ITPA-IOS group is doing time dependent integrated simulations (TSC/TRANSP, Corsica, CRONOS, ASTRA/ZIMPUR, RAPTOR) of the burn regime in ITER to understand impacts of physics uncertainties and to develop and test control strategies for the device. Entry to burn simulations are performed to examine the dependences on injected power, rate of rise of the density, argon impurity timing and amount, and feedback control. At SOF densities of n20(0) ~ 0.45 showed marginal access for 43 MW of injected power, while 73 MW was capable of entering and sustaining an H-mode regardless of the density rise trajectory. Simulations examining early, medium, and late Ar injection showed that with 73 MW of input power the earlier injection did not hinder or significantly affect the entry to H-mode, while at 43 MW the timing and amount of Ar strongly affected the access. Simulations examined the impact of the multi-regime H-mode by considering type I ELMy H-mode for Pnet/Pthr > 1.3 with H98 = 1, type III ELMy H-mode with H98 = 0.8 for 0.5 < Pnet/Pthr < 1.3 , and hysteresis that maintains H98 = 0.8 until Pnet/Pthr < 0.5 where H98 drops to 0.5 (L-mode). With 73 MW of input power, and 0.05 and 0.15% argon, fractions the plasma could enter type I H-mode and remain there, for 0.05%, while it dropped back to type III H-mode or lower, with lower energy confinement, at 0.15%. Simulations of a steady state scenario in flattop were conducted with simultaneous multi-variable feedback of the density by fueling, the fusion power by NB injection, power losses to the divertor by Ar impurity seeding, and the loop voltage with lower hybrid current drive power. Perturbations were introduced as an impurity burst at low and high levels, 0.7% and 2.0% to the electron density. The diagonal version of the controller could be used, and simulations of similar perturbations were examined at high Q, all showing good plasma controllability. Work is partially supported by the US Department of Energy under DE-AC02-CH0911466
        Speaker: Dr Florian Koechl (Austria)
      • 08:30
        Experimental and Modelling Results on Wall Conditioning for ITER Operation 4h
        Wall conditioning will be required in ITER to control fuel - and impurity recycling and to improve plasma performance and reproducibility. In the nuclear phase, wall conditioning will also contribute to the control of the tritium (T) inventory within the fuelling cycle. This paper reviews experimental and modelling research activities on wall conditioning in preparation of ITER operation. Baking and Glow Discharge Conditioning (GDC), the primary wall conditioning techniques that ITER will use for cleaning, have been in particular studied in JET-ILW, providing results of particular relevance to ITER operation. The use of Be as PFC material lead to significant reduced needs for wall conditioning after the initial plasma restart, following baking at 200°C and D2-GDC, dramatically contrasting with restart and operation in JET-C with CFC-dominated walls. A novel 2D multi-fluid model has been developed and benchmarked against experimental data, with the aim to assess uniformity and wall coverage with the ITER glow discharge system currently re-designed. We present benchmarking results either from a small laboratory chamber or from large toroidal machines and show that H2-GDC in ITER will be fairly homogeneous in terms of electron density and temperature and toroidal distribution of the ion fluxes to the wall, determining the rate of cleaning. The efficiency of isotopic exchange with GDC or Ion Cyclotron Wall Conditioning (ICWC) for Tritium removal has been assessed in various devices, in particular JET-ILW and ASDEX-Upgrade. A 1D model of isotope exchange in Be has been developed. The model includes processes like hydrogen implantation, trapping to the ion-induced trap sites, detrapping to a solute (mobile) state, diffusion in Be and recombination to molecular form at the surface. Calculated hydrogen depth profiles are compared with those obtained on the linear plasma device PISCES-B. Extrapolation to the ITER, from a database on fuel removal efficiency of isotopic exchange experiments with Ion Cyclotron Wall Conditioning on current tokamaks, in particular JET-ILW and ASDEX-Upgrade, indicates that up to 0.4 gT could be removed between pulses, whereas the estimated T-retention lies between 0.14 and 0.5 gT per 400 s long ITER D:T shots.
        Speaker: Mr David DOUAI (CEA, IRFM, Association Euratom-CEA, 13108 St Paul lez Durance, France)
        summary
      • 08:30
        Fast Particle-Driven Ion Cyclotron Emission (ICE) in Tokamak Plasmas and the Case for an ICE Diagnostic in ITER 4h
        Fast particle-driven waves in the ion cyclotron frequency range (ion cyclotron emission or ICE) have provided a valuable diagnostic of confined and escaping fast ions in many tokamaks. This is a passive, non-invasive diagnostic that would be compatible with the high radiation environment of DT plasmas in ITER, and could provide important information on fusion alpha-particles and beam ions in that device. In JET ICE from confined fusion products scaled linearly with fusion reaction rate over six orders of magnitude [1] and provided evidence that alpha-particle confinement was close to classical [2]. In TFTR ICE was observed from super-Alfvenic alpha-particles in the plasma edge [3]. The intensity of beam-driven ICE in DIII-D is more strongly correlated with drops in neutron rate during fishbone excitation than signals from more direct beam ion loss diagnostics [4]. In ASDEX Upgrade ICE is produced by both super-Alfvenic DD fusion products and sub-Alfvenic D beam ions [5]. The magnetoacoustic cyclotron instability (MCI), driven by the resonant interaction of population-inverted energetic ions with fast Alfven waves, provides a credible explanation for ICE. One-dimensional PIC and hybrid simulations have been used to explore the nonlinear stage of the MCI [6,7], thereby providing a more exact comparison with measured ICE spectra and opening the prospect of exploiting ICE more fully as a fast ion diagnostic. For realistic values of fast ion concentration, the nonlinearly-saturated ICE spectrum closely resembles the measured spectrum. The PIC/hybrid approach should soon make it possible to simulate the nonlinear physics of ICE in full toroidal geometry. Emission has been observed at a wide range of poloidal locations, and so there is flexibility in the requirements of an ICE detector. Such a detector could be implemented in ITER by installing a toroidal loop or adding a detection capability to the ICRH antennae. This work was part-funded by the RCUK Energy Programme and by the European Union's Horizon 2020 programme. [1] G.A. Cottrell et al., NF 33 (1993) 1365 [2] K.G. McClements et al. PRL 82 (1999) 2099 [3] S.J. Zweben et al., NF 40 (2000) 91 [4] W.W. Heidbrink et al., PP&CF 53 (2011) 085028 [5] R. D’Inca et al., Proc. 38th EPS Conf. Plasma Phys., P1.053 (2011) [6] L. Carbajal et al., PoP 21 (2014) 012106 [7] J.W.S. Cook et al., PP&CF 55 (2013) 065003.
        Speaker: Dr Ken McClements (CCFE)
        summary
      • 08:30
        Formation and Termination of Runaway Beams in Tokamak Disruptions and Implications for ITER 4h
        Large runaway electron (RE) currents could be formed during the current quench (CQ) phase of ITER disruptions. Although the main interest of studying REs is related to their final deposition on the plasma facing components (PFCs), much less attention has been paid to their termination phase, when the current and the REs are lost. During this phase, conversion of magnetic energy of the runaway plasma into runaway kinetic energy can occur which can increase substantially the energy fluxes deposited by the REs on the PFCs. In this work, an inter-machine comparison for various devices (JET, DIII and FTU) has been performed which, together with simple 0D modeling of the termination phase, has allowed to identify the physical processes determining the magnetic into kinetic energy conversion. It is predicted that, in ITER, for fast RE losses, below 1 ms, it is essentially the plateau runaway kinetic energy that will be deposited on the PFCs. For long enough RE losses, the avalanche generation of runaways will play an important role, increasing the energy deposited by the REs onto the PFCs, and energies up to 300 MJ for a plateau RE current of 10 MA are predicted. With the aim of improving our understanding of the physics underlying the runaway heat loads onto the PFCs in ITER disruptions, an integrated analysis in which the results of the modeling of the disruption CQ and runaway formation provide the basic inputs for the termination phase of the disruption, is carried out for selected ITER scenarios. This is done by means of simplified models, but retaining the essential physical processes, including the effect of the main runaway generation mechanisms expected in ITER, effects associated with current profile shape during the formation and termination of the runaway current, as well as, in the case of high impurity content, corrections to the runaway dynamics to account for the collisions of the REs with the partially stripped impurity ions. The ultimate goal is to provide a guidance for the most severe foreseen ITER scenarios as well as for the most suitable schemes for the minimization of the effects of runaway impact on the ITER first wall. This work was supported by EURATOM and carried out within the framework of the European Fusion Development Agreement. 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)
        summary
      • 08:30
        From Edge Non-Stiffness to Improved IN-Mode: a New Perspective on Global Tokamak Radial Transport 4h
        Dedicated experiments have been performed on the TCV tokamak to compare the inverse scalelengths of the main plasma region with the one observed in the edge region in standard L-mode plasmas. The main plasma region is known to be characterized by a central region with relatively flat profiles, influenced by the sawtooth activity, and a stiff region where the inverse scalelength is relatively independent on the heat flux. TCV has demonstrated recently that the edge region, inside the last closed flux surface, is not stiff and is key to understanding global confinement properties [1]. It is shown that the inverse scalelength in this region increases with increasing I_p, increasing P_ECH, increasing density and with a change of the plasma triangularity from positive to negative. The role of this non-stiff edge region is also key to understanding the saturation of the ohmic confinement at high density [1]. In these experiments, the ion transport is seen to be essentially neoclassical and the dependence of T_i profiles with I_p will be discussed as well. A new improved L-mode, called the IN-mode, has been obtained on TCV with global confinement time scaling near H-mode values, H_98y2=0.9. This mode will be discussed in detail and compared with the edge non-stiffness discussed above. On TCV, the edge T_e does not show a steep gradient, but the edge density is maintained high, hence the name IN-mode. This high edge density is favourable for keeping high T_i values and good global confinement. The IN-mode has been obtained over a wide range of q_95 and density values, thanks to either a short transition into H-mode or a high gas puffing rate applied directly after break-down and sustained during the I_p ramp-up. Indications are that low l_i are sustained in this way. Core and edge transport properties of these L-mode plasmas are studied in detail with ASTRA simulations [2] and help to better characterize the non-stiff edge properties. The evolution of the profiles up to the L- to H-mode transition is analyzed as well. In particular the role of the edge bootstrap current on the edge q profile is analyzed, with the bootstrap current building up thanks to the edge non-stiff region having steep gradients. [1] O. Sauter et al, accepted in PoP. [2] G.V. Pereverzev and P.N. Yushmanov, IPP Report, 2002.
        Speaker: Dr Antoine Merle (Ecole Polytechnique Fédérale de Lausanne (EPFL), Centre de Recherches en Physique des Plasmas (CRPP), CH-1015 Lausanne, Switzerland)
        summary
      • 08:30
        Fusion Alpha Loss in ITER with Local Marginal Stability to Alfven Eigenmodes 4h
        A simple 1D radial transport code predicting the fusion alpha density profiles in an ITER burning plasma unstable to Alfvén eigenmodes (AEs) is illustrated. With the local AE thresholds exceeded only in the mid-core in the baseline case, we find only moderate mid-core flattening of the alpha density (and heating) profile and negligible alpha heating loss. Neglecting “ripple loss,” only microturbulent low energy alpha transport remains at the edge for this baseline case, so escaping alpha particles are best described as very hot helium. Edge energy loss in the alpha channel is then 1000 times smaller than in the thermal channels. This work extends earlier work by Angioni et al. [1] treating the fusion alpha transport from high-n micro-turbulence to include marginal stability (or “stiff”) transport from alpha driven low-n Alfvén eigenmodes. The local alpha density gradient AE thresholds are provided by physically realistic linear gyrokinetic GYRO code calculations. The transported alpha density profiles are compared to the (no transport) classical (or collisional) slowing-down alpha profiles. The baseline thermal plasma (and hence fusion alpha source) profiles are taken from the Kinsey et al. [2] ITER performance projection. The distinction between the alpha particle and the much smaller alpha energy microturbulent transport loss is emphasized. Any high-energy alpha losses would both reduce needed plasma heating and increase the risk of material damage to plasma-facing surfaces. We predict no such losses in this baseline case. This work was supported by the US Department of Energy under DE-FG02-95ER54309 and DE-FC02-08ER54977. [1] C. Angioni, A.G. Peters, G.V. Pereverzev, A. Botino, J. Candy, R. Dux, E. Fable, T. Hein, and R.E. Waltz, Nucl. Fusion 49, 055013 (2009). [2] J.E. Kinsey, G.M. Staebler, R.E. Waltz, and J. Candy, Nucl. Fusion 51, 083001 (2011)
        Speaker: Dr Eric M. Bass (University of California San Diego)
        summary
      • 08:30
        High Density Regime in Ohmic TCV Discharges with Positive and Negative Triangularity 4h
        Studies of high density plasmas approaching the Greenwald limit are timely and necessary in view of future reactor operation, both to gain a phenomenological understanding of their behavior and to validate and improve the theoretical treatment of their stability, confinement and control, as well as the physics underlying the density limit process itself. One of the key features of the TCV tokamak is the possibility to change plasma shaping in a wide range of plasma elongation and triangularity. In previous TCV experiments it was shown that in both ohmic and ECRH L-mode discharges the plasma confinement time increases with plasma elongation and decreases with triangularity. In the present work the operational space of TCV has been extended to densities close to the Greenwald limit, and the evolution of the Ohmic confinement with the density increase has been investigated in limited discharges with positive and negative triangularity. We find that the limit density value has the same increasing dependence on plasma current in both cases; however, the dependence is weaker than predicted by the Greenwald formula. The limit density approached the Greenwald limit only in the delta_95>0 case at low current. In discharges with delta_95<0 the value of the limit density appeared to be lower in the whole density range explored. The increase in plasma density in both cases was found to be accompanied by a change in sawtooth behavior, namely an increase in the sawtooth period, a modification of the relaxation dynamics, and a reduction in regularity (variable period and amplitude); this was then followed by the disappearance of sawtooth oscillations altogether. Energy and particle confinement are also affected by the density increase. In discharges with delta_95>0 a transition from linear to saturated ohmic confinement is observed at a line-averaged density (4-4.5)*10^19 m^-3, and the start of the sawtooth-free phase is followed by a decrease in the energy confinement time. In the delta_95<0 case the confinement behavior was found to be strongly dependent on plasma current: at high I_p (q_95~3) the density dependence of tau_E is similar to the delta_95>0 case, whereas at low current (q_95~5.5) a pronounced confinement degradation with density is observed. The possible role of MHD activity and the effect of the gas-puffing rate will be discussed.
        Speaker: Dr Natalia Kirneva (NRC "Kurchatov Institute")
        summary
      • 08:30
        High Power ICRF Systems and Heating Experiments in EAST 4h
        The ICRF system of 12.0 MW has been developed for EAST. To support the long pulse operation over 1000s, the ICRF heating system is upgraded with active cooling, especially for ICRF antenna systems. The ICRF system of 6.0MW has been operating in the 2012 experimental campaign. A new 6.0 MW system has been successfully commissioned at full power on water dummy load. In the upcoming campaign,the ICRF system is capable of delivering more than 10 MW of rf power to the plasma. The relevant experimental results from the upcoming campaign will be given. Heating power modulation experiments using ion cyclotron resonance heating (ICRH) in the Hydrogen minority scheme have been performed in the 2012 campaign of the EAST. The power deposition profile in the ion cyclotron range of frequencies (ICRF) has been investigated experimentally. The D (H) minority heating scheme provides a dominant localized ion heating. In this scheme, electron heating occurs only through collisions with the minority ion tail. The results shows that the peak of the experimental power deposition profile is always occurred around to ω=ωcH. The global energy confinement time was calculated in the ICRF modulation experiment, and compared with the scaling law, ITER-89. It was found that the calculated results coincided with ITER scaling law. The global energy confinement time decreases by a factor of 2 approximately from that in Ohmic plasmas with the ICRF heating power increases up to 1.6 MW. The heating efficiency was somewhat lower than expected. The effect on the sawtooth period was demonstrated in the experiments analyzed in this paper.
        Speaker: Dr Xinjun Zhang (Institute of Plasma Physics Chinese Academy of Sciences (ASIPP))
      • 08:30
        Impact of W on Scenario Simulations for ITER 4h
        AUG and JET, the largest present devices with high-Z PFC components, have identified requirements for stable H-mode operation, i.e. to keep heavy impurity concentrations sufficiently low, to avoid central accumulation, radiative collapses and disruptions. Limitations in the operational space which can be accessed in H-mode have been identified, e.g. (i) the need of operating at sufficiently high levels of gas puff, impeding access to low density regimes at low gas puff levels; (ii) central electron heating and/or frequent sawteeth may be needed to avoid W core accumulation. This paper starts with a short review of experimental results on erosion sources, edge (pedestal) transport, and core W transport. Then implications for ITER are discussed, concentrating on the effect of core W accumulation on the discharge evolution. In different ways the critical W concentration in ITER was assessed i.e. the maximal tolerable level without significantly perturbing the evolution of li, or the q and Te profiles. First, impurity transport (both neoclassical and anomalous) was modelled with the ZIMPUR code, in combination with ASTRA for the description of the bulk plasma parameters evolution with a scaling based transport model. The calculated critical W concentration is ~7 10-5 for the inductive scenario, and a factor 2-3 lower for the hybrid and steady state scenarios. Second, the current ramp-up phase in JET, AUG and ITER was modelled with the CRONOS suite of codes for different W concentrations. For ITER the expected plasma parameters for the baseline ITER ramp-up were used; for JET and AUG the experimental data (Te, ne, Zeff) were taken. The Bohm-gyroBohm model for thermal transport was used, and nW/ne profiles were assumed either flat or of the same (peaked) shape as measured in JET. The effects of flat and peaded W profiles are very different. The modelling results are in excellent agreement with experimental findings. As maximum tolerable W concentrations have now been calculated for different ITER scenarios, future work can concentrate on further quantify these limitations, using current understanding of neo-classical and anomalous W transport. This work was supported by EURATOM and carried out within the framework of the European Fusion Development Agreement. The views and opinions expressed herein do not necessarily reflect those of the European Commission.
        Speaker: Dr Gerrit Hogeweij (FOM)
        summary
      • 08:30
        Improved beta (local beta > 1) and density in electron cyclotron resonance heating on the RT-1 magnetosphere plasma 4h
        This study reports the recent progress in improved plasma parameters of the RT-1 device. Increased input power and the optimized polarization of electron cyclotron resonance heating (ECRH) with an 8.2 GHz klystron produced a significant increase in electron beta, which is evaluated by an equilibrium analysis of Grad-Shafranov equation. The peak value of the local electron beta e was found to exceed 1. In the high beta and high-density regime, the density limit was observed for H, D, and He plasmas. The line average density was close to the cutoff density for 8.2 GHz ECRH. A density limit exists even at the low beta region. This result indicates the density limit is caused by the cutoff density rather than the beta limit. From the analysis of interferometer data, the uphill diffusion produces a peaked density profile beyond the cutoff density.
        Speaker: Dr Masaki Nishiura (The University of Tokyo)
        summary
      • 08:30
        Influence of a Tungsten Divertor on the Performance of ITER H-Mode Plasmas 4h
        The effect of a tungsten divertor on the performance of H-mode plasmas in ITER has been investigated by combining scrape-off layer transport calculations performed with SOLPS with core transport simulations using the ASTRA code, which was coupled to the impurity transport code STRAHL. The penetration of W into the central plasma mainly depends on two mechanisms: prompt re-deposition of W to the target and the radial transport in the edge transport barrier (ETB) during and between ELMs. Within an ELM, the transport of W is strongly enhanced and concomitantly, the physical sputtering of W at the divertor target strongly increases. Monte Carlo simulations of prompt W re-deposition, which included the effects of multiple W ionisation and electric field force on the ions in the magnetic pre-sheath, were carried out for ITER controlled ELM conditions. It was found that W re-deposition causes a significant, factor 10000, reduction in the net W erosion. The avalanche effect, where W self-sputtering could lead to a runaway process of increasing W sputtering, can be ruled out. Thus, the SOLPS simulations found that controlled ELMs conditions present very little danger of plasma contamination with sputtered W. Between ELMs, it can be assumed that the W-transport in the ETB is due to neoclassical transport which was studied for a large range of pedestal profiles. The radial convection velocity of W was found to be outward directed for the major part of the tested profiles. This is due to a combination of high pedestal temperatures and high separatrix densities making the outward directed temperature screening term to be the predominant contribution of the convection. The high densities at the separatrix are needed to control the power exhaust and the sputtering in the divertor and the high pedestal temperatures are expected to be achieved to meet the fusion performance objectives. Combined ASTRA+STRAHL transport simulations in presence of ELMs of varying frequencies have been carried out. Both neoclassical and ad-hoc anomalous transport models have been included to simulate the evolution of the W profile in the pedestal region. When using the sources as calculated with SOLPS, there was no influence on the total plasma radiation. Even for the most pessimistic case of no redeposition, ELM frequencies in the range 10-30 Hz lead still to tolerable W concentrations.
        Speaker: Dr Ralph Dux (Max-Planck-Institut für Plasmaphysik, Garching, Germany)
        summary
      • 08:30
        Influence of Magnetic Perturbations on Particle Pump-out in Magnetic Fusion Devices 4h
        Control of Type-I Edge Localized Modes (ELMs) is an important task for next step fusion devices like ITER. Following the results obtained on DIII-D resonant magnetic perturbations (RMPs) became a very popular tool to control plasma exhaust in tokamaks like ASDEX-Upgrade, MAST, KSTAR and probably ITER. RMPs produce a stochastic boundary which reduces the pressure gradients in the pedestal region allowing the suppression or mitigation of ELMs while keeping the outward transport enhanced. The density reduction in the pedestal area is a large contributor to reducing the pressure gradient below the peeling-ballooning stability limit. There were already several attempts both experimentally and theoretically to understand the interaction of magnetic perturbation with L-mode and H-mode plasmas, this work is aimed to summarize recent experimental results from ASDEX-Upgrade, DIII-D, KSTAR, MAST, NSTX, LHD and TEXTOR with resonant and non-resonant magnetic perturbations. In L-mode plasmas the influence on the edge and core plasma is typically more pronounced, when comparing to H-mode plasmas, what is attributed to much better coupling of non-axisymmetric perturbation with the plasma equilibrium. Recent MAST results show that the largest degree of pump-out coincides with the best alignment of the external field to the equilibrium field, which agrees with findings from TEXTOR. During H-mode discharges numerous experiments report changes in transport during phases where type I ELMs are mitigated (JET, DIII-D) or suppressed (ASDEX Upgrade, DIII-D, KSTAR) by external perturbations. As observed in ASDEX Upgrade, DIII-D and LHD the pump-out seems to depend on pedestal collisionality/density. Additionally on ASDEX-Upgrade depending on the spectrum of the magnetic perturbation one gets either reduction (for a non-resonant case) or increase (for a resonant case) of pedestal and central electron densities. This cannot be explained with increased frequency of mitigated ELMs as the particle losses per event decrease with ELM frequency as reported on ASDEX Upgrade, MAST and LHD.
        Speaker: Dr Marcin Jakubowski (Max-Planck-Institut für Plasmaphysik)
        summary
      • 08:30
        Integrated Core-SOL-Divertor Modelling for ITER Including Impurity: Effect of Tungsten on Fusion Performance in H-Mode and Hybrid Scenario 4h
        Different plasma performance (energy confinement, discharge duration) has been generally observed in operationally close JET discharges with carbon (C) and ITER-like wall (ILW). Presence of tungsten (W) in ILW discharges is one of the major changes introduced with the wall replacement which may partly explain the differences observed at JET and have an impact on ITER operation. Effect of W on ITER H-mode and hybrid fusion performance is analysed here via integrated core-SOL-divertor modelling performed with ASTRA and JETTO (theory-based core simulations for main species) and COREDIV (impurities and core-SOL-divertor integration) codes used iteratively. Be, He, W and Ne impurities are simulated self-consistently with main plasma species. Core toroidal rotation velocity is also predicted, which is another novelty of this work. The core thermal and particle transport is simulated with the GLF23 transport model which has been extensively validated on a number of H-mode plasmas and hybrid scenarios (HS) in various tokamaks. Here, this model is tested in self-consistent simulations of temperatures, density and toroidal rotation velocity in JET HS. The correlation of the ExB shear amplification factor alpha_E with Mach number has been found in these simulations. The alpha_E and Prandtl number validated on JET are projected to ITER. The COREDIV model has been successfully benchmarked with the JET H-mode discharges and advanced scenarios. Modelling results show that the core-SOL-divertor coupling in ITER plasmas is very strong in presence of W impurity and it affects the operational domains for considered scenarios. The long-pulse H-mode scenario analysed in the density range n=(8.3–11.9)x10^19 m-3 has narrow operational window limited by the tolerable power to divertor (achieved due to optimised neon gas puff and marginally sustained H-mode under given modelling assumptions. The consequences of strong core-SOL-divertor coupling for low density H-mode and hybrid operation where the improvement in fusion performance due to a good core confinement is counteracted by large W production and core radiation are investigated. Sensitivity studies will be presented showing under what conditions the core-SOL-divertor coupling is reduced allowing high H-mode and hybrid performance.
        Speaker: Dr Roman Zagorski (Institute of Plasma Physics and Laser Microfusion)
        summary
      • 08:30
        Integrated Modelling of ITER Disruption Mitigation 4h
        Feasibility of the ITER disruption mitigation system (DMS) to a)mitigate heat loads on the divertor target plates and plasma facing components during the thermal quench (TQ) phase of the disruption; b) reduce electromagnetic forces on the vacuum vessel during current quench; c) avoid or control the runaway electron (RE) generation are studied in the present report. Complex variety of physical phenomena comprising disruption of a tokamak discharge requires integrated modeling approach. The well-validated DINA code [1] is used as an integrating core module for disruption simulator development. Whenever possible the DINA results are verified by ASTRA code [2] simulations. Impurity charge state dynamics, radiation and transport are calculated by the ZIMPUR code [3]. Newly developed gas flow model allows accurate accounting for the technical specifications of MGI system foreseen for ITER DMS. RE generation, evolution and suppression are simulated with use of Monte-Carlo solver for RE kinetic equation integrated with DINA code. Full disruption scenarios from “prediction” of expected disruption, till complete termination of the plasma current are simulated to determine operation domain for the ITER DMS based on MGI. It is shown that optimization of MGI parameters (geometry, gas mix content and quantities) allows to draw consistent scenario of mitigated disruption with use of two-component MGI system. The first one is aimed on the TQ heat load mitigation, while the second one provides safe plasma current termination without excessive forces on the construction and suppression of REs beams if they appeared. [1] Khayrutdinov, R.R. and Lukash, V.E., Journal of Computational Physics, 109, (1993) 193. [2] Pereversev, G.V., Yushmanov, P.N., Preprint IPP 5/98, Garching. Germany (2002). [3] Leonov, V.M., Zhogolev, V.E., Plasma Phys. Control. Fusion, 47 (2005) 903
        Speaker: Mr Sergey Konovalov (NRC "Kurchatov Institute")
      • 08:30
        Investigation of Argon Seeding in Different Divertor Configurations in EAST and Corresponding SOLPS 5.0 Modeling 4h
        Introducing external impurities into plasma provides an effective means to reduce divertor power load for present and future fusion devices [1]. Dedicated argon (Ar) seeding experiments focusing on the effects of the plasma configuration and seeding position have been carried out in EAST, with the corresponding simulations using SOLPS 5.0 code package being also ongoing. The double null (DN) divertor configuration is found to have a cooler divertor plasma before Ar seeding comparing to lower single null (LSN), as expected [2]. When Ar is seeded, the parallel heat flux to the lower divertor, measured by target Langmuir probes, exhibits a slightly more dramatic decrease in the case of DN configuration than in LSN configuration. At the outer midplane, the measurements by reciprocating probe reveal that the electron temperature and density are higher in DN configuration than in LSN configuration after Ar seeding. This indicates a greater gradient of temperature along the field line in DN case, which is beneficial for reducing divertor power load. In addition, the plasma stored energy Wdia increase with Ar seeding in DN case, meaning the confinement is improved. Comparisons have also been made between Ar puffing into the divertor volume in LSN configuration and into main chamber in USN configuration. The radiation in the latter case increases by 42.6% together with a significant increase in Zeff, while in both cases the plasma stored energy almost stays unchanged. Though electron temperature and density at midplane is higher, the parallel heat flux at divertor plate is lower in USN case. However, after Ar seeding no decrease in parallel heat flux to upper outer divertor plate is seen in this case. At inner plate only the peak value of parallel heat flux reduces by 13%. It is evident that Ar can readily penetrate into the core when it is puffed into the main chamber volume, but does not affect the heat flux to the divertor very much in this case. we will include cases of different Ar puff locations as well as cases of Ar seeding in different plasma configurations in the SOLPS simulation to compare with present and further experiment results. [1] ITER Physics Basis 1999 Nucl. Fusion 39 2208 [2] S.C. Liu et al. Phys. Plasmas 21, 022509 (2014)
        Speaker: Ms Lingyan Xiang (Institute of Plasma Physics, Chinese Academy of Science)
      • 08:30
        Investigation of LHW-Plasma Coupling and Current Drive Related to H-Mode Experiments in EAST 4h
        LHW-plasma coupling and high density are two important issues in achieving LHCD H-mode plasma in EAST. Firstly, effects of LHW on the density at the grill mouth are investigated by a Langmuir probe installed in the top of the LHW antenna. Results show that the measured density with anti-clockwise Bt is lower than those with clockwise Bt, suggesting the asymmetric density behaviour in SOL. Simulation with a diffusive-convective model suggest that such asymmetry is mainly due to E×B drift. Secondly, high density experiments with LHCD are further analyzed by means of simulation, showing that parametric instability (PI), collision absorption (CA) in edge region, and density fluctuation could be responsible for the low current drive (CD) efficiency at high density. (i) Frequencies and growth rates of coupled modes could be identified near the antenna and the LCFS. Modelling results show that the line broadening of the operating LH frequency and the downshifted sidebands, observed during the experiments, could be produced by PI effects driven by ion-sound and ion-cyclotron (IC) quasimodes, respectively, near the plasma edge in the low field side. The growth rates are larger in the case of poor lithiation, consistently with the observed reduced CD efficiency. Simulations also show that the growth rate peak of the IC sideband occurs near the LCFS, and, in case of poor lithiation, has a smaller frequency shift from the pump, in agreement with the RF probe spectra. (ii) The fraction of LH power calculated with GENRAY code indicates that more LHW power is absorbed in SOL by collision in the case of poor lithiation, whether at low density or high density, making some contribution to the low CD efficiency in this case. (iii) Modification of spectrum due to density fluctuation makes the power deposition and driven current profile predicted by C3PO/LUKE ray-tracing and Fokker-Planck move inward, but the total value of driven current decreases (~30%). This may partly explain the small CD efficiency in the case of SMBI. In addition, CD efficiency with considering bootstrap current and hot electrical conductivity has been investigated in EAST H mode. Results show the efficiency maximum locates at n_e ~ 2.2×10^19m^-3, above which the efficiency drops significantly with density increase, nearly consistent with the above HXR emission. Further study will be continued.
        Speaker: Dr Bojiang Ding (Institute of Plasma Physics, Chinese Academy of Sciences)
      • 08:30
        ITER Energetic Particle Confinement in the Presence of ELM Control Coils and European TBMs 4h
        The new physics introduced by ITER operation, of which there is very little prior experience, is related to the very energetic (3.5 MeV) alpha particles produced in large quantities in fusion reactions. These particles not only constitute a massive energy source inside the plasma, but also present a potential hazard to the material structures that provide the containment of the burning plasma. In addition, the negative neutral beam injection (NBI) produces 1 MeV deuterons and the application of ICRH produces minority ions in multi-MeV range, both of which have to be well confined to ensure successful operation of ITER. Since energetic ions are very sensitive to the details of the magnetic field, in this contribution the field was calculated in unprecedented detail, including all the known magnetic perturbations such as ferritic inserts, TBMs and ELM Control Coils (ECC). The FEM solver COMSOL was used to first calculate the magnetization of the ferromagnetic components due to plasma current and currents flowing in the field coils. The perturbation field due to the magnetization was then calculated and added to the unperturbed field integrated from the coils using the Biot-Savart law. The cases reported here correspond to the 15 MA standard H-mode operation and the 9 MA advanced scenario in ITER. Both thermonuclear fusion alphas and NBI ions from ITER heating beams are addressed. Both species are simulated using the Monte Carlo orbit-following code ASCOT in the full 3D magnetic configuration given by the COMSOL calculations. The first wall also has full 3D features. The ferritic components are found not to jeopardize the integrity of the first wall, but the application of ECC needs further attention, in particular for the potential resonance amplification.
        Speaker: Dr Taina Kurki-Suonio (Aalto University)
        summary
      • 08:30
        Kinetic Modelling of Runaway Electrons and their Mitigation in ITER 4h
        Runaway electrons (RE) can be generated during plasma disruptions in ITER. A large portion of the toroidal current can then be carried by RE, so that a substantial fraction of the magnetic energy would be associated with the RE current. The uncontrolled loss of such REs in ITER needs to be avoided to minimize the detrimental effects from disruptions. This calls for an adequate means to control or suppress a RE beams in ITER. Massive Gas Injection (MGI) has been successfully employed to mitigate disruptions in present tokamaks, and is therefore considered to be an essential component of the envisioned ITER Disruption Mitigation System (DMS). Recent experiments has shown, that MGI of high-Z impurities (Ar,Ne,...) into a RE beam leads to a fast current decay. In the present work, the ITER MGI target parameters are assessed taking into account the interaction of REs with high-Z nuclei. The 2D Fokker-Planck equation is solved, which describes the kinetics of REs, coupled to an electric field evolution model. Enhanced scattering of REs on high-Z impurities was shown to modify the RE distribution function, resulting in the broadening of the velocity distribution as well as a decrease in the average kinetic energy of REs. As an alternative mechanism of enhanced scattering or stopping REs, the possible excitations of kinetic instabilities (whistlers or magnetized waves) are considered. It was shown, that an instability window for whistlers exists within possible ITER disruptions parameters. Simulations of RE suppression by MGI revealed that both the RE current and kinetic energy can be successfully dissipated by the injection of a moderate amount of a noble gas, well below ITER technical limitations.
        Speaker: Dr Pavel Aleynikov (ITER Organization)
        summary
      • 08:30
        Magnetic System of Multipole Trap--Galatea on the Basis of Levitating Quadrupole 4h
        The possibility of the creation of the magnetic system of multipole trap-Galatea on the basis of the levitating quadrupole from the superconducting coils-rings is considered. Based upon the superconductor property to conserve the trapped magnetic flux the analytical dependence of the potential energy of the proposed configurations from the coordinates of the levitating coils and the deflection angle of their axis has been obtained. The calculations in Mathcad system have shown that under the definite values of the physical parameters (the trapped magnetic flux, dimensions and masses of coils, etc.) this dependence has local minimums, which correspond to the stable equilibrium states of levitating coils. For carrying out experiments with levitation several multiturn short-circuited coils-rings have been made from the high-temperature superconducting (HTSC) wire of the SCS4050-i-AP 2G HTS type. HTSC rings have been made also from the preliminary synthesized powder of HTSC phase YBa2Cu3Oy with the help of melt textured growth (MTG) method. Using the experimental data on the trapped magnetic fluxes for HTSC rings, their dimensions and masses, and also the parameters of the ordinary coil with the current, with the help of calculations of the pointed out dependence for the potential energy the search of the equilibrium states for the different cases has been carried out. Under the magnetic fluxes of the same polarity in coils, the stable levitating states of: 1) single HTSC ring both in the field of other HTSC ring and in the filed of the ordinary coil with the current; 2) two HTSC rings in the filed of the ordinary coil with the constant current are observed experimentally in positions corresponding to calculated values.
        Speaker: Dr Marina Kozintseva (Moscow State Technical University of Radioengineering, Electronics and Automation, Moscow, Russia)
      • 08:30
        Micro- and Macro-Instability, and Large Density and Beta in Improved Confinement MST RFP Plasmas 4h
        In MST plasmas where inductive current profile control routinely produces tokamak-like confinement with enhanced density and temperature gradients, a far-forward-collective-scattering diagnostic reveals a broadband reduction of density fluctuations, with as much as a 100-fold drop in amplitude. This drop is precipitated largely by the reduction of current-gradient-driven tearing modes. However, in the region where thermal gradients are largest, there is a localized peak in the density fluctuation power spectrum. The source of these fluctuations is not established, but gyrokinetic modeling with the GENE code suggests the trapped electron mode, driven by the local density gradient. These plasmas, whose density is well below the Greenwald limit, routinely exhibit a large total beta (average plasma pressure normalized to the total edge field pressure) of about 15 percent, but this beta is limited only by the finite Ohmic heating power. With the goal of trying to establish the limits on density and beta in these plasmas, frozen deuterium pellets were injected to increase the density and the Ohmic heating power. This has resulted in an RFP-record density of 1.6 times the Greenwald limit, but the upper limit on the density in these plasmas still has yet to be established. Over a broad range of density, from well below to well above the Greenwald limit, the total beta is routinely enhanced. And now, with the addition of NBI, the total beta has reached an RFP record of about 28 percent (with toroidal beta ~ 115 percent), and it appears that this represents a soft limit on beta. Magnetic fluctuations increase as the density increases, leading to increased energy transport, and these discharges do not exhibit premature terminations. According to MHD modeling with the NIMROD code, both pressure-driven tearing and interchange modes can be linearly unstable in these plasmas, unlike the discharges at lower beta. These instabilities can contribute to the enhancement of magnetic fluctuations. This soft beta limit phenomenology is similar to that observed in some stellarators.
        Speaker: Dr Brett Chapman (UW-Madison)
      • 08:30
        Modelling of Melt Damage of Tungsten Armour under Multiple Transients Expected in ITER and Validations against JET-ILW Experiments 4h
        The ITER Organization has recently decided to install a full-tungsten (W) divertor from the start of operations. One of the key issues with such a strategy is the possibility of W melting and melt splashing during transients, which can lead to modifications of surface topology and which may lead to higher disruption frequency or compromise subsequent plasma operation. Although every effort will be made to avoid leading edges, ITER plasma stored energies are sufficient that transients can drive shallow melting on the top surfaces of components. A new experiment has now been performed on JET-ILW in the ITER-Like Wall (ILW) environment, in which a deliberately misaligned W element (lamella) in the outer divertor has been used to perform controlled ELM transient melting experiments for the first time in a tokamak. This paper reports on the application of the 3D MEMOS code to modeling of these experiments. Input heat loads are obtained from experimental data, notably high resolution IR camera thermography. Importantly, the code indicates that that shielding by the evaporated tungsten prevents bulk melting between ELMs. Encouragingly, the simulations are also able to quantitatively reproduce the dimensions of the damaged area observed by high resolution photography after the first pulse in which melting was achieved. MEMOS simulations on the consequences of multiple mitigated major disruptions (MD), mitigated vertical displacement events (VDE) and major disruptions expected in ITER on damage of tungsten castellated armour have been performed for several scenarios of impact conditions specified by IO. This work, supported by the European Communities under the EFDA Task Agreement between EURATOM and Karlsruhe Institute of Technology (KIT) and contract between IO and KIT, was carried out within the framework of the European Fusion Development Agreement. The views and opinions expressed herein do not necessarily reflect those of the European Commission or of the ITER Organization.
        Speaker: Dr Boris Bazylev (Karlsruhe Institute of Technology, Germany)
      • 08:30
        Modelling of Transitions Between L- and H-Mode Including W Behaviour in ITER Scenarios 4h
        The dynamics of the access to and exit from high QDT regimes in the H-mode confinement regime in ITER is expected to be qualitatively different to present experiments: neutral fuelling is much less effective, Psep/PL-H < 2.0 even in stationary QDT ~ 10 burning conditions, the density evolution determines not only PL-H but also Palpha which in turn affects dWth/dt after a transition, and plasma position control may be challenging in case of an unexpected back transition to L-mode. In addition, the presence of W may impose additional operational constraints due to possible core accumulation and increased radiation during transients (possibility of a sudden return to L-mode confinement, plasma-wall contact and/or a disruption). To determine under which conditions the transition to stationary high QDT H-mode regime and its safe termination can be achieved, how the plasma evolution to/from H-mode can be optimised, and to assess the problem of possible core W accumulation, modelling studies have been carried out with the JINTRAC suite of codes, simulating the core and core+SOL plasma evolution for the entire period of density evolution following transitions to/from H-mode in the ITER 15 MA/5.3 T and 7.5 MA/2.65 T scenarios. Simulation scans for the L-H transition have been performed with varying target waveforms for the density evolution, applying a feedback on pellet fuelling. Depending on boundary and operational conditions, limits for the density ramp rate and/or a delay time before the application of increased fuelling could be established. Below these limits, the plasma remains in dithering conditions with Psep~PL-H for a long while before it enters a good quality H-mode regime at Psep>PL-H, leading to increased flux consumption and a significantly reduced burn duration. In extreme cases, the plasma never reaches high performance H-mode and returns back to L-mode. The back transition to L-mode has also been assessed. The fast reduction in core energy would cause Psep to remain close to PL-H. The plasma would then not immediately reach L-mode but stay in H-mode for a while, followed by a dithering phase before the ETB completely disappears. The energy loss could become accelerated though by an immediate transition to dithering mode e.g. after a strong MHD event. Subsequent W accumulation could then lead to an immediate transition to L-mode and a disruption.
        Speaker: Dr Florian Koechl (Vienna University of Technology, Institute of Atomic and Subatomic Physics)
        Poster
        summary
      • 08:30
        Modelling Toroidal Rotation Damping in ITER Due to External 3D Fields 4h
        Three-dimensional external magnetic field perturbations, can either be intentionally applied such as in the experiments of mitigating edge localised modes (ELM) using resonant magnetic perturbations (RMP), or be un-intentionally generated such as the intrinsic error fields (EF). One crucial consequence of applying these (nearly) static 3D fields is the plasma flow damping. In this work, we model the toroidal rotation damping in ITER plasmas using the recently developed MARS-Q code. This code solves the n = 0 toroidal momentum balance equation together with the single fluid MHD equations describing the plasma response to external 3D fields. The code includes the momentum diffusion and pinch terms, as well as various momentum sink/source terms (torques) that contribute to the momentum balance and consequently determine the time evolution of the flow profile and its amplitude. We include the resonant electromagnetic JxB torque, the neoclassical toroidal viscous (NTV) torque, the (v,grad)v type of Reynolds stress torque associated with the plasma inertia, the torque source due to neutral beam injection (NBI), the torque due to energetic particle losses in the 3D fields, as well as the torque source associated with the intrinsic rotation. Both the RMP field and the intrinsic error field are considered, with somewhat different toroidal spectra: n = 3 and 4 for the former and n = 1 and 2 for the latter. For an ITER 15MA plasma with the pedestal temperature of 4.5kA, that we have modeled, preliminary results from the MARS-Q runs show a minor damping of the plasma flow with 30kAt RMP coil current in the n=4 configuration. On the other hand, the plasma rotation at the edge is fully damped with 45kAt current after about 40ms of applying the n=4 RMP field. A further increase of the coil current to 60kAt leads to a quicker damping of the edge flow (~20ms). Similar simulations with the n = 3 coil configuration at 45kAt current also show edge rotation braking. The modelling suggests that the primary damping comes from the NTV torque.
        Speaker: Dr Yueqiang Liu (CCFE Culham Science Centre)
        summary
      • 08:30
        Multi-Diagnostic Study of Core Turbulence and Geodesic Acoustic Modes in the TCV Tokamak 4h
        TCV is equipped with a suite of diagnostics capable of making fluctuation measurements of several plasma parameters. The suite has been used in a large variety of TCV discharges. Emphasis has been placed on the study of turbulence as a function of plasma shape and in particular of edge triangularity. Correlation ECE (CECE) measurements have shown that the relative electron temperature fluctuation amplitude decreases as edge triangularity goes from positive to negative (+/- 0.34). At the same time tangential phase contrast imaging (TPCI) measurements show an analogous reduction in the density fluctuation component deep into the plasma core (rho ~ 0.3). Local, non-linear, flux tube, gyrokinetic simulations have reproduced the fluctuation reduction in the plasma edge but not in the core (rho < 0.7). The geodesic acoustic mode (GAM) has been identified in TCV discharges through its toroidal symmetry and the linear scaling of its frequency with sound speed. It has been simultaneously detected in radiative temperature, electron density, magnetic field and plasma flow velocity measurements, appearing as a coherent mode in the 20-30kHz range close to the plasma edge. The multiple diagnostic identification of the GAM has allowed its radial location and poloidal distribution and its propagation direction to be determined. The poloidal mode number of its magnetic component is predominantly 2 as predicted by theory. In some cases the GAM no longer appears as a single coherent mode but as a continuum with radially varying frequency. The transition from coherent to continuum nature has been observed in a discharge during the course of a current ramp. Global simulations with the PIC gyrokinetic code ORB5 have been performed to study the GAM characteristics. Results are in good, semi-quantitative agreement with experimental findings. Synthetic diagnostics are being developed to allow comparison of numerical simulations with experimental results. Prototype CECE and TPCI synthetic diagnostics have been developed as post processing modules for use with the GENE code and are planned for the ORB5 code. First comparisons with experimental data will be presented.
        Speaker: Dr Laurie Porte (CRPP-EPFL)
        Poster
        summary
      • 08:30
        Off-Axis Current Generation by Helicons and LH Waves in Core of Modern Tokamaks and Reactors FNSF-AT, ITER, DEMO and by Alfven Waves in Pedestal Plasmas. Scenarios, Modeling and Antennae 4h
        The Innovative concept and 3D full wave code modeling Off-axis current drive by RF waves in large scale tokamaks, reactors FNSF-AT, ITER and DEMO for steady state operation with high efficiency is proposed. The scheme uses the helicons radiation (fast magnetosonic waves at high (20-40) IC frequency harmonics) at frequencies of 500-1000 MHz, propagating in the outer regions of the plasmas with a rotational transform [1]. Modeling with full wave three-dimensional codes PSTELION and STELEC2 showed flexible control of the current profile in DIII-D, T-15 and KSTAR and reactor plasmas of ITER, FNSF-AT and DEMO, using multiple frequencies, the positions of the antennae and toroidal waves slow down. Commercially available klystrons of MW/tube range, CW working, are promising for commercial stationary fusion reactors. The compact antennae of waveguide type in Traveling Wave regime are proposed, and the example of possible RF system for today's tokamaks and proposed for russian FNSF-AT project is given. For spherical tokamaks the Helicons excitation scheme does not provide efficient Off-axis CD profile tayloring flexibility due to strong coupling of helicons with O-mode, also through the boundary conditions in low aspect machines, and intrinsic large amount of trapped electrons, as will be shown by STELION modeling for the NSTX tokamak. Alfven Resonance Heating/CD method for ELMs and EHO control pedestal plasma is proposed. New possibility is to control the plasma current in pedestal being an essential element in peeling/ballooning modes stability. We propose to use the Alfven resonance scheme based on well known shear Alfven wave relation for frequencies well below of ion cyclotron ones: ω = K//VA . The slow waves and KAW are absorbed mainly by the electrons in pedestal area. Respectively, with properly toroidally phased antenna the SLOW/KAW waves highly efficiently (similar to LH) drive non inductive current to be exploited for ELMs and EHOs control in tokamaks. The antenna with ICRF-like poloidal loops or RMP coils arrays excites the near fields which penetrate into plasma. Thus exploration of Alfven Resonance scenario to low frequencies from megaherz to tenth kHz evidences the importance of KAW excitation in tokamak pedestal area, with possible influence on MHD unstable modes like ELMs and EHO. [1] Vdovin V. Plas Phys. Rep Vol.39 (2013), N2 95
        Speaker: Prof. Victor Vdovin (NRC Kurchatov Institute)
      • 08:30
        On the Equilibrium and Stability of ITER Relevant Plasmas with Flow 4h
        We present recent results on steady states of translational symmetric and axisymmetric ITER relevant plasmas with incompressible sheared flow in connection with a generalized Grad-Shafranov equation and on their stability [1-2]. The presentation includes equilibria either with monotonically increasing safety factor profiles pertinent to the L-H transition or with reversed magnetic shear. Linear and nonlinear solutions of the generalized Grad-Shafranov equation including non parallel flows of plasmas surrounded by a diverted boundary are constructed analytically, quasi analytically and numerically. It turns out that the electric field makes the safety factor flatter and increases the magnitude and shear of the toroidal velocity in qualitative agreement with experimental evidence on the formation of Internal Transport Barriers in tokamaks, thus indicating a stabilizing effect of the electric field. For parallel flows the linear stability is examined by applying a sufficient condition [3]. In this case one equilibrium corresponding to the H-state is potentially stable in the sense that the stability condition is satisfied in an appreciable part of the plasma while another solution corresponding to the L-state does not satisfy the condition. In the majority of the equilibria considered stabilization is caused by the variation of the magnetic field in the direction perpendicular to the magnetic surfaces (related to the magnetic shear) in conjunction with the sheared flow, depends on the plasma shaping and is sensitive to the up-down asymmetry. [1] G. N. Throumoulopoulos, H. Tasso, Phys. Plasmas 19, 014504 (2012). [2] AP Kuiroukidis and G. N. Throumoulopoulos, Phys. Plasmas 19, 022508 (2012); J. Plasma Phys. 79, 257 (2013); J. Plasma Phys. 80, 27 (2014); Phys. Plasmas 21, 032509 (2014). [3] G. N. Throumoulopoulos and H. Tasso, Phys. Plasmas 14, 122104 (2007).
        Speaker: Prof. George N. Throumoulopoulos (University of Ioannina)
        summary
      • 08:30
        On the Possibility of Alpha-Particle Confinement Study in ITER by NPA Measurements of Knock-on Ion Tails 4h
        One of the issues of the neutral particle diagnostics on ITER is to measure the distribution functions of the fast deuterium (D) and tritium (T) ions in MeV energy range. High energy tails in D,T-ion energy distributions (so-called knock-on ions) appear as a result of the close elastic collisions between the thermal fuel ions and the fusion alpha particles. The knock-on ion density depends directly on the density and energy distribution of the alpha particles. Therefore measurements of the neutralized knock-on D,T-ion fluxes escaping the plasma volume can provide information on the alpha particle confinement in DT-plasma. This report presents results of the numerical simulation for the neutralized fast D,T-ion fluxes in case of ITER fusion plasma and considers measurements of these fluxes in respect to neutral particle diagnostics and its capabilities on ITER.
        Speaker: Mr Vladislav Nesenevich (Ioffe Physical-Technical Institute of the Russian Academy of Sciences)
        summary
      • 08:30
        Physical Characteristics of Neoclassical Toroidal Viscosity in Tokamaks for Rotation Control and the Evaluation of Plasma Response 4h
        Favorable use of low magnitude (deltaB/B ~ O(10^-3)) three-dimensional (3D) magnetic fields in tokamaks includes mitigation of ELMs and Alfvénic modes, and alteration of the plasma rotation profile to strongly affect the stability of NTMs and RWMs. However, in ITER, these fields can significantly reduce the fusion gain, Q, by increasing alpha particle transport. These effects have been theoretically addressed using neoclassical toroidal viscosity (NTV) theory [K.C. Shaing and C.T. Hsu, Nucl. Fusion 54 (2014) 033012]. NTV magnitude and profile that determines the critical 3D applied field level for Q reduction, or for rotation feedback control, depends on the field spectrum, plasma collisionality, and plasma response to the field. The present work focuses on these critical questions with new analysis of results from NSTX and KSTAR. Experimental angular momentum alteration is directly compared to theoretical NTV torque density profiles, T_NTV, created by a range of applied 3D field spectra and plasma parameters in NSTX including configurations with dominant n = 2 and n = 3 field components. Large radial variations of T_NTV are found in ideal MHD models when the flux surface displacement is derived using an assumption of a fully penetrated deltaB. In contrast, experimentally measured T_NTV does not show strong torque localization. NSTX experiments yield a computed displacement ~ 0.3cm, smaller than the ion banana width, and averaging T_NTV over the banana width more closely matches the measured dL/dt profile. Results from a model-based rotation controller designed using NTV from applied 3D fields as an actuator for instability control are shown. A favorable observation for rotation control, clearly illustrated by KSTAR experiments, is the lack of hysteresis of the rotation when altered by non-resonant NTV. These experiments also show the theoretical scaling of T_NTV with deltaB^2 and ion temperature ~ T_i^2.5. Due to this strong dependence of the T_NTV profile on deltaB, the T_NTV measurements significantly constrain the allowable field amplification. Plasma response models being tested against experiment include the fully-penetrated deltaB model, and various physics models in the M3D-C1 resistive MHD code. Analysis shows that the M3D-C1 single-fluid model produces a flux surface-averaged |deltaB| consistent with the measured T_NTV.
        Speaker: Mr Steven Sabbagh (USA)
      • 08:30
        Plasma Confinement in the Trimix-3M Multipole Galatea Trap 4h
        Quasi-stationary plasma confinement with a high β value was implemented in the Trimix-3M multipole magnetic trap. In the trap, magnetic surfaces of complex cross-section delimit a closed toroidal region with zero magnetic field at its center. The lifetime of the plasma in the trap is ~1 ms and the average value of β is ~0.35. The Rogowski loop is used to measure the value of the toroidal current arising after the injection of a plasmoid through the magnetic crust of the Trimix-3M magnetic multipole trap. This current is due to plasma diamagnetism. A relation is established between the value of the diamagnetic current and the maximal plasma pressure realized on the separatrix of the magnetic field of the trap. It is thus shown that magnetic measurements in the multipole trap, for a known concentration value, allow us to determine the plasma temperature in the trap and the energy confinement time. Injection of a plasmoid is implemented through the magnetic crust of the trap. Both theory and experiment have shown that the depth of penetration of the plasmoid into a transverse magnetic field is proportional to the plasmoid energy and is inversely proportional to the magnetic pressure and the cross-section area of the plasmoid. This fact can be used to optimize the process of trapping a plasmoid.
        Speaker: Dr Andrey Bishaev (Moscow State Technical University of Radioengineering, Electronics and Automation, Moscow, Russia)
      • 08:30
        Plasma Vertical Stabilization in ITER 4h
        This paper describes the progress in analysis of the ITER plasma vertical stabilization (VS) system since its Design Review 2007 – 2008. Two indices characterising plasma vertical stabilization were studied. These are 1) the maximum value of plasma vertical displacement due to free drift that can be stopped by the VS system and 2) the maximum Root Mean Square value of low frequency noise in the dZ/dt “diagnostic” signal used in the VS feedback loop. The first VS index was calculated using the PET code for 15MA high-li low-beta plasmas. The second VS index was studied in the simulations of the most demanding for plasma magnetic control 15MA scenarios having the fastest plasma current ramp-up with early X-point formation, the fastest plasma current ramp-down in divertor configuration and the H to L mode transition at the current flattop. The simulations were performed from the beginning of the central solenoid discharge till the end of plasma current ramp-down using the DINA code with feedback control of the plasma current, position and shape, taking into account engineering limits imposed on the coils, power supplies and plasma-wall gaps. The studies performed demonstrate that the VS in-vessel coils, adopted recently in the baseline design, increase of the VS controllability range by about a factor of 6 providing operating margins sufficient to achieve ITER's goals.
        Speaker: Dr Yury Gribov (ITER Organization)
        summary
      • 08:30
        Predator-Prey Time Dynamics and Locking Control of Spontaneous Helical States in the RFP 4h
        Reversed Field Pinch (RFP) plasmas tend toward self-organized behavior depending on the nonlinear coupling between mutually interacting tearing modes. In multiple helicity plasmas one or more linearly unstable tearing modes may drive energy into stable modes through this coupling. In contrast, at high current and low density plasmas tend towards a state with a single dominant core mode. Although secondary modes are present, their amplitudes are reduced in this Quasi-Single Helicity (QSH) state. Recent work on modeling the shear-suppression mechanism has produced a predator-prey model of the QSH dynamics that reproduces the observed time dynamic behavior, in particular the increased persistence of the QSH state with increased plasma current. To diagnose these plasmas, we have established an error field control mechanism that locks the structure to a particular helical phase, to the advantage of the advanced diagnostic set on MST. With this diagnostic set, we have obtained evidence of helical structure in electron temperature, density, and impurity temperature.
        Speaker: Mr Mark Nornberg (USA)
        Poster
        summary
      • 08:30
        Progress in Active Control of Divertor Power Load in the EAST Superconducting Tokamak 4h
        Divertor power and particle exhaust is a critical issue facing the operation of next-step fusion devices such as ITER and DEMO. Active control of excessive heat and particle fluxes under high power steady-state plasma conditions has become a frontier hotspot in magnetic confinement fusion development. Significant progress has been made in controlling transient and stationary divertor heat fluxes in the Experimental Advanced Superconducting Tokamak (EAST) since the last IAEA-FEC, using many innovative techniques such as deuterium/lithium (D2/Li) pellet injection, Supersonic Molecular Beam Injection (SMBI), and integration of radiative divertor scenario with three-dimensional (3D) magnetic topology change induced by lower hybrid current drive (LHCD). We find that the injection of solid D2 pellets can directly induce divertor plasma detachment in the RF-heated plasmas, significantly reducing steady-state heat fluxes on the divertor target plates. Plasma detachment can be modulated periodically by multi-pellet injection. Furthermore, D2 pellet injection tends to trigger compound ELMs in H-modes, mitigating transient heat fluxes, compared to the standard Type I ELMs. In addition, Li granule injection has been demonstrated, for the first time in tokamaks, to be effective at triggering small ELMs with near 100% efficiency, thus providing a novel means for divertor power load control. ELM mitigation with SMBI has also been successfully achieved in EAST, significantly reducing the amplitude and increasing the frequency of ELMs. With multi-pulse SMBI, we demonstrated for the first time that the stationary divertor heat footprint can be actively modified by transferring heat from the outer strike point to the striated heat flux area in the far scrape-off layer, which is induced by LHCD. Similar results have been observed with divertor argon seeding. This provides an additional knob for the control of the stationary divertor power load, beyond or in addition to the achievement of highly radiating divertor conditions, which may be of great interest for future fusion devices such as ITER.
        Speaker: Dr Liang Wang (Institute of Plasma Physics, Chinese Academy of Sciences (ASIPP))
      • 08:30
        Progress in Snowflake Divertor Studies on TCV 4h
        Dissipating the energy emitted from a fusion reactor remains critical to achieving a commercially viable design with acceptable machine lifetime. Using the extreme shaping capability of TCV together with an ‘open-vessel’ design, the Snowflake (SF) configuration divertor was first demonstrated on TCV. Although a nearly ‘exact’ SF configuration was obtained, most of this work concentrated on SF configurations with a clear separation between the X-points (which may however vary during ELMs). This paper reports on the range of experiments performed over the last years on TCV, designed to understand the processes involved in power distribution between the four legs of a SF-divertor configuration by comparing the geometrical characteristics with measured power depositions. To date, these experiments are mostly performed at relatively low power density and particle density where the distribution to the divertor is expected to be dominated by transport parallel to the magnetic field lines and should be consistent with present divertor theory. Initial analysis of L-mode discharges indicated an enhanced transport into the private flux region and a reduction of peak heat. This enhanced transport, although relatively small, cannot be explained by the modified field line geometry alone and likely requires an additional or enhanced cross-field transport channel. A first attempt to model the configuration by matching the power and particle profiles at the primary strike points was unable to explain the observed power to the secondary strike points. This work was extended to both the L-mode and H-mode plasma confinement regimes. During ELM activity, up to ~20% of the exhausted energy was redistributed to the additional SPs and the peak heat flux to the inner primary was SP reduced by a factor of 2–3. Further avenues for progress, including an upgrade to the diagnostic array with the installation of a reciprocating probe and improved IR cameras, are in hand and, using the experience gained with SF configurations, research on other divertor geometries is under consideration. The need to find a working solution for fusion reactor exhaust and these highly encouraging TCV results demonstrate that this work in alternative divertor concepts is providing vital experimental and theoretical research information.
        Speaker: Dr Basil Duval (Ecole Polytechnique Fédérale de Lausanne – Centre des Recherches en Physique des Plasmas(CRPP), Association Euratom-Confédération Suisse(EPFL) CH-1015 Lausanne, Switzerland)
        summary
      • 08:30
        Recent Advances in the Understanding and Optimization of RMP ELM Suppression for ITER 4h
        Recent experiments with applied Resonant Magnetic Perturbations (RMPs) in low-collisionality ITER Similar Shape (ISS) plasmas on DIII-D have advanced the understanding of and increased confidence in obtaining ELM suppression in the ITER standard operating regime. ELM suppression is obtained with a reduced coil set (5–11 coils) on DIII-D, demonstrating the effectiveness of mixed harmonics (n=1,2,3) with a partial coil set and mitigating the risk of reduced coil availability on ITER. Recent advances in linear two-fluid MHD simulations indicate that resonant field penetration and amplification at the top of the pedestal is ubiquitous in these ISS plasmas, together with resonant field screening and kink amplification in the steep pressure gradient region. Measurements with the X-ray imaging camera reveal new information on the plasma response to 3D fields. There is good agreement between X-ray imaging and M3D-C1 simulation in the steep pressure gradient region of the pedestal, validating theoretical predictions of resonant screening and a dominant edge-kink response. While direct imaging of islands in ELM suppressed plasmas remains elusive, measurements with the newly upgraded magnetic sensors are suggestive of partially screened fields at the top of the pedestal, consistent with M3D-C1 simulations. Indirect evidence of island formation and resonant field penetration is also provided by the observed flattening of the electron pressure profile at the top of the pedestal and concomitant shrinkage of the pedestal width when the RMP is applied. In addition, the flutter model of electron transport also predicts an electron thermal diffusivity “hill” at the top of the pedestal that is comparable to experimental values when the resonant field amplification at the top of the pedestal is included in the calculation. Optimization of the pedestal pressure is an important issue for ELM suppression in ITER given that a reduction in the pedestal pressure is commonly observed in ISS plasmas with applied RMPs. Recent experiments demonstrate that the pedestal pressure can be maintained at the level before the onset of the RMP if the effect of density pumpout is counteracted with density feedback. This work was supported by the US Department of Energy under DE-AC02-09CH11466, DE-FG02-92ER54139, DE-FC02-04ER54698, DE-SC0007880, DE-FG02-07ER54917, and DE-AC05-00OR22725
        Speaker: Mr Mickey R. Wade (USA)
        Slides
      • 08:30
        Redefinition of the ITER Requirements and Diagnostics for Erosion, Deposition, Dust and Tritium Measurements Accounting for the Change to Tungsten Divertor 4h
        Dust and tritium inventories in the vacuum vessel have upper limits in ITER. Erosion, migration and re-deposition of wall material and co-deposition of fuel material are closely linked to the these inventories. The related suite of diagnostic and the respective set of plasma-wall-interaction physics related measurement requirements is now redefined as a whole because the decision to change from carbon to tungsten as divertor target material has been taken and the construction schedule requires developing the diagnostic concepts. This paper presents the result of this redefiniton.
        Speaker: Dr Roger Reichle (ITER Organization)
        summary
      • 08:30
        Simulation of the Pre-Thermal Quench Stage of Disruptions at Massive Gas Injection and Projections for ITER 4h
        During disruption mitigation by massive gas injection (MGI) the thermal energy is expected to be radiated with high efficiency in order to prevent excessive heat loads to first wall and divertor PFCs. The energy loss will take place in two phases: a) the so-called pre-thermal quench phase that lasts from the arrival of the first gas to the onset of increased transport due to MHD activity during b) the second phase the thermal quench (TQ). Quantification of the duration of the pre-TQ phase is essential for the design of the ITER disruption mitigation system (DMS). The DMS has to be designed such that the impurity amount accumulated during pre-TQ stage should be sufficient for re-radiation of more than 90% of heat flux at subsequent TQ phase of ITER disruption. The modelling with the code ASTRA together with ZIMPUR impurity transport and radiation code allows the description of the cooling process at the plasma edge, including the penetration of impurities and the shrinking of the current channel. Newly developed model for the gas flow at the end of delivery tube of MGI system well reproduce experimentally measured evolution. The validation of the simulation approach on available experimental data has demonstrated its ability to produce quantitative estimations of the pre-TQ stage duration and of the accumulated in the plasma amounts of Ar and deuterium under MGI. The comparison of the simulation results with experiment will allow the identification of how the pre-TQ duration scales with plasma minor and major radius, plasma current, thermal energy, plasma density and temperature profiles on which an extrapolation to ITER can be based. Simulation results of the pre-TQ stage in reference ITER scenarios are presented. The ability of the ITER MGI systems to provide injection of necessary impurity amount during pre-TQ stage are discussed.
        Speaker: Mr Victor Zhogolev (NRC "Kurchatov Institute")
      • 08:30
        Status of R&D for ITER Disruption Loads, Disruption Mitigation and Runaway Electron Avoidance 4h
        The energy stored during a burning plasma pulse in ITER will significantly exceed that in present devices. Rapid release of this energy during a disruption has the potential to cause surface melting of plasma-facing components (PFCs) and will cause high electromagnetic loads, in some cases close to the design limits. Heat load specifications for ITER, which enter, for example, in the design process for blanket modules and full-W divertor, are based on empirical data on footprint broadening, deposition time and confinement degradation prior to the thermal quench. Runaway electrons (RE) can cause localized high energy deposition and, under some circumstances, up to 300 MJ of magnetic energy could be converted to RE kinetic energy. Electromagnetic loads are quantified in terms of halo current and current quench time, for which a broad database exists. However, understanding the origin of current asymmetries, which can be particularly challenging for the mechanical structures, remains an open issue. To ensure the required lifetime of PFCs, therefore, reliable disruption prediction will be required to allow action for disruption avoidance or, as a last resort, to trigger the disruption mitigation system (DMS). Three systems are under consideration for the ITER DMS: massive gas injection (MGI), shattered pellet injection (SPI) and Be injection as a back-up option. MGI experiments have shown that electromagnetic and thermal load mitigation is feasible. However, it remains to be confirmed for the latter that 90% radiation efficiency, as envisaged for ITER, can be reliably achieved. RE mitigation remains an open issue. Several options like densification, position control, and destabilization of MHD during RE formation have been tested in recent years and their applicability was shown to be – if at all – limited for ITER. However, more recent experiments indicate that lower densities than predicted by the present theory can already suppress RE formation. Another promising effect is pitch-angle scattering on high-Z impurities, resulting in RE energy dissipation by synchrotron radiation. Simulations including this effect are in qualitative agreement with experimental observations. Overall, achieving mitigation of all disruption loads in ITER is challenging and will need a careful choice of injection technique, injected species, amount of material and injection timing.
        Speaker: Dr David Campbell (ITER Organization)
        summary
      • 08:30
        Studies of Impact of Edge Current Profiles, Plasma Shaping, Nonlinearity on Edge Localized Modes with BOUT++ Code 4h
        This work extends previous BOUT++ work to systematically study the edge current driven kink mode, the effect of plasma shaping on the ELMs, to benchmark with other codes, and to validate the BOUT++ nonlinear ELM simulation on EAST tokamak experiments. A sequences of equilibrium with different edge current is generated with the CORSICA code, by keeping total current and pressure profile fixed. With the edge current increasing, the dominant modes are changed from high-n ballooning modes to low-n kink modes. We found the edge current provides stabilizing effects on high-n ballooning modes, but not always provides the destabilizing effect on edge kink mode. We benchmarked BOUT++ linear results with GATO and ELITE codes. It is showed that the vacuum model has important effect on the edge kink mode calculation. The resistivity vacuum has destabilizing effect on both the kink modes and ballooning mode. Nonlinear calculation shows that with the edge current increasing, the linear growth rate of the ELM size decreases. However, at the final saturated stage after the nonlinear evolution, the ELM size increases with the edge current. We studied the role the plasma geometry, by choosing the circular, elongated and shaped with X-point equilibrium. The shaped plasma and the X-point geometry has stabilizing effect on the ELMs. We benchmarked those calculation with other codes. 3D ELMs nonlinear simulation on EAST tokamak has first been studied, based on the discharge #41019. The result shows that there are four phases during one ELM crash, which gives us a vivid dynamic process of the ELM crash. We also prove the exactness of the nonlinear simulation by comparing the results of experiment and simulation, including the energy loss and speed of the ELM effluxes. The experimental energy loss of ELM crash is consistent with the nonlinear simulation within same order of magnitude. The speed of the simulated ELM effluxes is comparable with the experimental data by Gas Puffing Image (GPI) diagnosis, and the ELM crashes start at the outer mid-plane.
        Speaker: Dr Guoqiang Li (Institute of Plasma Physica, Chinese Academy of Sciences)
        summary
      • 08:30
        Study of ITER First Plasma Initiation Using a 3D Electromagnetic Model 4h
        The First Plasma operation is expected to be performed in ITER without blanket modules (with temporary limiters), with a toroidal magnetic field of 2.65 T at 6.2 m (half of the nominal value), starting from a partly charged central solenoid (producing half of the nominal poloidal magnetic flux). The First Plasma initiation will be more challenging in ITER than in present tokamaks. One reason is due to high stray magnetic fields caused by 1.5 MA of eddy currents induced in the vacuum vessel and by 70 MA of currents in the central solenoid (CS). Moreover, the First Plasma initiation is complicated by a low value of the toroidal electric field (0.3 Vm-1), by a high volume of the vacuum vessel (1700 m3) and by a high content of impurities. On the other hand, several MW of ECRF heating will be available to assist plasma breakdown and burnthrough. Nevertheless, a carefully tuned poloidal field (PF) scenario will be essential to assure successful plasma initiation. The PF system should reduce the stray magnetic fields to about 2 mT in a large breakdown region with a minor radius of about 1.6 m and support a stable plasma equilibrium with increasing plasma current after the breakdown. Taking into account all of the above mentioned difficulties of plasma initiation in ITER, it is important to develop numerical tools to design a proper scenario of PF operation. So far, the design of scenarios of the PF system operation during plasma initiation and simulations of the plasma initiation were performed with the TRANSMAK code using 2D axisymmetric models of ITER conducting structures and 0D plasma transport model. In this paper, another approach is presented with first results of the design and simulation of PF scenarios of the First Plasma initiation using a 3D electromagnetic model developed on the basis of the TYPHOON codes. This model includes 3D conducting structures of the vacuum vessel and cryostat, the toroidal field (TF) coils, and the ferromagnetic inserts located between the shells of the vacuum vessel. The values of connection lengths confirm a good quality of the magnetic field null obtained using the 3D electromagnetic model: in the breakdown area with the effective minor radius 1.8 m the toroidally averaged components of the magnetic field is less than 2 mT and the connection length averaged over the boundary of this region is 2000 m.
        Speaker: Dr Anatoly Mineev (Joint Stock Company “D.V.Efremov Institute of Electrophysical Apparatus”)
        summary
      • 08:30
        Study of Pedestal Turbulence on EAST Tokamak 4h
        Pedestal turbulence was studied by a microwave reflectometry on EAST tokamak. The characteristics of edge pedestal turbulence during L-I-H transition, ELM-free H-mode phase and inter-ELM phase have been studied on EAST recently. (1) An edge spatial structure of density fluctuation and its dithering temporal evolution is observed on EAST tokamak during the L-H transition phase. (2) A coherent mode and/or a multi-harmonic mode often appears during the ELM-free phase before the first ELM on EAST tokamak. Analysis shows that there is a critical density for two modes on EAST. (3) In plasma with type-III ELMs, a precursor mode before ELM is usually observed. The frequency of the precursor was born about 150 kHz and gradually decreased up to the ELM. The mode amplitude increases or shows saturation before ELM. In the plasma with compound ELMs composed of high and low frequency ELMs, the precursor was also available before the high frequency ELM while the harmonic oscillations with frequencies of 20, 40 and 60 kHz appears before the low frequency ELM. This work was supported by National Magnetic Confinement Fusion Program of China (No. 2014GB106000 and 2014GB106003) and the National Natural Science Foundation of China (No. 11275234, 11305215, 11321092).
        Speaker: Prof. Xiang Gao (Institute of Plasma Physics, Chinese Academy of Sciences)
      • 08:30
        Successful ELM Suppressions in a Wide Range of q95 Using Low n RMPs in KSTAR and its Understanding as a Secondary Effect of RMP 4h
        As the most plausible technique to control the edge localized modes (ELMs) of high confinement mode (h-mode) plasmas, which is critical for ITER and beyond, non-axisymmetric resonant magnetic perturbations (RMPs) have been actively investigated in KSTAR. Since the first success of ELM suppression using n=1 RMPs in 2011 [1], our effort has been devoted to extend the operation regime of ELM control for both external magnetic configurations and plasma parameters. As results, it shows a possibility that the ELM-suppression can be achieved in a wide range of q95 (=3.5~6.5) if the RMP field is configured properly. Several unique features of RMP-driven ELM-suppression are found, by which the physics mechanisms of ELM suppression and mitigation can be explicitly distinguished. These are a long time delay (a secondary effect) and an improved confinement (transport bifurcation). Furthermore, a persistent, rapidly repeating bursty event, localized in the plasma edge, is observed and suggested as a key player in the underlying physics mechanism of RMP-driven ELM-suppression. [1] Y.M. Jeon et al., Phys. Rev. Lett. 109, 035004 (2012)
        Speaker: Mr YoungMu Jeon (Korea, Republic of)
      • 08:30
        The Latest Development of EAST Neutral Beam Injector 4h
        As the first full superconducting non-circular cross section tokamak in the world, EAST is used to explore the forefront physics and engineering issues on the construction of tokamak fusion reactor. Neutral beam injection has been recognized as one of the most effective means for plasma heating. According to the research plan of the EAST physics experiment, two sets of neutral beam injector (4-8MW,10-100s) will be built and operational in 2014. The paper presents the development of EAST neutral beam injector and the latest experiment results obtained on the test-stand about long pulse beam extraction and high power beam extraction, those results show that all targets reach or almost reach the design targets. All these will lay a solid foundation for the achievement of plasma heating and current drive for EAST in 2014.
        Speaker: Dr Yongjian Xu (Institute of Plasma Physics, Chinese Academy of Sciences)
      • 08:30
        Theoretical Analysis of the ICRH Antenna’s Impedance Matching for ELMy Plasmas on EAST 4h
        A well-optimized design of an ion cyclotron resonance heating (ICRH) antenna is very important for steady-state plasma heating with high radio frequency (RF) power of several tens of megawatts. However, a sharp decrease in the coupling RF power because of impedance mismatch of ICRH system is an issue that must be resolved for present-day fusion reactors and International Thermonuclear Experimental Reactor. This paper has theoretically analyzed the ICRH antenna’s impedance matching for ELMy plasmas on experimental advanced superconducting tokamak (EAST) by the transmission line theory. The results indicate that judicious choice of the optimal feeder location is found useful for adjustable capacitors’ tolerance to the variations of the antenna input impedance during edge-localized mode (ELM) discharge, which is expected to be good for the design of ICRH antenna system and for real-time feedback control during ELM discharge on EAST.
        Speaker: Prof. XUEYU GONG (Southwestern Institute of Physics)
      • 08:30
        X-Divertors in ITER - Without any Hardware Changes or Additions - and in Current Machines, and DEMO Reactors 4h
        We show that X-Divertors (XD), that mainly expand magnetic flux near the SOL peak heat flux strike point on the divertor plates, can be created for ITER, many existing tokamaks (e.g., SST-1, EAST, AUG, … ), and future Reactors (KDEMO, ARIES) without modifying their Poloidal Field (PF) coil sets (or adding internal coils). For ITER, KDEMO, SST-1, and EAST, we will present sequences of XDs, starting from their SDs, in which the flux expansions on the divertor plates increases smoothly. All these XDs ( including ITER) can be made without adding any internal PF coils, with all currents in existing PF coils under their maximum limits, have nearly the same strike points as the SD cases so no modifications of their divertors or pumping will be needed, and have core plasma shapes that maintain the specified conformity with their given first walls. Further, because XDs preferentially expand flux near the SOL strike point, they are expected to increase stability of detached plasmas, thus allowing a possible way to overcome “shadowing” issues arising from lower angles of field line incidence. Note that the incidence angle at higher flux expansions at SOL strike points can only be reduced by increased major radius of the Super X-Divertor (SXD). The demonstration of these XDs opens the possibility that they could be tested and, then, used to assist in high-power operation on ITER. Some further issues need examination, e.g., full time-dependent plasma startup and operation scenarios, vertical stability, etc. The SOLPS code will be used to gauge the effects of XDs on SOL, and possibly plasma detachment.
        Speaker: Dr Mike Kotschenreuther (University of Texas)
    • 10:15 10:45
      Coffee Break 30m
    • 10:45 12:30
      Core Turbulence: EX/2 & TH/1 Blue 1-5

      Blue 1-5

      Hotel Park Inn Pribaltiyskaya

      Saint Petersburg, Russian Federation
      Convener: Mr Evgeniy Gusakov (Russian Federation)
      • 10:45
        Dynamic Method to Study Turbulence and Turbulence Transport 20m
        Here we developed research methods of plasma turbulence transport associated with the non-local features. The ECH modulation experiment and the higher harmonic analysis of the heat wave indicated: (i) propagation of the change of Te at the time of switch-off/on of ECH power is about 5 times faster than that of perturbation itself, (ii) propagation of the higher (7th) harmonic of the Te perturbation is 5 times faster than prediction by the diffusive model. New bi-spectral analysis of fluctuations demonstrated a non-linear coupling of micro-fluctuations at different radial locations. These results are beneficial for control of plasma dynamics in future fusion reactors.
        Speaker: Mr Shigeru Inagaki (Japan)
        Slides
        summary
      • 11:05
        Turbulence Behavior and Transport Response Approaching Burning Plasma Relevant Parameters 20m
        Multi-scale turbulence properties are significantly altered and typically exhibit increased amplitude in high-beta inductive plasmas as parameters approach those anticipated in burning plasmas. These increases, observed with multiple fluctuation diagnostics in high performance H-mode plasmas on DIII-D, explain the consequent local transport and global energy time confinement response. Burning plasmas will exhibit equilibrated ion and electron temperatures, low average injected torque and toroidal rotation, as well as low ρ*, low ν* and low q95. Increased fluctuation amplitudes are observed as a result of reducing core toroidal rotation (and consequent ExB shear) and increasing Te/Ti. Density and temperature fluctuation measurements were obtained over a broad wavenumber and radial range, and provide a basis for quantitative comparisons with nonlinear simulations to test turbulent transport models. The energy confinement time is reduced by about 40% as the toroidal rotation is decreased by nearly a factor of three, while core turbulence increases in matched advanced-inductive plasmas (β≈2.7, q95=5.1). Density, electron and ion temperature profiles, as well as relevant dimensionless parameters (β, ρ*, q95, Te/Ti, and ν*) were maintained nearly fixed. Low-wavenumber (ion gyroradius scale) density fluctuations near mid-radius show significant amplitude reduction along with a slight reduction in radial correlation length at high rotation, while fluctuations in the outer region of the plasma, ρ>0.6, exhibit little change in amplitude. In related experiments, low-k density fluctuations are observed to increase over the radial range 0.3<ρ<0.8 as the Te/Ti ratio is raised by 25% towards unity via application of 3.5 MW of off-axis electron cyclotron heating. The spatial correlation properties are modified, suggesting a change in the dominant underlying instability driving the observed turbulence. Transport likewise increases in all channels while the energy confinement time is reduced by 30%. Initial modeling with TGLF shows significant changes to linear growth rates and saturated turbulence levels as Te/Ti is increased. This work was supported by the US Department of Energy under DE-FG02-08ER54999, DE-FG02-89ER53296, DE-FG02-07ER54917, DE-FG02-08ERA54984, DE-FC02-04ER54698, DE-FG02-04ER54235 and DE-AC02-09CH11466.
        Speaker: Mr George R. McKee (USA)
        Slides
      • 11:25
        Multi-Scale ITG/TEM/ETG Turbulence Simulations with Real Mass Ratio and Beta Value 20m
        Understanding electron heat transport is one of the critical issues in ITER. Although electron temperature gradient (ETG) driven turbulence has been regarded as a candidate of electron heat transport, recent multi-scale plasma turbulence simulations have reported that ion-scale instabilities such as ion temperature gradient modes (ITGs) and trapped electron modes (TEMs) dominate heat transport not only of ions but also of electrons. However, their simulations are limited to the reduced ion-to-electron mass ratios and electrostatic (beta=0) approximation, the following points are not yet clarified: (i) Are ETGs negligible even with the real mass ratio, where ion and electron scales are separated by a factor of the square root of the ion-to-electron mass ratio (~ 43)? (ii) Does ion-scale turbulence dominate heat transport even with the real beta value, where electromagnetic (finite-beta) effects stabilize ITGs but not ETGs? To resolve these issues, we have carried out multi-scale plasma turbulence simulations employing the real mass ratio and beta value, which is first realized by developing massively parallel finite-difference/spectral methods for the electromagnetic gyrokinetic simulation code GKV. Our study reveals real-mass-ratio and real-beta effects on multi-scale turbulence: (i) Ion-scale turbulence eliminates electron-scale turbulence and dominates electron heat transport even when employing the real mass ratio and beta value. (ii) When growth rates of ITGs are reduced by the finite-beta effects, full-spectrum analyses of multi-scale turbulence are essentially required. The results suggest that the simulations resolving only ion scales give good estimates only if ion-scale modes are highly unstable. On the other hand, in high-beta regimes where ion-scale modes are stabilized, proper treatments of the contributions of electron scales are important for evaluating transport levels and modeling turbulent transport.
        Speaker: Mr Shinya Maeyama (Japan)
        Slides
        summary
      • 11:45
        Controlling H-Mode Particle Transport with Modulated Electron Heating in DIII-D and Alcator C-Mod via TEM Turbulence 20m
        This work develops a quantitative understanding of the mechanisms for increased particle transport with electron heating in (quiescent) H-mode plasmas. Our DIII-D experiments demonstrate that H-mode core particle transport and density peaking can be locally controlled by modulated electron cyclotron heating (ECH). GYRO simulations show density gradient driven trapped electron modes (TEM) are the only unstable drift modes in the inner half-radius. Transport driven by TEMs increases strongly with electron temperature, reducing the density gradient during ECH. Thus α heating could reduce density peaking, self-regulating fusion power. The DIII-D experiments complement Alcator C-Mod experiments which controlled H mode core particle transport with modulated minority ICRF heating. High-resolution profile reflectometer density profiles and local fluctuation measurements were obtained in DIII D. Core density fluctuations intensify during electron heating in both C-Mod and DIII-D at TEM wavelengths. Several hundred GS2 nonlinear TEM simulations for Alcator C-Mod reproduce measured core density fluctuation levels using a new synthetic diagnostic, while matching inferred energy fluxes. A nonlinear upshift in the TEM critical density gradient, associated with zonal flow dominated states, increases strongly with collisionality. In the C-Mod experiments, the density gradient is clamped by this nonlinear TEM critical density gradient. The DIII-D experiments test our predicted strong collisionality variation of the TEM nonlinear upshift at an order of magnitude lower collisionality, while allowing Te/Ti to nearly double during ECH, tripling the TEM growth rate. Intermittent, quasi-coherent DBS fluctuations near ρ~0.33 grow stronger with ECH, with adjacent toroidal mode numbers characteristic of TEMs, accompanied by broadband turbulence. Density fluctuations from beam emission spectroscopy near ρ~0.7 strongly increase. Core carbon density and metallic line intensities are modulated by ECH, consistent with TEM expectations. Finally, profile stiffness tests were performed via gas puff modulation that varied the density gradient as well as collisionality. Supported by the U.S. Department of Energy under DE-FC02-08ER54966, DE-FC02-04ER54698, DE-AC02-09CH11466, DE-FG02-08ER54984, DE-FG02-07ER54917, DE-FG02-99ER54531, DE-FG02-89ER53296, DE-FG02-08ER54999, and DE-FC02-99ER54512.
        Speaker: Mr Darin Ernst (USA)
        Slides
        summary
      • 12:05
        Advances in Stellarator Gyrokinetics 20m
        We summarise a number of recent advances in gyrokinetic theory and simulations of microinstabilities and turbulence in stellarators, and compare with tokamaks. Trapped-electron modes (TEMs) can be very different in different types of devices, because these instabilities are excited by trapped electrons in regions of unfavourable magnetic curvature. Tokamaks and stellarators are fundamentally different in the sense that the regions of trapping and bad curvature overlap in tokamaks but need not do so in stellarators. In particular, quasi-isodynamic stellarators can have the property that the bounce-averaged curvature is favourable for the great majority of all trapped particles. Analytical theory suggests that TEMs are then stable in large parts of parameter space, and this prediction is confirmed by GENE simulations, which show that these modes are more stable the more the regions of bad curvature and magnetic trapping are separated from each other. As predicted by analytical theory, the electrons are far less destabilising in the stellarator. Not only is the growth rate lower, but the most unstable mode has a shorter perpendicular wavelength, which could result in less transport. Linear simulations of linear ITG modes in W7-X and LHD geometry using the EUTERPE show that the growth rate is similar in these two devices and that the mode structure is affected by details in the magnetic geometry. GENE simulations of saturated turbulence show that the density fluctuations are much less evenly distributed over the flux surface than in typical tokamaks. Instead of covering the entire outboard side of the torus, the turbulence is localised to narrow bands in regions of bad magnetic curvature. Possibly as a consequence of this localisation, the transport scaling is more sensitive to the normalised gyroradius and becomes stiffer when this parameter is small. This happens although these simulations are local in the radial direction, and the situation is different from that in tokamaks, where flux-tube simulations yield the same results as simulations of entire flux surfaces.
        Speaker: Mr Per Helander (Germany)
        Slides
        summary
    • 12:30 14:00
      Lunch Break 1h 30m
    • 14:00 16:10
      Pedestal Physics: EX/3 & TH/2 Blue 1-5

      Blue 1-5

      Hotel Park Inn Pribaltiyskaya

      Saint Petersburg, Russian Federation
      Convener: Dr Richard J. Buttery (General Atomics)
      • 14:00
        Understanding of Impurity Poloidal Distribution in Edge Pedestal by Modeling 20m
        Modeling of the impurities distribution in the pedestal region of ASDEX-Upgrade H-mode shot was done using the B2SOLPS5.2 transport code. Strong LFS-HFS asymmetry of B+5 density and strong decoupling between parallel velocity distribution of the impurities and the main ions was obtained in qualitative agreement with the experimental observations [1]. Similar asymmetry was obtained earlier in the modeling of He ions distribution in the MAST H-mode by the same code [2]. It is demonstrated that the observed phenomena could be understood by the impact of neoclassical effects in the presence of strong density gradients in the barrier region. Obtained results are completely different from the predictions of the standard neoclassical theory. In particular the role of the poloidal drift and parallel thermal force are important. The physical mechanisms can also explain observed phenomena on C-Mod. The poloidal LFS-HFS asymmetry of impurities changes their radial transport so the standard neoclassical transport model for impurities should be revisited in the edge barrier region. [1] E. Viezzer et al PPCF 55 (2013) 124037 [2] V. Rozhansky et al Proc.37th EPS Conf. on Contr. Fus. and Plasma Phys. v. 34A (2010) P2.190
        Speaker: Mr Vladimir Rozhansky (Russian Federation)
        Slides
        summary