An Overview of Research and Design Activities at CTFusion

NASA Astrophysics Data System (ADS)

Sutherland, D. A.; Jarboe, T. R.; Hossack, A. C.

2016-10-01

CTFusion, a newly formed company dedicated to the development of compact, toroidal fusion energy, is a spin-off from the University of Washington that will build upon the successes of the HIT-SI research program. The mission of the company to develop net-gain fusion power cores that will serve as the heart of economical fusion power plants or radioactive-waste destroying burner reactors. The overarching vision and development plan of the company will be presented, along with a detailed justification and design for our next device, the HIT-TD (Technology Demonstration) prototype. By externally driving the edge current and imposing non-axisymmetric magnetic perturbations, HIT-TD should demonstrate the sustainment of stable spheromak configurations with Imposed-Dynamo Current Drive (IDCD), as was accomplished in the HIT-SI device, with higher current gains and temperatures than previously possible. HIT-TD, if successful, will be an instrumental step along this path to economical fusion energy, and will serve as the stepping stone to our Proof-Of-Principle device (HIT-PoP). Beyond the implications of higher performance, sustained spheromaks for fusion applications, the HIT-TD platform will provide a unique system to observe plasma self-organizational phenomena of interest for other fusion devices, and astrophysical systems as well. Lastly, preliminary nuclear engineering design simulations with the MCNP6 code of the HIT-FNSF (Fusion Nuclear Science Facility) device will be presented.

Spherical torus fusion reactor

DOEpatents

Martin Peng, Y.K.M.

1985-10-03

The object of this invention is to provide a compact torus fusion reactor with dramatic simplification of plasma confinement design. Another object of this invention is to provide a compact torus fusion reactor with low magnetic field and small aspect ratio stable plasma confinement. In accordance with the principles of this invention there is provided a compact toroidal-type plasma confinement fusion reactor in which only the indispensable components inboard of a tokamak type of plasma confinement region, mainly a current conducting medium which carries electrical current for producing a toroidal magnet confinement field about the toroidal plasma region, are retained.

Recent progress in understanding electron thermal transport in NSTX

DOE PAGES

Ren, Y.; Belova, E.; Gorelenkov, N.; ...

2017-03-10

The anomalous level of electron thermal transport inferred in magnetically confined configurations is one of the most challenging problems for the ultimate realization of fusion power using toroidal devices: tokamaks, spherical tori and stellarators. It is generally believed that plasma instabilities driven by the abundant free energy in fusion plasmas are responsible for the electron thermal transport. The National Spherical Torus eXperiment (NSTX) (Ono et al 2000 Nucl. Fusion 40 557) provides a unique laboratory for studying plasma instabilities and their relation to electron thermal transport due to its low toroidal field, high plasma beta, low aspect ratio and largemore » ExB flow shear. Recent findings on NSTX have shown that multiple instabilities are required to explain observed electron thermal transport, given the wide range of equilibrium parameters due to different operational scenarios and radial regions in fusion plasmas. Here we review the recent progresses in understanding anomalous electron thermal transport in NSTX and focus on mechanisms that could drive electron thermal transport in the core region. The synergy between experiment and theoretical/ numerical modeling is essential to achieving these progresses. The plans for newly commissioned NSTX-Upgrade will also be discussed.« less

Locked modes in two reversed-field pinch devices of different size and shell system

NASA Astrophysics Data System (ADS)

Malmberg, J.-A.; Brunsell, P. R.; Yagi, Y.; Koguchi, H.

2000-10-01

The behavior of locked modes in two reversed-field pinch devices, the Toroidal Pinch Experiment (TPE-RX) [Y. Yagi et al., Plasma Phys. Control. Fusion 41, 2552 (1999)] and Extrap T2 [J. R. Drake et al., in Plasma Physics and Controlled Nuclear Fusion Research 1996, Montreal (International Atomic Energy Agency, Vienna, 1996), Vol. 2, p. 193] is analyzed and compared. The main characteristics of the locked mode are qualitatively similar. The toroidal distribution of the mode locking shows that field errors play a role in both devices. The probability of phase locking is found to increase with increasing magnetic fluctuation levels in both machines. Furthermore, the probability of phase locking increases with plasma current in TPE-RX despite the fact that the magnetic fluctuation levels decrease. A comparison with computations using a theoretical model estimating the critical mode amplitude for locking [R. Fitzpatrick et al., Phys. Plasmas 6, 3878 (1999)] shows a good correlation with experimental results in TPE-RX. In Extrap T2, the magnetic fluctuations scale weakly with both plasma current and electron densities. This is also reflected in the weak scaling of the magnetic fluctuation levels with the Lundquist number (˜S-0.06). In TPE-RX, the corresponding scaling is ˜S-0.18.

Global linear gyrokinetic simulation of energetic particle-driven instabilities in the LHD stellarator

DOE PAGES

Spong, Donald A.; Holod, Ihor; Todo, Y.; ...

2017-06-23

Energetic particles are inherent to toroidal fusion systems and can drive instabilities in the Alfvén frequency range, leading to decreased heating efficiency, high heat fluxes on plasma-facing components, and decreased ignition margin. The applicability of global gyrokinetic simulation methods to macroscopic instabilities has now been demonstrated and it is natural to extend these methods to 3D configurations such as stellarators, tokamaks with 3D coils and reversed field pinch helical states. This has been achieved by coupling the GTC global gyrokinetic PIC model to the VMEC equilibrium model, including 3D effects in the field solvers and particle push. Here, this papermore » demonstrates the application of this new capability to the linearized analysis of Alfvénic instabilities in the LHD stellarator. For normal shear iota profiles, toroidal Alfvén instabilities in the n = 1 and 2 toroidal mode families are unstable with frequencies in the 75 to 110 kHz range. Also, an LHD case with non-monotonic shear is considered, indicating reductions in growth rate for the same energetic particle drive. Finally, since 3D magnetic fields will be present to some extent in all fusion devices, the extension of gyrokinetic models to 3D configurations is an important step for the simulation of future fusion systems.« less

Toroidal reactor

DOEpatents

Dawson, John M.; Furth, Harold P.; Tenney, Fred H.

1988-12-06

Method for producing fusion power wherein a neutral beam is injected into a toroidal bulk plasma to produce fusion reactions during the time permitted by the slowing down of the particles from the injected beam in the bulk plasma.

Gasdynamic Mirror (GDM) Fusion Propulsion Engine Experiment

NASA Technical Reports Server (NTRS)

1999-01-01

The Gasdynamic Mirror, or GDM, is an example of a magnetic mirror-based fusion propulsion system. Its design is primarily consisting of a long slender solenoid surrounding a vacuum chamber that contains plasma. The bulk of the fusion plasma is confined by magnetic field generated by a series of toroidal-shaped magnets in the center section of the device. the purpose of the GDM Fusion Propulsion Experiment is to confirm the feasibility of the concept and to demonstrate many of the operational characteristics of a full-size plasma can be confined within the desired physical configuration and still reman stable. This image shows an engineer from Propulsion Research Technologies Division at Marshall Space Flight Center inspecting solenoid magnets-A, an integrate part of the Gasdynamic Mirror Fusion Propulsion Engine Experiment.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lazarus, E; Peng, Yueng Kay Martin

Oak Ridge National Laboratory (ORNL) proposes to build the Spherical Torus Experiment (STX), a very low aspect ratio toroidal confinement device. This proposal concentrates on tokamak operation of the experiment; however, it can in principle be operated as a pinch or reversed-field pinch as well. As a tokamak, the spherical torus confines a plasma that is characterized by high toroidal beta, low poloidal beta, large natural elongation, high plasma current for a given edge q, and strong paramagnetism. These features combine to offer the possibility of a compact, low-field fusion device. The figure below shows that when compared to amore » conventional tokamak the spherical torus represents a major change in geometry. The primary goals of the experiment will be to demonstrate a capability for high beta (20%) in the first stability regime, to extend our knowledge of tokamak confinement scaling, and to test oscillating-field current drive. The experiment will operate in the high-beta, collisionless regime, which is achieved in STX at low temperatures because of the geometry. At a minimum, operation of STX will help to resolve fundamental questions regarding the scaling of beta and confinement in tokamaks. Complete success in this program would have a significant impact on toroidal fusion research in that it would demonstrate solutions to the problems of beta and steady-state operation in the tokamak. The proposed device has a major radius of 0.45 m, a toroidai field of 0.5 T, a plasma current of 900 kA, and heating by neutral beam injection. We estimate 30 months for design, construction, and assembly. The budget estimate, including contingency and escalation, is $6.8 million.« less

Research Technology

NASA Image and Video Library

1999-05-12

The Gasdynamic Mirror, or GDM, is an example of a magnetic mirror-based fusion propulsion system. Its design is primarily consisting of a long slender solenoid surrounding a vacuum chamber that contains plasma. The bulk of the fusion plasma is confined by magnetic field generated by a series of toroidal-shaped magnets in the center section of the device. the purpose of the GDM Fusion Propulsion Experiment is to confirm the feasibility of the concept and to demonstrate many of the operational characteristics of a full-size plasma can be confined within the desired physical configuration and still reman stable. This image shows an engineer from Propulsion Research Technologies Division at Marshall Space Flight Center inspecting solenoid magnets-A, an integrate part of the Gasdynamic Mirror Fusion Propulsion Engine Experiment.

History of Nuclear Fusion Research in Japan

NASA Astrophysics Data System (ADS)

Iguchi, Harukazu; Matsuoka, Keisuke; Kimura, Kazue; Namba, Chusei; Matsuda, Shinzaburo

In the late 1950s just after the atomic energy research was opened worldwide, there was a lively discussion among scientists on the strategy of nuclear fusion research in Japan. Finally, decision was made that fusion research should be started from the basic, namely, research on plasma physics and from cultivation of human resources at universities under the Ministry of Education, Science and Culture (MOE). However, an endorsement was given that construction of an experimental device for fusion research would be approved sooner or later. Studies on toroidal plasma confinement started at Japan Atomic Energy Research Institute (JAERI) under the Science and Technology Agency (STA) in the mid-1960s. Dualistic fusion research framework in Japan was established. This structure has lasted until now. Fusion research activities over the last 50 years are described by the use of a flowchart, which is convenient to glance the historical development of fusion research in Japan.

Spherical torus fusion reactor

DOEpatents

Peng, Yueng-Kay M.

1989-04-04

A fusion reactor is provided having a near spherical-shaped plasma with a modest central opening through which straight segments of toroidal field coils extend that carry electrical current for generating a toroidal magnet plasma confinement fields. By retaining only the indispensable components inboard of the plasma torus, principally the cooled toroidal field conductors and in some cases a vacuum containment vessel wall, the fusion reactor features an exceptionally small aspect ratio (typically about 1.5), a naturally elongated plasma cross section without extensive field shaping, requires low strength magnetic containment fields, small size and high beta. These features combine to produce a spherical torus plasma in a unique physics regime which permits compact fusion at low field and modest cost.

Spherical torus fusion reactor

DOEpatents

Peng, Yueng-Kay M.

1989-01-01

A fusion reactor is provided having a near spherical-shaped plasma with a modest central opening through which straight segments of toroidal field coils extend that carry electrical current for generating a toroidal magnet plasma confinement fields. By retaining only the indispensable components inboard of the plasma torus, principally the cooled toroidal field conductors and in some cases a vacuum containment vessel wall, the fusion reactor features an exceptionally small aspect ratio (typically about 1.5), a naturally elongated plasma cross section without extensive field shaping, requires low strength magnetic containment fields, small size and high beta. These features combine to produce a spherical torus plasma in a unique physics regime which permits compact fusion at low field and modest cost.

ELMO Bumpy Square proposal

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dory, R.A.; Uckan, N.A.; Ard, W.B.

The ELMO Bumpy Square (EBS) concept consists of four straight magnetic mirror arrays linked by four high-field corner coils. Extensive calculations show that this configuration offers major improvements over the ELMO Bumpy Torus (EBT) in particle confinement, heating, transport, ring production, and stability. The components of the EBT device at Oak Ridge National Laboratory can be reconfigured into a square arrangement having straight sides composed of EBT coils, with new microwave cavities and high-field corners designed and built for this application. The elimination of neoclassical convection, identified as the dominant mechanism for the limited confinement in EBT, will give themore » EBS device substantially improved confinement and the flexibility to explore the concepts that produce this improvement. The primary goals of the EBS program are twofold: first, to improve the physics of confinement in toroidal systems by developing the concepts of plasma stabilization using the effects of energetic electrons and confinement optimization using magnetic field shaping and electrostatic potential control to limit particle drift, and second, to develop bumpy toroid devices as attractive candidates for fusion reactors. This report presents a brief review of the physics analyses that support the EBS concept, discussions of the design and expected performance of the EBS device, a description of the EBS experimental program, and a review of the reactor potential of bumpy toroid configurations. Detailed information is presented in the appendices.« less

Current-carrying element based on second-generation high-temperature superconductor for the magnet system of a fusion neutron source

NASA Astrophysics Data System (ADS)

Novikov, M. S.; Ivanov, D. P.; Novikov, S. I.; Shuvaev, S. A.

2015-12-01

Application of current-carrying elements (CCEs) made of second-generation high-temperature superconductor (2G HTS) in magnet systems of a fusion neutron source (FNS) and other fusion devices will allow their magnetic field and thermodynamic stability to be increased substantially in comparison with those of low-temperature superconductor (LTS) magnets. For a toroidal magnet of the FNS, a design of a helical (partially transposed) CCE made of 2G HTS is under development with forced-flow cooling by helium gas, a current of 20-30 kA, an operating temperature of 10-20 K, and a magnetic field on the winding of 12-15 T (prospectively ~20 T). Short-sized samples of the helical flexible heavy-current CCE are being fabricated and investigated; a pilot-line unit for production of long-sized CCE pieces is under construction. The applied fabrication technique allows the CCE to be produced which combines a high operating current, thermal and mechanical stability, manufacturability, and low losses in the alternating modes. The possibility of fabricating the CCE with the outer dimensions and values of the operating parameter required for the FNS (and with a significant margin) using already available serial 2G HTS tapes is substantiated. The maximum field of toroidal magnets with CCEs made of 2G HTS will be limited only by mechanical properties of the magnet's casing and structure, while the thermal stability will be approximately two orders of magnitude higher than that of toroidal magnets with LTS-based CCEs. The helical CCE made of 2G HTS is very promising for fusion and hybrid electric power plants, and its design and technologies of production, as well as the prototype coils made of it for the FNS and other tokamaks, are worth developing now.

Progress in extrapolating divertor heat fluxes towards large fusion devices

NASA Astrophysics Data System (ADS)

Sieglin, B.; Faitsch, M.; Eich, T.; Herrmann, A.; Suttrop, W.; Collaborators, JET; the MST1 Team; the ASDEX Upgrade Team

2017-12-01

Heat load to the plasma facing components is one of the major challenges for the development and design of large fusion devices such as ITER. Nowadays fusion experiments can operate with heat load mitigation techniques, e.g. sweeping, impurity seeding, but do not generally require it. For large fusion devices however, heat load mitigation will be essential. This paper presents the current progress of the extrapolation of steady state and transient heat loads towards large fusion devices. For transient heat loads, so-called edge localized modes are considered a serious issue for the lifetime of divertor components. In this paper, the ITER operation at half field (2.65 T) and half current (7.5 MA) will be discussed considering the current material limit for the divertor peak energy fluence of 0.5 {MJ}/{{{m}}}2. Recent studies were successful in describing the observed energy fluence in the JET, MAST and ASDEX Upgrade using the pedestal pressure prior to the ELM crash. Extrapolating this towards ITER results in a more benign heat load compared to previous scalings. In the presence of magnetic perturbation, the axisymmetry is broken and a 2D heat flux pattern is induced on the divertor target, leading to local increase of the heat flux which is a concern for ITER. It is shown that for a moderate divertor broadening S/{λ }{{q}}> 0.5 the toroidal peaking of the heat flux disappears.

Pushing Particles with Waves: Current Drive and α-Channeling

DOE PAGES

FISCH, Nathaniel J.

2016-01-01

It can be advantageous to push particles with waves in tokamaks or other magnetic confinement devices, relying on wave-particle resonances to accomplish specific goals. Waves that damp on electrons or ions in toroidal fusion devises can drive currents if the waves are launched with toroidal asymmetry. Theses currents are important for tokamaks, since they operate in the absence of an electric field with curl, enabling steady state operation. The lower hybrid wave and the electron cyclotron wave have been demonstrated to drive significant currents. Non-inductive current also stabilizes deleterious tearing modes. Waves can also be used to broker the energymore » transfer between energetic alpha particles and the background plasma. Alpha particles born through fusion reactions in a tokamak reactor tend to slow down on electrons, but that could take up to hundreds of milliseconds. Before that happens, the energy in these alpha particles can destabilize on collisionless timescales toroidal Alfven modes and other waves, in a way deleterious to energy confinement. However, it has been speculated that this energy might be instead be channeled instead into useful energy, that heats fuel ions or drives current. Furthermore, an important question is the extent to which these effects can be accomplished together.« less

Impact of physics and technology innovations on compact tokamak fusion pilot plants

NASA Astrophysics Data System (ADS)

Menard, Jonathan

2016-10-01

For magnetic fusion to be economically attractive and have near-term impact on the world energy scene it is important to focus on key physics and technology innovations that could enable net electricity production at reduced size and cost. The tokamak is presently closest to achieving the fusion conditions necessary for net electricity at acceptable device size, although sustaining high-performance scenarios free of disruptions remains a significant challenge for the tokamak approach. Previous pilot plant studies have shown that electricity gain is proportional to the product of the fusion gain, blanket thermal conversion efficiency, and auxiliary heating wall-plug efficiency. In this work, the impact of several innovations is assessed with respect to maximizing fusion gain. At fixed bootstrap current fraction, fusion gain varies approximately as the square of the confinement multiplier, normalized beta, and major radius, and varies as the toroidal field and elongation both to the third power. For example, REBCO high-temperature superconductors (HTS) offer the potential to operate at much higher toroidal field than present fusion magnets, but HTS cables are also beginning to access winding pack current densities up to an order of magnitude higher than present technology, and smaller HTS TF magnet sizes make low-aspect-ratio HTS tokamaks potentially attractive by leveraging naturally higher normalized beta and elongation. Further, advances in kinetic stabilization and feedback control of resistive wall modes could also enable significant increases in normalized beta and fusion gain. Significant reductions in pilot plant size will also likely require increased plasma energy confinement, and control of turbulence and/or low edge recycling (for example using lithium walls) would have major impact on fusion gain. Reduced device size could also exacerbate divertor heat loads, and the impact of novel divertor solutions on pilot plant configurations is addressed. For missions including tritium breeding, high-thermal-efficiency liquid metal breeding blankets are attractive, and novel immersion blankets offer the potential for simplified fabrication and maintenance and reduced cost. Lastly, the optimal aspect ratio for a tokamak pilot plant is likely a function of the device mission and associated cost, with low aspect ratio favored for minimizing TF magnet mass and higher aspect ratio favored for minimizing blanket mass. The interplay between a range of physics and technology innovations for enabling compact pilot plants will be described. This work was supported by U.S. DOE Contract Number DE-AC02-09CH11466.

Fast ion motion in the plasma part of a stellarator-mirror fission-fusion hybrid

NASA Astrophysics Data System (ADS)

Moiseenko, V. E.; Nemov, V. V.; Ågren, O.; Kasilov, S. V.; Garkusha, I. E.

2016-06-01

Recent developments of a stellarator-mirror (SM) fission-fusion hybrid concept are reviewed. The hybrid consists of a fusion neutron source and a powerful sub-critical fast fission reactor core. The aim is transmutation of spent nuclear fuel and safe fission energy production. In its fusion part, a stellarator-type system with an embedded magnetic mirror is used. The stellarator confines deuterium plasma with moderate temperature, 1-2 keV. In the magnetic mirror, a hot component of sloshing tritium ions is trapped. There, the fusion neutrons are generated. A candidate for a combined SM system is a DRACON magnetic trap. A basic idea behind an SM device is to maintain local neutron production in a mirror part, but at the same time eliminate the end losses by using a toroidal device. A possible drawback is that the stellarator part can introduce collision-free radial drift losses, which is the main topic for this study. For high energy ions of tritium with an energy of 70 keV, comparative computations of collisionless losses in the rectilinear part of a specific design of the DRACON type trap are carried out. Two versions of the trap are considered with different lengths of the rectilinear sections. Also the total number of current-carrying rings in the magnetic system is varied. The results predict that high energy ions from neutral beam injection can be satisfactorily confined in the mirror part during 0.1-1 s. The Uragan-2M experimental device is used to check key points of the SM concept. The magnetic configuration of a stellarator with an embedded magnetic mirror is arranged in this device by switching off one toroidal coil. The motion of particles magnetically trapped in the embedded mirror is analyzed numerically with use of motional invariants. It is found that without radial electric field particles quickly drift out of the SM, even if the particles initially are located on a nested magnetic surface. We will show that a weak radial electric field, which would be spontaneously created by the ambipolar radial particle losses, can make drift trajectories closed, which substantially improves particle confinement. It is remarkable that the improvement acts both for positive and negative charges.

3D toroidal physics: testing the boundaries of symmetry breaking

NASA Astrophysics Data System (ADS)

Spong, Don

2014-10-01

Toroidal symmetry is an important concept for plasma confinement; it allows the existence of nested flux surface MHD equilibria and conserved invariants for particle motion. However, perfect symmetry is unachievable in realistic toroidal plasma devices. For example, tokamaks have toroidal ripple due to discrete field coils, optimized stellarators do not achieve exact quasi-symmetry, the plasma itself continually seeks lower energy states through helical 3D deformations, and reactors will likely have non-uniform distributions of ferritic steel near the plasma. Also, some level of designed-in 3D magnetic field structure is now anticipated for most concepts in order to lead to a stable, steady-state fusion reactor. Such planned 3D field structures can take many forms, ranging from tokamaks with weak 3D ELM-suppression fields to stellarators with more dominant 3D field structures. There is considerable interest in the development of unified physics models for the full range of 3D effects. Ultimately, the questions of how much symmetry breaking can be tolerated and how to optimize its design must be addressed for all fusion concepts. Fortunately, significant progress is underway in theory, computation and plasma diagnostics on many issues such as magnetic surface quality, plasma screening vs. amplification of 3D perturbations, 3D transport, influence on edge pedestal structures, MHD stability effects, modification of fast ion-driven instabilities, prediction of energetic particle heat loads on plasma-facing materials, effects of 3D fields on turbulence, and magnetic coil design. A closely coupled program of simulation, experimental validation, and design optimization is required to determine what forms and amplitudes of 3D shaping and symmetry breaking will be compatible with future fusion reactors. The development of models to address 3D physics and progress in these areas will be described. This work is supported both by the US Department of Energy under Contract DE-AC05-00OR22725 with UT-Battelle, LLC and under the US DOE SciDAC GSEP Center.

Ion cyclotron emission studies: Retrospects and prospects

DOE PAGES

Gorelenkov, N. N.

2016-06-05

Ion cyclotron emission (ICE) studies emerged in part from the papers by A.B. Mikhailovskii published in the 1970s. Among the discussed subjects were electromagnetic compressional Alfv,nic cyclotron instabilities with the linear growth rate similar ~ √(n α/n e) driven by fusion products, -particles which draw a lot of attention to energetic particle physics. The theory of ICE excited by energetic particles was significantly advanced at the end of the 20th century motivated by first DT experiments on TFTR and subsequent JET experimental studies which we highlight. Recently ICE theory was advanced by detailed theoretical and experimental studies on spherical torusmore » (ST) fusion devices where the instability signals previously indistinguishable in high aspect ratio tokamaks due to high toroidal magnetic field became the subjects of experiments. Finally, we discuss prospects of ICE theory applications for future burning plasma (BP) experiments such as those to be conducted in ITER device in France, where neutron and gamma rays escaping the plasma create extremely challenging conditions fusion alpha particle diagnostics.« less

Skyshine study for next generation of fusion devices

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gohar, Y.; Yang, S.

1987-02-01

A shielding analysis for next generation of fusion devices (ETR/INTOR) was performed to study the dose equivalent outside the reactor building during operation including the contribution from neutrons and photons scattered back by collisions with air nuclei (skyshine component). Two different three-dimensional geometrical models for a tokamak fusion reactor based on INTOR design parameters were developed for this study. In the first geometrical model, the reactor geometry and the spatial distribution of the deuterium-tritium neutron source were simplified for a parametric survey. The second geometrical model employed an explicit representation of the toroidal geometry of the reactor chamber and themore » spatial distribution of the neutron source. The MCNP general Monte Carlo code for neutron and photon transport was used to perform all the calculations. The energy distribution of the neutron source was used explicitly in the calculations with ENDF/B-V data. The dose equivalent results were analyzed as a function of the concrete roof thickness of the reactor building and the location outside the reactor building.« less

Ion cyclotron emission studies: Retrospects and prospects

NASA Astrophysics Data System (ADS)

Gorelenkov, N. N.

2016-05-01

Ion cyclotron emission (ICE) studies emerged in part from the papers by A.B. Mikhailovskii published in the 1970s. Among the discussed subjects were electromagnetic compressional Alfvénic cyclotron instabilities with the linear growth rate √ {n_α /n_e } driven by fusion products, -particles which draw a lot of attention to energetic particle physics. The theory of ICE excited by energetic particles was significantly advanced at the end of the 20th century motivated by first DT experiments on TFTR and subsequent JET experimental studies which we highlight. More recently ICE theory was advanced by detailed theoretical and experimental studies on spherical torus (ST) fusion devices where the instability signals previously indistinguishable in high aspect ratio tokamaks due to high toroidal magnetic field became the subjects of experiments. We discuss further prospects of ICE theory applications for future burning plasma (BP) experiments such as those to be conducted in ITER device in France, where neutron and gamma rays escaping the plasma create extremely challenging conditions fusion alpha particle diagnostics.

EDITORIAL: The Nuclear Fusion Award The Nuclear Fusion Award

NASA Astrophysics Data System (ADS)

Kikuchi, M.

2011-01-01

The Nuclear Fusion Award ceremony for 2009 and 2010 award winners was held during the 23rd IAEA Fusion Energy Conference in Daejeon. This time, both 2009 and 2010 award winners were celebrated by the IAEA and the participants of the 23rd IAEA Fusion Energy Conference. The Nuclear Fusion Award is a paper prize to acknowledge the best distinguished paper among the published papers in a particular volume of the Nuclear Fusion journal. Among the top-cited and highly-recommended papers chosen by the Editorial Board, excluding overview and review papers, and by analyzing self-citation and non-self-citation with an emphasis on non-self-citation, the Editorial Board confidentially selects ten distinguished papers as nominees for the Nuclear Fusion Award. Certificates are given to the leading authors of the Nuclear Fusion Award nominees. The final winner is selected among the ten nominees by the Nuclear Fusion Editorial Board voting confidentially. 2009 Nuclear Fusion Award nominees For the 2009 award, the papers published in the 2006 volume were assessed and the following papers were nominated, most of which are magnetic confinement experiments, theory and modeling, while one addresses inertial confinement. Sabbagh S.A. et al 2006 Resistive wall stabilized operation in rotating high beta NSTX plasmas Nucl. Fusion 46 635-44 La Haye R.J. et al 2006 Cross-machine benchmarking for ITER of neoclassical tearing mode stabilization by electron cyclotron current drive Nucl. Fusion 46 451-61 Honrubia J.J. et al 2006 Three-dimensional fast electron transport for ignition-scale inertial fusion capsules Nucl. Fusion 46 L25-8 Ido T. et al 2006 Observation of the interaction between the geodesic acoustic mode and ambient fluctuation in the JFT-2M tokamak Nucl. Fusion 46 512-20 Plyusnin V.V. et al 2006 Study of runaway electron generation during major disruptions in JET Nucl. Fusion 46 277-84 Pitts R.A. et al 2006 Far SOL ELM ion energies in JET Nucl. Fusion 46 82-98 Berk H.L. et al 2006 Explanation of the JET n = 0 chirping mode Nucl. Fusion 46 S888-97 Urano H. et al 2006 Confinement degradation with beta for ELMy HH-mode plasmas in JT-60U tokamak Nucl. Fusion 46 781-7 Izzo V.A. et al 2006 A numerical investigation of the effects of impurity penetration depth on disruption mitigation by massive high-pressure gas jet Nucl. Fusion 46 541-7 Inagaki S. et al 2006 Comparison of transient electron heat transport in LHD helical and JT-60U tokamak plasmas Nucl. Fusion 46 133-41 Watanabe T.-H. et al 2006 Velocity-space structures of distribution function in toroidal ion temperature gradient turbulence Nucl. Fusion 46 24-32 2010 Nuclear Fusion Award nominees For the 2010 award, the papers published in the 2007 volume were assessed and the following papers were nominated, all of which are magnetic confinement experiments and theory. Rice J.E. et al 2007 Inter-machine comparison of intrinsic toroidal rotation in tokamaks Nucl. Fusion 47 1618-24 Lipschultz B. et al 2007 Plasma-surface interaction, scrape-off layer and divertor physics: implications for ITER Nucl. Fusion 47 1189-205 Loarer T. et al 2007 Gas balance and fuel retention in fusion devices Nucl. Fusion 47 1112-20 Garcia O.E et al 2007 Fluctuations and transport in the TCV scrape-off layer Nucl. Fusion 47 667-76 Zonca F. et al 2007 Electron fishbones: theory and experimental evidence Nucl. Fusion 47 1588-97 Maggi C.F. et al 2007 Characteristics of the H-mode pedestal in improved confinement scenarios in ASDEX Upgrade, DIII-D, JET and JT-60U Nucl. Fusion 47 535-51 Yoshida M. et al 2007 Momentum transport and plasma rotation profile in toroidal direction in JT-60U L-mode plasmas Nucl. Fusion 47 856-63 Zohm H. et al 2007 Control of MHD instabilities by ECCD: ASDEX Upgrade results and implications for ITER Nucl. Fusion 47 228-32 Snyder P.B. et al 2007 Stability and dynamics of the edge pedestal in the low collisionality regime: physics mechanisms for steady-state ELM-free operation Nucl. Fusion 47 961-8 Urano H. et al 2007 H-mode pedestal structure in the variation of toroidal rotation and toroidal field ripple in JT-60U Nucl. Fusion 47 706-13 Günter S. et al 2007 Interaction of energetic particles with large and small scale instabilities Nucl. Fusion 47 920-8

Tokamaks

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wesson, J

1987-01-01

The word tokamak derives from the Russian term, toroidalnaya kamera magnitaya (toroidal chamber magnetic). The device was invented in the Soviet Union in 1950 and has since developed into one of the chief ways in which it is hoped to obtain usable power from plasmas through thermonuclear fusion. The present is meant to be an introduction to those entering the field, to those already engaged in research, and to those who want to gain some understanding of what it's all about.

Improvement of Current Drive Efficiency in Projected FNSF Discharges

NASA Astrophysics Data System (ADS)

Prater, R.; Chan, V.; Garofalo, A.

2012-10-01

The Fusion Nuclear Science Facility - Advanced Tokamak (FNSF-AT) is envisioned as a facility that uses the tokamak approach to address the development of the AT path to fusion and fusion's energy objectives. It uses copper coils for a compact device with high βN and moderate power gain. The major radius is 2.7 m and central toroidal field is 5.44 T. Achieving the required confinement and stability at βN˜3.7 requires a current profile with negative central shear and qmin>1. Off-axis Electron Cyclotron Current Drive (ECCD), in addition to high bootstrap current fraction, can help support this current profile. Using the applied EC frequency and launch location as free parameters, a systematic study has been carried out to optimize the ECCD in the range ρ= 0.5-0.7. Using a top launch, making use of a large toroidal component to the launch direction, adjusting the vertical launch angle so that the rays propagate nearly parallel to the resonance, and adjusting the frequency for optimum total current give a high dimensionless efficiency of 0.44 for a broad ECCD profile peaked at ρ=0.7, and the driven current is 17 kA/MW for n20= 2.1 and Te= 10.3 keV locally.

HIGH TEMPERATURE SYSTEMS

DOEpatents

Thomson, G.P.; Blackman, M.

1961-07-25

BS>A device is descrined for producing nuclear fusion reactions by additional acceleration of a hydrogen isotope plasma formed and initially accelerated by a collapsing magnetic field. The plasma is enclosed in a toroidal cavity within a vessel composed of a plurality of insulated coaxial segments. The added acceleration is caused by providing progressing potentials to the insulated segments acting as electrodes by means of a segmented delay transmission line coupled to the electrode segments and excited by a two phase alternating current supply.

Tritium assay of Li/sub 2/O in the LBM/LOTUS experiments

DOE Office of Scientific and Technical Information (OSTI.GOV)

Quanci, J.; Azam, S.; Bertone, P.

1986-11-01

The Lithium Blanket Module (LBM) is an assembly of over 20,000 cylindrical lithium oxide pellets in an array representative of a limited-coverage breeding zone for a toroidal fusion device. A principal objective of the LBM program is to test the ability of advanced neutronics coding to model the tritium breeding characteristics of a fusion device blanket. The LBM has been irradiated at the Ecole Polytechnique Federale de Lausanne (EPFL) LOTUS facility with a 14 MeV point-neutron source. Princeton Plasma Physics Laboratory (PPPL) and EPFL assayed the tritium bred in lithium oxide diagnostic samples placed at various positions in the LBM.more » PPPL employed a thermal extraction technique while EPFL used a dissolution method. The results for the assay are reported and compared to MCNP Monte Carlo neutronics calculations for the LBM/LOTUS system.« less

Overview of ECRH experimental results

NASA Astrophysics Data System (ADS)

Lloyd, Brian

1998-08-01

A review of the present status of electron cyclotron heating and current drive experiments in toroidal fusion devices is presented. In addition to basic heating and current drive studies the review also addresses advances in wave physics and the application of electron cyclotron waves for instability control, transport studies, pre-ionization/start-up assist, etc. A comprehensive overview is given with particular emphasis on recent advances since the major review of Erckmann and Gasparino (1994) ( 36 1869), including results from the latest generation of high-power, high-frequency experiments.

Experimental Studies of Compact Toroidal Plasma on BCTX

NASA Astrophysics Data System (ADS)

Morse, Edward C.; Coomer, Eric D.; Hartman, Charles W.

1998-11-01

The Berkeley Compact Toroid Experiment (BCTX) is a spheromak-type magnetically confined fusion confinement experiment. The plasma is formed using a Marshall gun and injected into a 70 cm diameter copper flux conserver. The BCTX device has an RF heating sy stem which can deliver twenty megawatts of RF power for 100 μs pulse length. The RF system operates at 450 MHz, and energy is coupled into the plasma by lower hybrid waves. The purpose of the experiment is to assess the energy-confining capability of the spheromak plasma configuration by using the RF power as a heat pulse and determining the decay rate of the plasma temperature following the heat pulse. Electron temperatures up to 150 eV have been measured in BCTX using Thomson scattering. Core dens ities have been measured with the Raman-calibrated Thomson system in the 2 arrow 5 × 10^14 per cc range. Other diagnostics include magnetic probes, a laser interferometer electron density measurement, three UV spectrometers for impurity l ine radiation, and an ion Doppler temperature measurement. Some data will be presented which shows the effects of an axial pinch being present in the device, giving the device a nonzero q at the wall.

System and method for generating steady state confining current for a toroidal plasma fusion reactor

DOEpatents

Fisch, Nathaniel J.

1981-01-01

A system for generating steady state confining current for a toroidal plasma fusion reactor providing steady-state generation of the thermonuclear power. A dense, hot toroidal plasma is initially prepared with a confining magnetic field with toroidal and poloidal components. Continuous wave RF energy is injected into said plasma to establish a spectrum of traveling waves in the plasma, where the traveling waves have momentum components substantially either all parallel, or all anti-parallel to the confining magnetic field. The injected RF energy is phased to couple to said traveling waves with both a phase velocity component and a wave momentum component in the direction of the plasma traveling wave components. The injected RF energy has a predetermined spectrum selected so that said traveling waves couple to plasma electrons having velocities in a predetermined range .DELTA.. The velocities in the range are substantially greater than the thermal electron velocity of the plasma. In addition, the range is sufficiently broad to produce a raised plateau having width .DELTA. in the plasma electron velocity distribution so that the plateau electrons provide steady-state current to generate a poloidal magnetic field component sufficient for confining the plasma. In steady state operation of the fusion reactor, the fusion power density in the plasma exceeds the power dissipated in the plasma.

System and method for generating steady state confining current for a toroidal plasma fusion reactor

DOEpatents

Bers, Abraham

1981-01-01

A system for generating steady state confining current for a toroidal plasma fusion reactor providing steady-state generation of the thermonuclear power. A dense, hot toroidal plasma is initially prepared with a confining magnetic field with toroidal and poloidal components. Continuous wave RF energy is injected into said plasma to estalish a spectrum of traveling waves in the plasma, where the traveling waves have momentum components substantially either all parallel, or all anti-parallel to the confining magnetic field. The injected RF energy is phased to couple to said traveling waves with both a phase velocity component and a wave momentum component in the direction of the plasma traveling wave components. The injected RF energy has a predetermined spectrum selected so that said traveling waves couple to plasma electrons having velocities in a predetermined range .DELTA.. The velocities in the range are substantially greater than the thermal electron velocity of the plasma. In addition, the range is sufficiently broad to produce a raised plateau having width .DELTA. in the plasma electron velocity distribution so that the plateau electrons provide steady-state current to generate a poloidal magnetic field component sufficient for confining the plasma. In steady state operation of the fusion reactor, the fusion power density in the plasma exceeds the power dissipated inthe plasma.

Numerical study of the Columbia high-beta device: Torus-II

DOE Office of Scientific and Technical Information (OSTI.GOV)

Izzo, R.

1981-01-01

The ionization, heating and subsequent long-time-scale behavior of the helium plasma in the Columbia fusion device, Torus-II, is studied. The purpose of this work is to perform numerical simulations while maintaining a high level of interaction with experimentalists. The device is operated as a toroidal z-pinch to prepare the gas for heating. This ionization of helium is studied using a zero-dimensional, two-fluid code. It is essentially an energy balance calculation that follows the development of the various charge states of the helium and any impurities (primarily silicon and oxygen) that are present. The code is an atomic physics model ofmore » Torus-II. In addition to ionization, we include three-body and radiative recombination processes.« less

Research on stellarator-mirror fission-fusion hybrid

NASA Astrophysics Data System (ADS)

Moiseenko, V. E.; Kotenko, V. G.; Chernitskiy, S. V.; Nemov, V. V.; Ågren, O.; Noack, K.; Kalyuzhnyi, V. N.; Hagnestål, A.; Källne, J.; Voitsenya, V. S.; Garkusha, I. E.

2014-09-01

The development of a stellarator-mirror fission-fusion hybrid concept is reviewed. The hybrid comprises of a fusion neutron source and a powerful sub-critical fast fission reactor core. The aim is the transmutation of spent nuclear fuel and safe fission energy production. In its fusion part, neutrons are generated in deuterium-tritium (D-T) plasma, confined magnetically in a stellarator-type system with an embedded magnetic mirror. Based on kinetic calculations, the energy balance for such a system is analyzed. Neutron calculations have been performed with the MCNPX code, and the principal design of the reactor part is developed. Neutron outflux at different outer parts of the reactor is calculated. Numerical simulations have been performed on the structure of a magnetic field in a model of the stellarator-mirror device, and that is achieved by switching off one or two coils of toroidal field in the Uragan-2M torsatron. The calculations predict the existence of closed magnetic surfaces under certain conditions. The confinement of fast particles in such a magnetic trap is analyzed.

Current drive at plasma densities required for thermonuclear reactors.

PubMed

Cesario, R; Amicucci, L; Cardinali, A; Castaldo, C; Marinucci, M; Panaccione, L; Santini, F; Tudisco, O; Apicella, M L; Calabrò, G; Cianfarani, C; Frigione, D; Galli, A; Mazzitelli, G; Mazzotta, C; Pericoli, V; Schettini, G; Tuccillo, A A

2010-08-10

Progress in thermonuclear fusion energy research based on deuterium plasmas magnetically confined in toroidal tokamak devices requires the development of efficient current drive methods. Previous experiments have shown that plasma current can be driven effectively by externally launched radio frequency power coupled to lower hybrid plasma waves. However, at the high plasma densities required for fusion power plants, the coupled radio frequency power does not penetrate into the plasma core, possibly because of strong wave interactions with the plasma edge. Here we show experiments performed on FTU (Frascati Tokamak Upgrade) based on theoretical predictions that nonlinear interactions diminish when the peripheral plasma electron temperature is high, allowing significant wave penetration at high density. The results show that the coupled radio frequency power can penetrate into high-density plasmas due to weaker plasma edge effects, thus extending the effective range of lower hybrid current drive towards the domain relevant for fusion reactors.

MHD simulation of relaxation transition to a flipped relaxed state in spherical torus

NASA Astrophysics Data System (ADS)

Kanki, Takashi; Nagata, Masayoshi; Kagei, Yasuhiro

2008-11-01

Recently, it has been demonstrated in the HIST device that in spite of the violation of the Kruskal-Shafranov stability condition, a normal spherical torus (ST) plasma has relaxed to a flipped ST state through a transient reversed-field pinch-like state when the vacuum toroidal field is decreased and its direction is reversed [1]. It has been also observed during this relaxation transition process that not only the toroidal field but also the poloidal field reverses polarity spontaneously and that the ion flow velocity is strongly fluctuated and abruptly increased up to > 50 km/s. The purpose of the present study is to investigate the plasma flows and the relevant MHD relaxation phenomena to elucidate this transition mechanism by using three-dimensional MHD simulations [2]. It is found from the numerical results that the magnetic reconnection between the open and closed field lines occurs due to the non-linear growth of the n=1 kink instability of the central open flux, generating the toroidal flow ˜ 60 km/s in the direction of the toroidal current. The n=1 kink instability and the plasma flows driven by the magnetic reconnection are consider to be responsible for the self-reversal of the magnetic fields. [1] M. Nagata el al., Phys. Rev. Lett. 90, 225001 (2003). [2] Y. Kagei el al., Plasma. Phys. Control. Fusion 45, L17 (2003).

Resonant magnetic perturbations of edge-plasmas in toroidal confinement devices

DOE Office of Scientific and Technical Information (OSTI.GOV)

Evans, T. E.

Controlling the boundary layer in fusion-grade, high-performance, plasma discharges is essential for the successful development of toroidal magnetic confinement power generating systems. A promising approach for controlling the boundary plasma is based on the use of small, externally applied, edge resonant magnetic perturbation (RMP) fields (δmore » $$b_⊥^{ext}$$ ≈ $$10^{-4}$$ → $$10^{-3}$$ T). A long-term focus area in tokamak fusion research has been to find methods, involving the use of non-axisymmetric magnetic perturbations to reduce the intense particle and heat fluxes to the wall. Experimental RMP research has progressed from the early pioneering work on tokamaks with material limiters in the 1970s, to present day research in separatrix-limited tokamaks operated in high-confinement mode, which is primarily aimed at the mitigation of the intermittent fluxes due edge localized modes. At the same time the theoretical research has evolved from analytical models to numerical simulations, including the full 3D complexities of the problem. Following the first demonstration of ELM suppression in the DIII-D tokamak during 2003, there has been a rapid worldwide growth in theoretical, numerical and experimental edge RMP research resulting in the addition of ELM control coils to the ITER baseline design [A. Loarte, et al., Nucl. Fusion 54 (2014) 033007]. This review provides an overview of edge RMP research including a summary of the early theoretical and numerical background along with recent experimental results on improved particle and energy confinement in tokamaks triggered by edge RMP fields. The topics covered make up the basic elements needed for developing a better understanding of 3D magnetic perturbation physics, which is required in order to utilize the full potential of edge RMP fields in fusion relevant high performance, H-mode, plasmas.« less

Resonant magnetic perturbations of edge-plasmas in toroidal confinement devices

DOE PAGES

Evans, T. E.

2015-11-13

Controlling the boundary layer in fusion-grade, high-performance, plasma discharges is essential for the successful development of toroidal magnetic confinement power generating systems. A promising approach for controlling the boundary plasma is based on the use of small, externally applied, edge resonant magnetic perturbation (RMP) fields (δmore » $$b_⊥^{ext}$$ ≈ $$10^{-4}$$ → $$10^{-3}$$ T). A long-term focus area in tokamak fusion research has been to find methods, involving the use of non-axisymmetric magnetic perturbations to reduce the intense particle and heat fluxes to the wall. Experimental RMP research has progressed from the early pioneering work on tokamaks with material limiters in the 1970s, to present day research in separatrix-limited tokamaks operated in high-confinement mode, which is primarily aimed at the mitigation of the intermittent fluxes due edge localized modes. At the same time the theoretical research has evolved from analytical models to numerical simulations, including the full 3D complexities of the problem. Following the first demonstration of ELM suppression in the DIII-D tokamak during 2003, there has been a rapid worldwide growth in theoretical, numerical and experimental edge RMP research resulting in the addition of ELM control coils to the ITER baseline design [A. Loarte, et al., Nucl. Fusion 54 (2014) 033007]. This review provides an overview of edge RMP research including a summary of the early theoretical and numerical background along with recent experimental results on improved particle and energy confinement in tokamaks triggered by edge RMP fields. The topics covered make up the basic elements needed for developing a better understanding of 3D magnetic perturbation physics, which is required in order to utilize the full potential of edge RMP fields in fusion relevant high performance, H-mode, plasmas.« less

Theoretical studies of possible toroidal high-spin isomers in the light-mass region

DOE PAGES

Staszczak, A.; Wong, Cheuk-Yin

2016-05-11

We review our theoretical knowledge of possible toroidal high-spin isomers in the light mass region in 28≤A≤52 obtained previously in cranked Skyrme-Hartree-Fock calculations. We report additional toroidal high-spin isomers in 56Ni with I=114ℏ and 140ℏ, which follow the same (multi-particle) (multi-hole) systematics as other toroidal high-spin isomers. We examine the production of these exotic nuclei by fusion of various projectiles on 20Ne or 28Si as an active target in time-projection-chamber (TPC) experiments.

Turbulent Stresses in LAPD and CSDX

NASA Astrophysics Data System (ADS)

Light, A. D.; Sechrest, Y.; Schaffner, D. A.; Muller, S. H.; Rossi, G. D.; Guice, D.; Carter, T. A.; Tynan, G. R.; Vincena, S.; Munsat, T.

2011-10-01

Turbulent momentum transport can affect phenomena as diverse as intrinsic rotation in self-organized systems, stellar dynamo, astrophysical accretion, and the mechanism of internal transport barriers in fusion devices. Contributions from turbulent fluctuations, in the form of Reynolds and Maxwell stress terms, have been predicted theoretically and observed in toroidal devices. In an effort to gain general insight into the physics, we present new results from turbulent stress measurements on two linear devices: the LArge Plasma Device (LAPD) at the University of California, Los Angeles, and the Controlled Shear De-correlation eXperiment (CSDX) at the University of California, San Diego. Both experiments are well-characterized linear machines in which the plasma beta can be varied. Electrostatic and magnetic fluctuations are measured over a range of plasma parameters in concert with fast imaging. Maxwell and Reynolds stresses are calculated from probe data and fluctuations are compared with fast camera images using velocimetry techniques.

One-dimensional MHD simulations of MTF systems with compact toroid targets and spherical liners

NASA Astrophysics Data System (ADS)

Khalzov, Ivan; Zindler, Ryan; Barsky, Sandra; Delage, Michael; Laberge, Michel

2017-10-01

One-dimensional (1D) MHD code is developed in General Fusion (GF) for coupled plasma-liner simulations in magnetized target fusion (MTF) systems. The main goal of these simulations is to search for optimal parameters of MTF reactor, in which spherical liquid metal liner compresses compact toroid plasma. The code uses Lagrangian description for both liner and plasma. The liner is represented as a set of spherical shells with fixed masses while plasma is discretized as a set of nested tori with circular cross sections and fixed number of particles between them. All physical fields are 1D functions of either spherical (liner) or small toroidal (plasma) radius. Motion of liner and plasma shells is calculated self-consistently based on applied forces and equations of state. Magnetic field is determined by 1D profiles of poloidal and toroidal fluxes - they are advected with shells and diffuse according to local resistivity, this also accounts for flux leakage into the liner. Different plasma transport models are implemented, this allows for comparison with ongoing GF experiments. Fusion power calculation is included into the code. We performed a series of parameter scans in order to establish the underlying dependencies of the MTF system and find the optimal reactor design point.

Public Data Set: H-mode Plasmas at Very Low Aspect Ratio on the Pegasus Toroidal Experiment

DOE Data Explorer

Thome, Kathreen E. [University of Wisconsin-Madison; Oak Ridge Associated Universities] (ORCID:0000000248013922); Bongard, Michael W. [University of Wisconsin-Madison] (ORCID:0000000231609746); Barr, Jayson L. [University of Wisconsin-Madison] (ORCID:0000000177685931); Bodner, Grant M. [University of Wisconsin-Madison] (ORCID:0000000324979172); Burke, Marcus G. [University of Wisconsin-Madison] (ORCID:0000000176193724); Fonck, Raymond J. [University of Wisconsin-Madison] (ORCID:0000000294386762); Kriete, David M. [University of Wisconsin-Madison] (ORCID:0000000236572911); Perry, Justin M. [University of Wisconsin-Madison] (ORCID:0000000171228609); Reusch, Joshua A. [University of Wisconsin-Madison] (ORCID:0000000284249422); Schlossberg, David J. [University of Wisconsin-Madison] (ORCID:0000000287139448)

2016-09-30

This data set contains openly-documented, machine readable digital research data corresponding to figures published in K.E. Thome et al., 'H-mode Plasmas at Very Low Aspect Ratio on the Pegasus Toroidal Experiment,' Nucl. Fusion 57, 022018 (2017).

Addressing Research and Development Gaps for Plasma-Material Interactions with Linear Plasma Devices

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rapp, Juergen

Plasma-material interactions in future fusion reactors have been identified as a knowledge gap to be dealt with before any next step device past ITER can be built. The challenges are manifold. They are related to power dissipation so that the heat fluxes to the plasma-facing components can be kept at technologically feasible levels; maximization of the lifetime of divertor plasma-facing components that allow for steadystate operation in a reactor to reach the neutron fluence required; the tritium inventory (storage) in the plasma-facing components, which can lead to potential safety concerns and reduction in the fuel efficiency; and it is relatedmore » to the technology of the plasma-facing components itself, which should demonstrate structural integrity under the high temperatures and high neutron fluence. While the dissipation of power exhaust can and should be addressed in high power toroidal devices, the interaction of the plasma with the materials can be best addressed in dedicated linear devices due to their cost effectiveness and ability to address urgent research and development gaps more timely. However, new linear plasma devices are needed to investigate the PMI under fusion reactor conditions and test novel plasma-facing components. Existing linear devices are limited either in their flux, their reactor-relevant plasma transport regimes in front of the target, their fluence, or their ability to test material samples a priori exposed to high neutron fluence. The proposed Material Plasma Exposure eXperiment (MPEX) is meant to address those deficiencies and will be designed to fulfill the fusion reactor-relevant plasma parameters as well as the ability to expose a priori neutron activated materials to plasmas.« less

Erratum: Resonant magnetic perturbations of edge-plasmas in toroidal confinement devices (2015 Plasma Phys. Control. Fusion 57 123001)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Evans, T. E.

Controlling the boundary layer in fusion-grade, high-performance, plasma discharges is essential for the successful development of toroidal magnetic confinement power generating systems. A promising approach for controlling the boundary plasma is based on the use of small, externally applied, edge resonant magnetic perturbation (RMP) fields (δmore » $$b_⊥^{ext}$$ ≈ $$10^{-4}$$ → $$10^{-3}$$ T). A long-term focus area in tokamak fusion research has been to find methods, involving the use of non-axisymmetric magnetic perturbations to reduce the intense particle and heat fluxes to the wall. Experimental RMP research has progressed from the early pioneering work on tokamaks with material limiters in the 1970s, to present day research in separatrix-limited tokamaks operated in high-confinement mode, which is primarily aimed at the mitigation of the intermittent fluxes due edge localized modes. At the same time the theoretical research has evolved from analytical models to numerical simulations, including the full 3D complexities of the problem. Following the first demonstration of ELM suppression in the DIII-D tokamak during 2003, there has been a rapid worldwide growth in theoretical, numerical and experimental edge RMP research resulting in the addition of ELM control coils to the ITER baseline design [A. Loarte, et al., Nucl. Fusion 54 (2014) 033007]. This review provides an overview of edge RMP research including a summary of the early theoretical and numerical background along with recent experimental results on improved particle and energy confinement in tokamaks triggered by edge RMP fields. The topics covered make up the basic elements needed for developing a better understanding of 3D magnetic perturbation physics, which is required in order to utilize the full potential of edge RMP fields in fusion relevant high performance, H-mode, plasmas.« less

Erratum: Resonant magnetic perturbations of edge-plasmas in toroidal confinement devices (2015 Plasma Phys. Control. Fusion 57 123001)

DOE PAGES

Evans, T. E.

2016-03-01

Controlling the boundary layer in fusion-grade, high-performance, plasma discharges is essential for the successful development of toroidal magnetic confinement power generating systems. A promising approach for controlling the boundary plasma is based on the use of small, externally applied, edge resonant magnetic perturbation (RMP) fields (δmore » $$b_⊥^{ext}$$ ≈ $$10^{-4}$$ → $$10^{-3}$$ T). A long-term focus area in tokamak fusion research has been to find methods, involving the use of non-axisymmetric magnetic perturbations to reduce the intense particle and heat fluxes to the wall. Experimental RMP research has progressed from the early pioneering work on tokamaks with material limiters in the 1970s, to present day research in separatrix-limited tokamaks operated in high-confinement mode, which is primarily aimed at the mitigation of the intermittent fluxes due edge localized modes. At the same time the theoretical research has evolved from analytical models to numerical simulations, including the full 3D complexities of the problem. Following the first demonstration of ELM suppression in the DIII-D tokamak during 2003, there has been a rapid worldwide growth in theoretical, numerical and experimental edge RMP research resulting in the addition of ELM control coils to the ITER baseline design [A. Loarte, et al., Nucl. Fusion 54 (2014) 033007]. This review provides an overview of edge RMP research including a summary of the early theoretical and numerical background along with recent experimental results on improved particle and energy confinement in tokamaks triggered by edge RMP fields. The topics covered make up the basic elements needed for developing a better understanding of 3D magnetic perturbation physics, which is required in order to utilize the full potential of edge RMP fields in fusion relevant high performance, H-mode, plasmas.« less

PROGRESS IN THE PEELING-BALLOONING MODEL OF ELMS: TOROIDAL ROTATION AND 3D NONLINEAR DYNAMICS

DOE Office of Scientific and Technical Information (OSTI.GOV)

SNYDER,P.B; WILSON,H.R; XU,X.Q

2004-06-01

Understanding the physics of the H-Mode pedestal and edge localized modes (ELMs) is very important to next-step fusion devices for two primary reasons: (1) The pressure at the top of the edge barrier (''pedestal height'') strongly impacts global confinement and fusion performance, and (2) large ELMs lead to localized transient heat loads on material surfaces that may constrain component lifetimes. The development of the peeling-ballooning model has shed light on these issues by positing a mechanism for ELM onset and constraints on the pedestal height. The mechanism involves instability of ideal coupled ''peeling-ballooning'' modes driven by the sharp pressure gradientmore » and consequent large bootstrap current in the H-mode edge. It was first investigated in the local, high-n limit [1], and later quantified for non-local, finite-n modes in general toroidal geometry [2,3]. Important aspects are that a range of wavelengths may potentially be unstable, with intermediate n's (n {approx} 3-30) generally limiting in high performance regimes, and that stability bounds are strongly sensitive to shape [Fig l(a)], and to collisionality (i.e. temperature and density) [4] through the bootstrap current. The development of efficient MHD stability codes such as ELITE [3,2] and MISHKA [5] has allowed detailed quantification of peeling-ballooning stability bounds (e.g. [6]) and extensive and largely successful comparisons with observation (e.g. [2,6-9]). These previous calculations are ideal, static, and linear. Here we extend this work to incorporate the impact of sheared toroidal rotation, and the non-ideal, nonlinear dynamics which must be studied to quantify ELM size and heat deposition on material surfaces.« less

System and method of operating toroidal magnetic confinement devices

DOEpatents

Chance, M.S.; Jardin, S.C.; Stix, T.H.; Grimm, R.C.; Manickam, J.; Okabayashi, M.

1984-08-30

This invention pertains to methods and arrangements for attaining high beta values in plasma confinement devices. More specifically, this invention pertains to methods for accessing the second stability region of operation in toroidal magnetic confinement devices.

Toroidal band limiter for a plasma containment device

DOEpatents

Kelley, George G.

1978-01-01

This invention relates to a toroidal plasma confinement device having poloidal and toroidal magnetic fields for confining a toroidal plasma column with a plasma current induced therein along an endless, circular equilibrium axis in a torus vacuum cavity wherein the improvement comprises the use of a toroidal plasma band limiter mounted within the vacuum cavity in such a manner as to ensure that the plasma energy is distributed more uniformly over the limiter surface thereby avoiding intense local heating of the limiter while at the same time substantially preventing damage to the plasma containment wall of the cavity by the energetic particles diffusing out from the confined plasma. A plurality of poloidal plasma ring limiters are also utilized for containment wall protection during any disruptive instability that might occur during operation of the device.

System and method for generating current by selective electron heating

DOEpatents

Fisch, Nathaniel J.; Boozer, Allen H.

1984-01-01

A system for the generation of toroidal current in a plasma which is prepared in a toroidal magnetic field. The system utilizes the injection of high-frequency waves into the plasma by means of waveguides. The wave frequency and polarization are chosen such that when the waveguides are tilted in a predetermined fashion, the wave energy is absorbed preferentially by electrons traveling in one toroidal direction. The absorption of energy in this manner produces a toroidal electric current even when the injected waves themselves do not have substantial toroidal momentum. This current can be continuously maintained at modest cost in power and may be used to confine the plasma. The system can operate efficiently on fusion grade tokamak plasmas.

Soft X-ray studies on MST: Measuring the effects of toroidicity on tearing mode phase and installation of a multi-energy camera

NASA Astrophysics Data System (ADS)

Vanmeter, Patrick; Reusch, Lisa; Franz, Paolo; Sarff, John; Goetz, John; Delgado-Aparicio, Louis; den Hartog, Daniel

2017-10-01

The soft X-ray tomography (SXT) system on MST uses four cameras in a double-filter configuration to measure the emitted brightness along forty distinct lines of sight. These measurements can then be inverted to determine the emissivity, which depends on physical properties such as temperature, density, and impurity content. The SXR emissivity should correspond to the structure of the magnetic field; however, there is a discrepancy between the phase of the emissivity inversions and magnetic field reconstructions when using the typical cylindrical approximation to interpret the signal from the toroidal magnetics array. This discrepancy was measured for two distinct plasma conditions using all four SXT cameras, with results supporting the interpretation that it emerges from physical effects of the toroidal geometry. In addition, a new soft x-ray measurement system based on the PILATUS3 photon counting detector will be installed on MST. Emitted photons are counted by an array of pixels with individually adjustable energy cutoffs giving the device more spectral information than the double-filter system. This material is based upon work supported by the U.S. Department of Energy, Office of Science, Office of Fusion Energy Sciences program under Award Numbers DE-FC02-05ER54814 and DE-SC0015474.

Thermonuclear instabilities and plasma edge transport in tokamaks

NASA Astrophysics Data System (ADS)

Fulop, Tunde Maria

High-energy ions generated by fusion reactions in a burning fusion plasma may give rise to different types of wave instabilities. The present thesis investigates two types of such instabilities which recently have been observed in fusion experiments: the Toroidal Alfvén Eigenmode (TAE) instability and the magnetoacoustic cyclotron instability (MCI) which is predicted to give rise to ion cyclotron emission (ICE). The TAE instability may degrade the confinement of fusion-produced high energy alpha particles and adversely affect the possibilities of reaching ignition. The present work derives it generalized expression for the linear growth rate of the instability, by including the effects of finite orbit width and finite Larmor radius of energetic particles, as well as the effects of mode localization and the possible mode excitation by both passing and trapped energetic ions. ICE does not threaten the plasma performance, but it might be useful as a fast ion diagnostic. The ICE originates from the MCI involving fast magnetoacoustic waves driven unstable by toroidicity-affected cyclotron resonance with fast ions. In the present thesis a detailed numerical and analytical investigation of this instability is presented, that explains most of the experimental ICE features observed in JET and TFTR. Moreover, the radial and poloidal localization of the fast magnetoacoustic eigenmodes is investigated, including the effects of toroidicity, ellipticity, the presence of a subpopulation of high energy ions and various profiles of the bulk ion density. In a fusion reactor, the transport of the particles near the edge have a strong influence on the global confinement of the plasma. In the edge region, where neutral atoms and impurity ions are abundant and the temperature and density gradients are large, the assumptions of the standard neoclassical theory break down. In this thesis, we explore the effect of neutral particles on the ion flow shear in the edge region. Furthermore, the neoclassical transport theory in an impure, toroidally rotating plasma is extended to allow for steeper pressure and temperature gradients than are usually considered.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Morgan, O.B. Jr.; Berry, L.A.; Sheffield, J.

This annual report on fusion energy discusses the progress on work in the following main topics: toroidal confinement experiments; atomic physics and plasma diagnostics development; plasma theory and computing; plasma-materials interactions; plasma technology; superconducting magnet development; fusion engineering design center; materials research and development; and neutron transport. (LSP)

Theoretical study on the laser-driven ion-beam trace probe in toroidal devices with large poloidal magnetic field

NASA Astrophysics Data System (ADS)

Yang, X.; Xiao, C.; Chen, Y.; Xu, T.; Yu, Y.; Xu, M.; Wang, L.; Wang, X.; Lin, C.

2018-03-01

Recently, a new diagnostic method, Laser-driven Ion-beam Trace Probe (LITP), has been proposed to reconstruct 2D profiles of the poloidal magnetic field (Bp) and radial electric field (Er) in the tokamak devices. A linear assumption and test particle model were used in those reconstructions. In some toroidal devices such as the spherical tokamak and the Reversal Field Pinch (RFP), Bp is not small enough to meet the linear assumption. In those cases, the error of reconstruction increases quickly when Bp is larger than 10% of the toroidal magnetic field (Bt), and the previous test particle model may cause large error in the tomography process. Here a nonlinear reconstruction method is proposed for those cases. Preliminary numerical results show that LITP could be applied not only in tokamak devices, but also in other toroidal devices, such as the spherical tokamak, RFP, etc.

Long-wavelength microinstabilities in toroidal plasmas*

NASA Astrophysics Data System (ADS)

Tang, W. M.; Rewoldt, G.

1993-07-01

Realistic kinetic toroidal eigenmode calculations have been carried out to support a proper assessment of the influence of long-wavelength microturbulence on transport in tokamak plasmas. In order to efficiently evaluate large-scale kinetic behavior extending over many rational surfaces, significant improvements have been made to a toroidal finite element code used to analyze the fully two-dimensional (r,θ) mode structures of trapped-ion and toroidal ion temperature gradient (ITG) instabilities. It is found that even at very long wavelengths, these eigenmodes exhibit a strong ballooning character with the associated radial structure relatively insensitive to ion Landau damping at the rational surfaces. In contrast to the long-accepted picture that the radial extent of trapped-ion instabilities is characterized by the ion-gyroradius-scale associated with strong localization between adjacent rational surfaces, present results demonstrate that under realistic conditions, the actual scale is governed by the large-scale variations in the equilibrium gradients. Applications to recent measurements of fluctuation properties in Tokamak Fusion Test Reactor (TFTR) [Plasma Phys. Controlled Nucl. Fusion Res. (International Atomic Energy Agency, Vienna, 1985), Vol. 1, p. 29] L-mode plasmas indicate that the theoretical trends appear consistent with spectral characteristics as well as rough heuristic estimates of the transport level. Benchmarking calculations in support of the development of a three-dimensional toroidal gyrokinetic code indicate reasonable agreement with respect to both the properties of the eigenfunctions and the magnitude of the eigenvalues during the linear phase of the simulations of toroidal ITG instabilities.

Three-dimensional drift kinetic response of high-β plasmas in the DIII-D tokamak.

PubMed

Wang, Z R; Lanctot, M J; Liu, Y Q; Park, J-K; Menard, J E

2015-04-10

A quantitative interpretation of the experimentally measured high-pressure plasma response to externally applied three-dimensional (3D) magnetic field perturbations, across the no-wall Troyon β limit, is achieved. The self-consistent inclusion of the drift kinetic effects in magnetohydrodynamic (MHD) modeling [Y. Q. Liu et al., Phys. Plasmas 15, 112503 (2008)] successfully resolves an outstanding issue of the ideal MHD model, which significantly overpredicts the plasma-induced field amplification near the no-wall limit, as compared to experiments. The model leads to quantitative agreement not only for the measured field amplitude and toroidal phase but also for the measured internal 3D displacement of the plasma. The results can be important to the prediction of the reliable plasma behavior in advanced fusion devices, such as ITER [K. Ikeda, Nucl. Fusion 47, S1 (2007)].

Latest results and developments from the Hybrid Illinois Device for Research and Applications

NASA Astrophysics Data System (ADS)

Rizkallah, Rabel; Andruczyk, Daniel; Jeckell, Zachary Jon; Shone, Andrew John; Johnson, Daniel Scott; Allain, Jean Paul; Curreli, Davide; Ruzic, David N.; The Hidra Team

2017-10-01

The Hybrid Illinois Device for Research and Applications (HIDRA) is a five-period, l = 2, m = 5, toroidal fusion device operated at the University of Illinois at Urbana-Champaign (UIUC). It has a major radius R0 = 0.72 m and minor radius a = 0.19 m. Initial heating is achieved with 2.45 GHz electron cyclotron resonance heating (ECRH) at an on-axis magnetic field of B0 = 0.087 T which can go as high as B0 = 0.5 T. HIDRA will mainly be used as a classical stellarator, but can also run as a tokamak. This allows for both steady-state and transient regime operations. Experiments on HIDRA will primarily tackle the issue of plasma-material interactions (PMI) in fusion, and focus on developing innovative plasma facing component (PFC) technologies. Currently, research on flowing liquid lithium PFCs meant to be tested inside the machine in real-time operation, is being carried on. The first experiments run on HIDRA started in early 2016 in the low field region. Now, HIDRA is also capable of running in the high field zone, allowing for more interesting experiments and meaningful outcomes. Here, we present some of the initial results coming from the machine.

DOE Office of Scientific and Technical Information (OSTI.GOV)

None

The vision described here builds on the present U.S. activities in fusion plasma and materials science relevant to the energy goal and extends plasma science at the frontier of discovery. The plan is founded on recommendations made by the National Academies, a number of recent studies by the Fusion Energy Sciences Advisory Committee (FESAC), and the Administration’s views on the greatest opportunities for U.S. scientific leadership.This report highlights five areas of critical importance for the U.S. fusion energy sciences enterprise over the next decade: 1) Massively parallel computing with the goal of validated whole-fusion-device modeling will enable a transformation inmore » predictive power, which is required to minimize risk in future fusion energy development steps; 2) Materials science as it relates to plasma and fusion sciences will provide the scientific foundations for greatly improved plasma confinement and heat exhaust; 3) Research in the prediction and control of transient events that can be deleterious to toroidal fusion plasma confinement will provide greater confidence in machine designs and operation with stable plasmas; 4) Continued stewardship of discovery in plasma science that is not expressly driven by the energy goal will address frontier science issues underpinning great mysteries of the visible universe and help attract and retain a new generation of plasma/fusion science leaders; 5) FES user facilities will be kept world-leading through robust operations support and regular upgrades. Finally, we will continue leveraging resources among agencies and institutions and strengthening our partnerships with international research facilities.« less

High-Energy Space Propulsion Based on Magnetized Target Fusion

NASA Technical Reports Server (NTRS)

Thio, Y. C. F.; Landrum, D. B.; Freeze, B.; Kirkpatrick, R. C.; Gerrish, H.; Schmidt, G. R.

1999-01-01

Magnetized target fusion is an approach in which a magnetized target plasma is compressed inertially by an imploding material wall. A high energy plasma liner may be used to produce the required implosion. The plasma liner is formed by the merging of a number of high momentum plasma jets converging towards the center of a sphere where two compact toroids have been introduced. Preliminary 3-D hydrodynamics modeling results using the SPHINX code of Los Alamos National Laboratory have been very encouraging and confirm earlier theoretical expectations. The concept appears ready for experimental exploration and plans for doing so are being pursued. In this talk, we explore conceptually how this innovative fusion approach could be packaged for space propulsion for interplanetary travel. We discuss the generally generic components of a baseline propulsion concept including the fusion engine, high velocity plasma accelerators, generators of compact toroids using conical theta pinches, magnetic nozzle, neutron absorption blanket, tritium reprocessing system, shock absorber, magnetohydrodynamic generator, capacitor pulsed power system, thermal management system, and micrometeorite shields.

The Helicity Injected Torus (HIT) Program

NASA Astrophysics Data System (ADS)

Jarboe, T. R.; Gu, P.; Hamp, W.; Izzo, V.; Jewell, P.; Liptac, J.; McCollam, K. J.; Nelson, B. A.; Raman, R.; Redd, A. J.; Shumlak, U.; Sieck, P. E.; Smith, R. J.; Jain, K. K.; Nagata, M.; Uyama, T.

2000-10-01

The purpose of the Helicity Injected Torus (HIT) program is to develop current drive techniques for low-aspect-ratio toroidal plasmas. The present HIT-II spherical tokamak experiment is capable of both Coaxial Helicity Injection (CHI) and transformer action current drive. The HIT-II device itself is modestly sized (major radius R = 0.3 m, minor radius a = 0.2 m, with an on-axis magnetic field of up to Bo = 0.5 T), but has demonstrated toroidal plasma currents of up to 200 kA, using either CHI or transformer drive. An overview of ongoing research on HIT-II plasmas, including recent results, will be presented. An electron-locking model has been developed for helicity injection current drive; a description of this model will be presented, as well as comparisons to experimental results from the HIT and HIT-II devices. Empirical results from both the HIT program and past spheromak research, buttressed by theoretical developments, have led to the design of the upcoming HIT-SI (Helicity Injected Torus with Steady Inductive helicity injection) device (T.R. Jarboe, Fusion Technology 36, p. 85, 1999). HIT-SI will be able to form a high-beta spheromak, a low aspect ratio RFP or a spherical tokamak using constant inductive helicity injection. The HIT-SI design and construction progress will be presented.

Fully-kinetic Ion Simulation of Global Electrostatic Turbulent Transport in C-2U

NASA Astrophysics Data System (ADS)

Fulton, Daniel; Lau, Calvin; Bao, Jian; Lin, Zhihong; Tajima, Toshiki; TAE Team

2017-10-01

Understanding the nature of particle and energy transport in field-reversed configuration (FRC) plasmas is a crucial step towards an FRC-based fusion reactor. The C-2U device at Tri Alpha Energy (TAE) achieved macroscopically stable plasmas and electron energy confinement time which scaled favorably with electron temperature. This success led to experimental and theoretical investigation of turbulence in C-2U, including gyrokinetic ion simulations with the Gyrokinetic Toroidal Code (GTC). A primary objective of TAE's new C-2W device is to explore transport scaling in an extended parameter regime. In concert with the C-2W experimental campaign, numerical efforts have also been extended in A New Code (ANC) to use fully-kinetic (FK) ions and a Vlasov-Poisson field solver. Global FK ion simulations are presented. Future code development is also discussed.

System and method for generating current by selective minority species heating

DOEpatents

Fisch, Nathaniel J.

1983-01-01

A system for the generation of toroidal current in a plasma which is prepared in a toroidal magnetic field. The system utilizes the injection of low-frequency waves into the plasma by means of phased antenna arrays or phased waveguide arrays. The plasma is prepared with a minority ion species of different charge state and different gyrofrequency from the majority ion species. The wave frequency and wave phasing are chosen such that the wave energy is absorbed preferentially by minority species ions traveling in one toroidal direction. The absorption of energy in this manner produces a toroidal electric current even when the injected waves themselves do not have substantial toroidal momentum. This current can be continuously maintained at modest cost in power and may be used to confine the plasma. The system can operate efficiently on fusion grade tokamak plasmas.

Benefits and drawbacks of low magnetic shears on the confinement in magnetic fusion toroidal devices

NASA Astrophysics Data System (ADS)

Firpo, Marie-Christine; Constantinescu, Dana

2012-10-01

The issue of confinement in magnetic fusion devices is addressed within a purely magnetic approach. As it is well known, the magnetic field being divergence-free, the equations of its field lines can be cast in Hamiltonian form. Using then some Hamiltonian models for the magnetic field lines, the dual impact of low magnetic shear is demonstrated. Away from resonances, it induces a drastic enhancement of magnetic confinement that favors robust internal transport barriers (ITBs) and turbulence reduction. However, when low-shear occurs for values of the winding of the magnetic field lines close to low-order rationals, the amplitude thresholds of the resonant modes that break internal transport barriers by allowing a radial stochastic transport of the magnetic field lines may be much lower than the ones obtained for strong shear profiles. The approach can be applied to assess the robustness versus magnetic perturbations of general almost-integrable magnetic steady states, including non-axisymmetric ones such as the important single helicity steady states. This analysis puts a constraint on the tolerable mode amplitudes compatible with ITBs and may be proposed as a possible explanation of diverse experimental and numerical signatures of their collapses.

Toroidal Alfven Waves in Advanced Tokamaks

NASA Astrophysics Data System (ADS)

Berk, Herbert L.

2003-10-01

In burning plasma experiments, alpha particles have speeds that readily resonate with shear Alfven waves. It is essential to understand this Alfven wave spectrum for toroidal plasma confinement. Most interest has focused on the Toroidal Alfven Eigenmode (TAE), and a method of analysis has been developed to understand the structure of this mode at a flux surface with a given magnetic shear. However, this model fails when the shear is too low or reversed. In this case a new method of analysis is required, which must incorporate novel fluid-like effects from the energetic particles [1] and also include effects that are second order in the inverse toroidal aspect ratio. With this new method [2] we can obtain spectral features that agree with experimental results. In particular, this theory gives an explanation for the so-called Cascade modes that have been observed in JT-60 [3], JET [4], and TFTR [5]. For these Cascade modes, slow upward frequency sweeping is observed, beginning from frequencies below the TAE range but then often blending into the TAE range of frequencies. The theoretical understanding of the Cascades modes has evolved to the point where these modes can be used as a diagnostic "signature" [6] to experimentally optimize the formation of thermal barriers in reversed-shear operation when the minimum q value is an integer. [1] H. L. Berk et al., Phys. Rev. Lett. 87, 185 (2002). [2] B. N. Breizman et al., submitted to Phys. Plasmas (2003). [3] H. Kimura et al., Nucl. Fusion 38, 1303 (1998). [4] S. Sharapov et al., Phys. Lett. A 289, 127 (2001); S. Sharapov, Phys. Plasmas 9, 2027 (2002). [5] R. Nazikian, H. L. Berk, et al., Bull. Am. Phys. Soc. 47, 327 (2002). [6] E. Joffrin et al., Plasma Phys. Contr. Fusion 44, 1739 (2002); E. Joffrin et al., in Proc. 2002 IAEA Fusion Energy Conference, submitted to Nucl. Fusion.

Magnetic fusion commercial power plants

NASA Astrophysics Data System (ADS)

Sheffield, J.

Toroidal magnetic systems present the best opportunity to make a commercial fusion power plant. They offer potential solutions to the main requirements that confront a power plant designer. An ideal system may be postulated in which the coils are a very small part of the cost, and the cost stems primarily from the inescapable components: minimal plasma heating (and sustaining system), tritium breeding blanket, shield, particle input, removal and treatment system, heat transfer system, generators, buildings, and balance of plant. No present system meets the ideal standards; however, toroidal systems contain among them the elements required. Consequently, a logical program may be based upon an evolutionary development, building on the contributions of the tokamak, which has been the mainline of research for a number of years.

Advances in boronization on NSTX-Upgrade

DOE PAGES

Skinner, C. H.; Bedoya, F.; Scotti, F.; ...

2017-01-27

Boronization has been effective in reducing plasma impurities and enabling access to higher density, higher confinement plasmas in many magnetic fusion devices. The National Spherical Torus eXperiment, NSTX, has recently undergone a major upgrade to NSTX-U in order to develop the physics basis for a ST-based Fusion Nuclear Science Facility (FNSF) with capability for double the toroidal field, plasma current, and NBI heating power and increased pulse duration from 1–1.5 s to 5–8 s. A new deuterated tri-methyl boron conditioning system was implemented together with a novel surface analysis diagnostic. We report on the spatial distribution of the boron depositionmore » versus discharge pressure, gas injection and electrode location. The oxygen concentration of the plasma facing surface was measured by in-vacuo XPS and increased both with plasma exposure and with exposure to trace residual gases. Furthermore, this increase correlated with the rise of oxygen emission from the plasma.« less

Verification of gyrokinetic particle simulation of current-driven instability in fusion plasmas. I. Internal kink mode

DOE Office of Scientific and Technical Information (OSTI.GOV)

McClenaghan, J.; Lin, Z.; Holod, I.

The gyrokinetic toroidal code (GTC) capability has been extended for simulating internal kink instability with kinetic effects in toroidal geometry. The global simulation domain covers the magnetic axis, which is necessary for simulating current-driven instabilities. GTC simulation in the fluid limit of the kink modes in cylindrical geometry is verified by benchmarking with a magnetohydrodynamic eigenvalue code. Gyrokinetic simulations of the kink modes in the toroidal geometry find that ion kinetic effects significantly reduce the growth rate even when the banana orbit width is much smaller than the radial width of the perturbed current layer at the mode rational surface.

The role of fluctuation-induced transport in a toroidal plasma with strong radial electric fields

NASA Technical Reports Server (NTRS)

Roth, J. R.; Krawczonek, W. M.; Powers, E. J.; Hong, J. Y.; Kim, Y. C.

1981-01-01

Previous work employing digitally implemented spectral analysis techniques is extended to demonstrate that radial fluctuation-induced transport is the dominant ion transport mechanism in an electric field dominated toroidal plasma. Such transport can be made to occur against a density gradient, and hence may have a very beneficial effect on confinement in toroidal plasmas of fusion interest. It is shown that Bohm or classical diffusion down a density gradient, the collisional Pedersen-current mechanism, and the collisionless electric field gradient mechanism described by Cole (1976) all played a minor role, if any, in the radial transport of this plasma.

Plasma Studies in the SPECTOR Experiment as Target Development for MTF

NASA Astrophysics Data System (ADS)

Ivanov, Russ; Young, William; the Fusion Team, General

2016-10-01

General Fusion (GF) is developing a Magnetized Target Fusion (MTF) concept in which magnetized plasmas are adiabatically compressed to fusion conditions by the collapse of a liquid metal vortex. To study and optimize the plasma compression process, GF has a field test program in which subscale plasma targets are rapidly compressed with a moving flux conserver. GF has done many field tests to date on plasmas with sufficient thermal confinement but with a compression geometry that is not nearly self-similar. GF has a new design for our subscale plasma injectors called SPECTOR (for SPhErical Compact TORoid) capable of generating and compressing plasmas with a more spherical form factor. SPECTOR forms spherical tokamak plasmas by coaxial helicity injection into a flux conserver (a = 9 cm, R = 19 cm) with a pre-existing toroidal field created by 0.5 MA current in an axial shaft. The toroidal plasma current of 100 - 300 kA resistively decays over a time period of 1.5 msec. SPECTOR1 has an extensive set of plasma diagnostics including Thomson scattering and polarimetry. MHD stability and lifetime of the plasma was explored in different magnetic configurations with a variable safety factor q(Ψ) . Relatively hot (Te >= 350 eV) and dense ( 1020 m-3) plasmas have achieved energy confinement times τE >= 100 μsec and are now ready for field compression tests. russ.ivanov@generalfusion.com.

Techniques for the measurement of disruption halo currents in the National Spherical Torus Experiment.

PubMed

Gerhardt, S P; Fredrickson, E; Guttadora, L; Kaita, R; Kugel, H; Menard, J; Takahashi, H

2011-10-01

This paper describes techniques for measuring halo currents, and their associated toroidal peaking, in the National Spherical Torus Experiments [M. Ono et al., Nucl. Fusion 40, 557 (2000)]. The measurements are based on three techniques: (1) measurement of the toroidal field created by the poloidal halo current, either with segmented Rogowski coils or discrete toroidal field sensors, (2) the direct measurement of halo currents into specially instrument tiles, and (3) small Rogowski coils placed on the mechanical supports of in-vessel components. For the segmented Rogowski coils and discrete toroidal field detectors, it is shown that the toroidal peaking factor inferred from the data is significantly less than the peaking factor of the underlying halo current distribution, and a simple model is developed to relate the two. For the array of discrete toroidal field detectors and small Rogowski sensors, the compensation steps that are used to isolate the halo current signal are described. The electrical and mechanical design of compact under-tile resistive shunts and mini-Rogowski coils is described. Example data from the various systems are shown.

A two-dimensional statistical framework connecting thermodynamic profiles with filaments in the scrape off layer and application to experiments

NASA Astrophysics Data System (ADS)

Militello, F.; Farley, T.; Mukhi, K.; Walkden, N.; Omotani, J. T.

2018-05-01

A statistical framework was introduced in Militello and Omotani [Nucl. Fusion 56, 104004 (2016)] to correlate the dynamics and statistics of L-mode and inter-ELM plasma filaments with the radial profiles of thermodynamic quantities they generate in the Scrape Off Layer. This paper extends the framework to cases in which the filaments are emitted from the separatrix at different toroidal positions and with a finite toroidal velocity. It is found that the toroidal velocity does not affect the profiles, while the toroidal distribution of filament emission renormalises the waiting time between two events. Experimental data collected by visual camera imaging are used to evaluate the statistics of the fluctuations, to inform the choice of the probability distribution functions used in the application of the framework. It is found that the toroidal separation of the filaments is exponentially distributed, thus suggesting the lack of a toroidal modal structure. Finally, using these measurements, the framework is applied to an experimental case and good agreement is found.

Toroidal magnetized plasma device with sheared magnetic field lines using an internal ring conductor.

PubMed

Pierre, Th

2013-01-01

In a new toroidal laboratory plasma device including a poloidal magnetic field created by an internal circular conductor, the confinement efficiency of the magnetized plasma and the turbulence level are studied in different situations. The plasma density is greatly enhanced when a sufficiently large poloidal magnetic field is established. Moreover, the instabilities and the turbulence usually found in toroidal devices without sheared magnetic field lines are suppressed by the finite rotational transform. The particle confinement time is estimated from the measurement of the plasma decay time. It is compared to the Bohm diffusion time and to the value predicted by different diffusion models, in particular neoclassical diffusion involving trapped particles.

Scaling Study of Reconnection Heating in Torus Plasma Merging Experiments

NASA Astrophysics Data System (ADS)

Ono, Yasushi; Akimitsu, Moe; Sawada, Asuka; Cao, Qinghong; Koike, Hideya; Hatano, Hironori; Kaneda, Taishi; Tanabe, Hiroshi

2017-10-01

We have been investigating toroidal plasma merging and reconnection for high-power heating of spherical tokamak (ST) and field-reversed configuration (FRC), using TS-3 (ST, FRC: R =0.2m, 1985-), TS-4 (ST, FRC: R =0.5m, 2000-), UTST (ST: R =0.45m, 2008-) and MAST (ST: R =0.9m, 2000-) devices. The series of merging experiments made clear the promising scaling and characteristics of reconnection heating: (i) its ion heating energy that scales with square of the reconnecting magnetic field Brec, (ii) its energy loss lower than 10%, (iii) its ion heating energy (in the downstream) 10 time larger than its electron heating energy (at around X-point) and (iv) low dependence of ion heating on the guide (toroidal) field Bg. The Brec2-scalingwas obtained when the current sheet was compressed to the order of ion gyrodadius. When the sheet was insufficiently compressed, the measured ion temperature was lower than the scaling prediction. Based on this scaling, we realized significant ion heating up to 1.2keV in MAST after 2D elucidation of ion heating up to 250eV in TS-3 [3,4]. This promising scaling leads us to new high Brec reconnection heating experiments for future direct access to burning plasma: TS-U (2017-) in Univ. Tokyo and ST-40 in Tokamak Energy Inc. (2017-). This presentation reviews major progresses in those toroidal plasma merging experiments for physics and fusion applications of magnetic reconnection.

Influence of toroidal rotation on tearing modes

NASA Astrophysics Data System (ADS)

Cai, Huishan; Cao, Jintao; Li, Ding

2017-10-01

Tearing modes stability analysis including toroidal rotation is studied. It is found that rotation affects the stability of tearing modes mainly through the interaction with resistive inner region of tearing mode. The coupling of magnetic curvature with centrifugal force and Coriolis force provides a perturbed perpendicular current, and a return parallel current is induced to affect the stability of tearing modes. Toroidal rotation plays a stable role, which depends on the magnitude of Mach number and adiabatic index Γ, and is independent on the direction of toroidal rotation. For Γ >1, the scaling of growth rate is changed for typical Mach number in present tokamaks. For Γ = 1 , the scaling keeps unchanged, and the effect of toroidal rotation is much less significant, compared with that for Γ >1. National Magnetic Confinement Fusion Science Program and National Science Foundation of China under Grants No. 2014GB106004, No. 2013GB111000, No. 11375189, No. 11075161 and No. 11275260, and Youth Innovation Promotion Association CAS.

Transport and Dynamics in Toroidal Fusion Systems

DOE Office of Scientific and Technical Information (OSTI.GOV)

Schnack, Dalton D

2006-05-16

This document reports the successful completion of the OFES Theory Milestone for FY2005, namely, Perform parametric studies to better understand the edge physics regimes of laboratory experiments. Simulate at increased resolution (up to 20 toroidal modes), with density evolution, late into the nonlinear phase and compare results from different types of edge modes. Simulate a single case including a study of heat deposition on nearby material walls. The linear stability properties and nonlinear evolution of Edge Localized Modes (ELMs) in tokamak plasmas are investigated through numerical computation. Data from the DIII-D device at General Atomics (http://fusion.gat.com/diii-d/) is used for themore » magnetohydrodynamic (MHD) equilibria, but edge parameters are varied to reveal important physical effects. The equilibrium with very low magnetic shear produces an unstable spectrum that is somewhat insensitive to dissipation coefficient values. Here, linear growth rates from the non-ideal NIMROD code (http://nimrodteam.org) agree reasonably well with ideal, i.e. non-dissipative, results from the GATO global linear stability code at low toroidal mode number (n) and with ideal results from the ELITE edge linear stability code at moderate to high toroidal mode number. Linear studies with a more realistic sequence of MHD equilibria (based on DIII-D discharge 86166) produce more significant discrepancies between the ideal and non-ideal calculations. The maximum growth rate for the ideal computations occurs at toroidal mode index n=10, whereas growth rates in the non-ideal computations continue to increase with n unless strong anisotropic thermal conduction is included. Recent modeling advances allow drift effects associated with the Hall electric field and gyroviscosity to be considered. A stabilizing effect can be observed in the preliminary results, but while the distortion in mode structure is readily apparent at n=40, the growth rate is only 13% less than the non-ideal MHD result. Computations performed with a non-local kinetic closure for parallel electron thermal conduction that is valid over all collisionality regimes identify thermal diffusivity ratios of {chi}{sub ||}/{chi}{sub {perpendicular}} ~ 10{sup 7} - 10{sup 8} as appropriate when using collisional heat flux modeling for these modes. Adding significant parallel viscosity proves to have little effect. Nonlinear ELM computations solve the resistive MHD model with toroidal resolution 0{<=}n{<=}21, including anisotropic thermal conduction, temperature-dependent resistivity, and number density evolution. The computations are based on a realistic equilibrium with high pedestal temperature from the linear study. When the simulated ELM grows to appreciable amplitude, ribbon-like thermal structures extend from the separatrix to the wall as the spectrum broadens about a peak at n=13. Analysis of the results finds the heat flux on the wall to be very nonuniform with greatest intensity occurring in spots on the top and bottom of the chamber. Net thermal energy loss occurs on a time-scale of 100 {micro}s, and the instantaneous loss rate exceeds 1 GW.« less

Continuum absorption in the vicinity of the toroidicity-induced Alfvén gap

DOE PAGES

Li, M.; Breizman, B. N.; Zheng, L. J.; ...

2015-12-04

Excitation of Alfvén modes is commonly viewed as a concern for energetic particle confinement in burning plasmas. The 3.5 MeValpha particles produced by fusion may be affected as well as other fast ions in both present and future devices. Continuum damping of such modes is one of the key factors that determine their excitation thresholds and saturation levels. This work examines the resonant dissipative response of the Alfvén continuum to an oscillating driving current when the driving frequency is slightly outside the edges of the toroidicity-induced spectral gap. The problem is largely motivated by the need to describe the continuummore » absorption in the frequency sweeping events. Akey element of this problem is the negative interference of the two closely spaced continuum crossing points.Weexplain why the lower and upper edges of the gap can have very different continuum absorption features. Lastly, the difference is associated with an eigenmode whose frequency can be arbitrarily close to the upper edge of the gap whereas the lower edge of the gap is always a finite distance away from the closest eigenmode.« less

Compressional Alfvén eigenmodes in rotating spherical tokamak plasmas

DOE PAGES

Smith, H. M.; Fredrickson, E. D.

2017-02-07

Spherical tokamaks often have a considerable toroidal plasma rotation of several tens of kHz. Compressional Alfvén eigenmodes in such devices therefore experience a frequency shift, which if the plasma were rotating as a rigid body, would be a simple Doppler shift. However, since the rotation frequency depends on minor radius, the eigenmodes are affected in a more complicated way. The eigenmode solver CAE3B (Smith et al 2009 Plasma Phys. Control. Fusion 51 075001) has been extended to account for toroidal plasma rotation. The results show that the eigenfrequency shift due to rotation can be approximated by a rigid body rotationmore » with a frequency computed from a spatial average of the real rotation profile weighted with the eigenmode amplitude. To investigate the effect of extending the computational domain to the vessel wall, a simplified eigenmode equation, yet retaining plasma rotation, is solved by a modified version of the CAE code used in Fredrickson et al (2013 Phys. Plasmas 20 042112). Lastly, both solving the full eigenmode equation, as in the CAE3B code, and placing the boundary at the vessel wall, as in the CAE code, significantly influences the calculated eigenfrequencies.« less

New Technique of AC drive in Tokamak using Permanent Magnets

NASA Astrophysics Data System (ADS)

Matteucci, Jackson; Zolfaghari, Ali

2013-10-01

This study investigates a new technique of capturing the rotational energy of alternating permanent magnets in order to inductively drive an alternating current in tokamak devices. The use of rotational motion bypasses many of the pitfalls seen in typical inductive and non-inductive current drives. Three specific designs are presented and assessed in the following criteria: the profile of the current generated, the RMS loop voltage generated as compared to the RMS power required to maintain it, the system's feasibility from an engineering perspective. All of the analysis has been done under ideal E&M conditions using the Maxwell 3D program. Preliminary results indicate that it is possible to produce an over 99% purely toroidal current with a RMS d Φ/dt of over 150 Tm2/s, driven by 20 MW or less of rotational power. The proposed mechanism demonstrates several key advantages including an efficient mechanical drive system, the generation of pure toroidal currents, and the potential for a quasi-steady state fusion reactor. The following quantities are presented for various driving frequencies and magnet strengths: plasma current generated, loop voltage, torque and power required. This project has been supported by DOE Funding under the SULI program.

Air core poloidal magnetic field system for a toroidal plasma producing device

DOEpatents

Marcus, Frederick B.

1978-01-01

A poloidal magnetics system for a plasma producing device of toroidal configuration is provided that reduces both the total volt-seconds requirement and the magnitude of the field change at the toroidal field coils. The system utilizes an air core transformer wound between the toroidal field (TF) coils and the major axis outside the TF coils. Electric current in the primary windings of this transformer is distributed and the magnetic flux returned by air core windings wrapped outside the toroidal field coils. A shield winding that is closely coupled to the plasma carries a current equal and opposite to the plasma current. This winding provides the shielding function and in addition serves in a fashion similar to a driven conducting shell to provide the equilibrium vertical field for the plasma. The shield winding is in series with a power supply and a decoupling coil located outside the TF coil at the primary winding locations. The present invention requires much less energy than the usual air core transformer and is capable of substantially shielding the toroidal field coils from poloidal field flux.

Preface to Special Topic: Advances in Radio Frequency Physics in Fusion Plasmas

NASA Astrophysics Data System (ADS)

Tuccillo, Angelo A.; Phillips, Cynthia K.; Ceccuzzi, Silvio

2014-06-01

It has long been recognized that auxiliary plasma heating will be required to achieve the high temperature, high density conditions within a magnetically confined plasma in which a fusion "burn" may be sustained by copious fusion reactions. Consequently, the application of radio and microwave frequency electromagnetic waves to magnetically confined plasma, commonly referred to as RF, has been a major part of the program almost since its inception in the 1950s. These RF waves provide heating, current drive, plasma profile control, and Magnetohydrodynamics (MHD) stabilization. Fusion experiments employ electromagnetic radiation in a wide range of frequencies, from tens of MHz to hundreds of GHz. The fusion devices containing the plasma are typically tori, axisymmetric or non, in which the equilibrium magnetic fields are composed of a strong toroidal magnetic field generated by external coils, and a poloidal field created, at least in the symmetric configurations, by currents flowing in the plasma. The waves are excited in the peripheral regions of the plasma, by specially designed launching structures, and subsequently propagate into the core regions, where resonant wave-plasma interactions produce localized heating or other modification of the local equilibrium profiles. Experimental studies coupled with the development of theoretical models and advanced simulation codes over the past 40+ years have led to an unprecedented understanding of the physics of RF heating and current drive in the core of magnetic fusion devices. Nevertheless, there are serious gaps in our knowledge base that continue to have a negative impact on the success of ongoing experiments and that must be resolved as the program progresses to the next generation devices and ultimately to "demo" and "fusion power plant." A serious gap, at least in the ion cyclotron (IC) range of frequencies and partially in the lower hybrid frequency ranges, is the difficulty in coupling large amount of power to the plasma while minimizing the interaction between the plasma and launching structures. These potentially harmful interactions between the plasma and the vessel and launching structures are challenging: (i) significant and variable loss of power in the edge regions of confined plasmas and surrounding vessel structures adversely affect the core plasma performance and lifetime of a device; (ii) the launcher design is partly "trial and error," with the consequence that launchers may have to be reconfigured after initial tests in a given device, at an additional cost. Over the broader frequency range, another serious gap is a quantitative lack of understanding of the combined effects of nonlinear wave-plasma processes, energetic particle interactions and non-axisymmetric equilibrium effects on determining the overall efficiency of plasma equilibrium and stability profile control techniques using RF waves. This is complicated by a corresponding lack of predictive understanding of the time evolution of transport and stability processes in fusion plasmas.

Toroid cavity/coil NMR multi-detector

DOEpatents

Gerald, II, Rex E.; Meadows, Alexander D.; Gregar, Joseph S.; Rathke, Jerome W.

2007-09-18

An analytical device for rapid, non-invasive nuclear magnetic resonance (NMR) spectroscopy of multiple samples using a single spectrometer is provided. A modified toroid cavity/coil detector (TCD), and methods for conducting the simultaneous acquisition of NMR data for multiple samples including a protocol for testing NMR multi-detectors are provided. One embodiment includes a plurality of LC resonant circuits including spatially separated toroid coil inductors, each toroid coil inductor enveloping its corresponding sample volume, and tuned to resonate at a predefined frequency using a variable capacitor. The toroid coil is formed into a loop, where both ends of the toroid coil are brought into coincidence. Another embodiment includes multiple micro Helmholtz coils arranged on a circular perimeter concentric with a central conductor of the toroid cavity.

Observations of toroidicity-induced Alfvén eigenmodes in a reversed field pinch plasma

NASA Astrophysics Data System (ADS)

Regnoli, G.; Bergsâker, H.; Tennfors, E.; Zonca, F.; Martines, E.; Serianni, G.; Spolaore, M.; Vianello, N.; Cecconello, M.; Antoni, V.; Cavazzana, R.; Malmberg, J.-A.

2005-04-01

High frequency peaks in the spectra of magnetic field signals have been detected at the edge of Extrap-T2R [P. R. Brunsell, H. Bergsåker, M. Cecconello, J. R. Drake, R. M. Gravestijn, A. Hedqvist, and J.-A. Malmberg, Plasma Phys. Controlled Fusion, 43, 1457 (2001)]. The measured fluctuation is found to be mainly polarized along the toroidal direction, with high toroidal periodicity n and Alfvénic scaling (f∝B/√mini ). Calculations for a reversed field pinch plasma predict the existence of an edge resonant, high frequency, high-n number toroidicity-induced Alfvén eigenmode with the observed frequency scaling. In addition, gas puffing experiments show that edge density fluctuations are responsible for the rapid changes of mode frequency. Finally a coupling with the electron drift turbulence is proposed as drive mechanism for the eigenmode.

Physics of Tokamak Plasma Start-up

NASA Astrophysics Data System (ADS)

Mueller, Dennis

2012-10-01

This tutorial describes and reviews the state-of-art in tokamak plasma start-up and its importance to next step devices such as ITER, a Fusion Nuclear Science Facility and a Tokamak/ST demo. Tokamak plasma start-up includes breakdown of the initial gas, ramp-up of the plasma current to its final value and the control of plasma parameters during those phases. Tokamaks rely on an inductive component, typically a central solenoid, which has enabled attainment of high performance levels that has enabled the construction of the ITER device. Optimizing the inductive start-up phase continues to be an area of active research, especially in regards to achieving ITER scenarios. A new generation of superconducting tokamaks, EAST and KSTAR, experiments on DIII-D and operation with JET's ITER-like wall are contributing towards this effort. Inductive start-up relies on transformer action to generate a toroidal loop voltage and successful start-up is determined by gas breakdown, avalanche physics and plasma-wall interaction. The goal of achieving steady-sate tokamak operation has motivated interest in other methods for start-up that do not rely on the central solenoid. These include Coaxial Helicity Injection, outer poloidal field coil start-up, and point source helicity injection, which have achieved 200, 150 and 100 kA respectively of toroidal current on closed flux surfaces. Other methods including merging reconnection startup and Electron Bernstein Wave (EBW) plasma start-up are being studied on various devices. EBW start-up generates a directed electron channel due to wave particle interaction physics while the other methods mentioned rely on magnetic helicity injection and magnetic reconnection which are being modeled and understood using NIMROD code simulations.

Alpha-driven magnetohydrodynamics (MHD) and MHD-induced alpha loss in the Tokamak Fusion Test Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Chang, Z.; Nazikian, R.; Fu, G.Y.

1997-02-01

Alpha-driven toroidal Alfven eigenmodes (TAEs) are observed as predicted by theory in the post neutral beam phase in high central q (safety factor) deuterium-tritium (D-T) plasmas in the Tokamak Fusion Test Reactor (TFTR). The mode location, poloidal structure and the importance of q profile for TAE instability are discussed. So far no alpha particle loss due to these modes was detected due to the small mode amplitude. However, alpha loss induced by kinetic ballooning modes (KBMs) was observed in high confinement D-T discharges. Particle orbit simulation demonstrates that the wave-particle resonant interaction can explain the observed correlation between the increasemore » in alpha loss and appearance of multiple high-n (n {ge} 6, n is the toroidal mode number) modes.« less

Full-wave simulations of ICRF heating regimes in toroidal plasma with non-Maxwellian distribution functions

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bertelli, N.; Valeo, E. J.; Green, D. L.

At the power levels required for significant heating and current drive in magnetically-confined toroidal plasma, modification of the particle distribution function from a Maxwellian shape is likely (Stix 1975 Nucl. Fusion 15 737), with consequent changes in wave propagation and in the location and amount of absorption. In order to study these effects computationally, both the finite-Larmor-radius and the high-harmonic fast wave (HHFW), versions of the full-wave, hot-plasma toroidal simulation code TORIC (Brambilla 1999 Plasma Phys. Control. Fusion 41 1 and Brambilla 2002 Plasma Phys. Control. Fusion 44 2423), have been extended to allow the prescription of arbitrary velocity distributionsmore » of the form f(v(parallel to), v(perpendicular to) , psi, theta). For hydrogen (H) minority heating of a deuterium (D) plasma with anisotropic Maxwellian H distributions, the fractional H absorption varies significantly with changes in parallel temperature but is essentially independent of perpendicular temperature. On the other hand, for HHFW regime with anisotropic Maxwellian fast ion distribution, the fractional beam ion absorption varies mainly with changes in the perpendicular temperature. The evaluation of the wave-field and power absorption, through the full wave solver, with the ion distribution function provided by either a Monte-Carlo particle and Fokker-Planck codes is also examined for Alcator C-Mod and NSTX plasmas. Non-Maxwellian effects generally tend to increase the absorption with respect to the equivalent Maxwellian distribution.« less

Full-wave simulations of ICRF heating regimes in toroidal plasmas with non-Maxwellian distribution functions

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bertelli, N.; Valeo, E.J.; Green, D.L.

At the power levels required for significant heating and current drive in magnetically-confined toroidal plasma, modification of the particle distribution function from a Maxwellian shape is likely [T. H. Stix, Nucl. Fusion, 15 737 (1975)], with consequent changes in wave propagation and in the location and amount of absorption. In order to study these effects computationally, both the finite-Larmor-radius and the high-harmonic fast wave (HHFW), versions of the full-wave, hot-plasma toroidal simulation code TORIC [M. Brambilla, Plasma Phys. Control. Fusion 41, 1 (1999) and M. Brambilla, Plasma Phys. Control. Fusion 44, 2423 (2002)], have been extended to allow the prescriptionmore » of arbitrary velocity distributions of the form f(v||, v_perp, psi , theta). For hydrogen (H) minority heating of a deuterium (D) plasma with anisotropic Maxwellian H distributions, the fractional H absorption varies significantly with changes in parallel temperature but is essentially independent of perpendicular temperature. On the other hand, for HHFW regime with anisotropic Maxwellian fast ion distribution, the fractional beam ion absorption varies mainly with changes in the perpendicular temperature. The evaluation of the wave-field and power absorption, through the full wave solver, with the ion distribution function provided by either aMonte-Carlo particle and Fokker-Planck codes is also examined for Alcator C-Mod and NSTX plasmas. Non-Maxwellian effects generally tends to increase the absorption with respect to the equivalent Maxwellian distribution.« less

Full-wave simulations of ICRF heating regimes in toroidal plasma with non-Maxwellian distribution functions

NASA Astrophysics Data System (ADS)

Bertelli, N.; Valeo, E. J.; Green, D. L.; Gorelenkova, M.; Phillips, C. K.; Podestà, M.; Lee, J. P.; Wright, J. C.; Jaeger, E. F.

2017-05-01

At the power levels required for significant heating and current drive in magnetically-confined toroidal plasma, modification of the particle distribution function from a Maxwellian shape is likely (Stix 1975 Nucl. Fusion 15 737), with consequent changes in wave propagation and in the location and amount of absorption. In order to study these effects computationally, both the finite-Larmor-radius and the high-harmonic fast wave (HHFW), versions of the full-wave, hot-plasma toroidal simulation code TORIC (Brambilla 1999 Plasma Phys. Control. Fusion 41 1 and Brambilla 2002 Plasma Phys. Control. Fusion 44 2423), have been extended to allow the prescription of arbitrary velocity distributions of the form f≤ft({{v}\\parallel},{{v}\\bot},\\psi,θ \\right) . For hydrogen (H) minority heating of a deuterium (D) plasma with anisotropic Maxwellian H distributions, the fractional H absorption varies significantly with changes in parallel temperature but is essentially independent of perpendicular temperature. On the other hand, for HHFW regime with anisotropic Maxwellian fast ion distribution, the fractional beam ion absorption varies mainly with changes in the perpendicular temperature. The evaluation of the wave-field and power absorption, through the full wave solver, with the ion distribution function provided by either a Monte-Carlo particle and Fokker-Planck codes is also examined for Alcator C-Mod and NSTX plasmas. Non-Maxwellian effects generally tend to increase the absorption with respect to the equivalent Maxwellian distribution.

Full-wave simulations of ICRF heating regimes in toroidal plasma with non-Maxwellian distribution functions

DOE PAGES

Bertelli, N.; Valeo, E. J.; Green, D. L.; ...

2017-04-03

At the power levels required for significant heating and current drive in magnetically-confined toroidal plasma, modification of the particle distribution function from a Maxwellian shape is likely (Stix 1975 Nucl. Fusion 15 737), with consequent changes in wave propagation and in the location and amount of absorption. In order to study these effects computationally, both the finite-Larmor-radius and the high-harmonic fast wave (HHFW), versions of the full-wave, hot-plasma toroidal simulation code TORIC (Brambilla 1999 Plasma Phys. Control. Fusion 41 1 and Brambilla 2002 Plasma Phys. Control. Fusion 44 2423), have been extended to allow the prescription of arbitrary velocity distributionsmore » of the form f(v(parallel to), v(perpendicular to) , psi, theta). For hydrogen (H) minority heating of a deuterium (D) plasma with anisotropic Maxwellian H distributions, the fractional H absorption varies significantly with changes in parallel temperature but is essentially independent of perpendicular temperature. On the other hand, for HHFW regime with anisotropic Maxwellian fast ion distribution, the fractional beam ion absorption varies mainly with changes in the perpendicular temperature. The evaluation of the wave-field and power absorption, through the full wave solver, with the ion distribution function provided by either a Monte-Carlo particle and Fokker-Planck codes is also examined for Alcator C-Mod and NSTX plasmas. Non-Maxwellian effects generally tend to increase the absorption with respect to the equivalent Maxwellian distribution.« less

Impact of an integrated core/SOL description on the R and B T optimization of tokamak fusion reactors

NASA Astrophysics Data System (ADS)

Siccinio, M.; Fable, E.; Angioni, C.; Saarelma, S.; Scarabosio, A.; Zohm, H.

2018-01-01

An updated and improved version of the 0D divertor and scrape-off layer (SOL) model published in Siccinio et al (2016 Plasma Phys. Control. Fusion 58 125011) was coupled with the 1.5D transport code ASTRA (Pereverzev 1991 IPP Report 5/42, Pereverzev and Yushmanov 2002 IPP Report 5/98 and Fable et al 2013 Plasma Phys. Control. Fusion 55 124028). The resulting numerical tool was employed for various scans in the major radius R and in the toroidal magnetic field B T—for different safety factors q, allowable loop voltages V loop and H factors—in order to identify the most convenient choices for an electricity producing tokamak. Such a scenario analysis was carried out evaluating self-consistently, and simultaneously, the core profile and transport effects, which significantly impact on the fusion power outcome, and the divertor heat loads, which represent one of the most critical issues in view of the realization of fusion power plants (Zohm et al 2013 Nucl. Fusion 53 073019 and Wenninger et al 2017 Nucl. Fusion 57 046002). The main result is that, when divertor limits are enforced, the curves at constant electrical power output are closed on themselves in the R-BT plane, and a maximum achievable power exists—i.e. no benefits would be obtained from a further increase in R and B T once the optimum is reached. This result appears as an intrinsic physical limit for all those devices where a radiative SOL is needed to deal with the power exhaust, and where a lower limit on the power crossing the separatrix (e.g. because of the L-H transition) is present.

An Ultrasound Image-Based Dynamic Fusion Modeling Method for Predicting the Quantitative Impact of In Vivo Liver Motion on Intraoperative HIFU Therapies: Investigations in a Porcine Model

PubMed Central

N'Djin, W. Apoutou; Chapelon, Jean-Yves; Melodelima, David

2015-01-01

Organ motion is a key component in the treatment of abdominal tumors by High Intensity Focused Ultrasound (HIFU), since it may influence the safety, efficacy and treatment time. Here we report the development in a porcine model of an Ultrasound (US) image-based dynamic fusion modeling method for predicting the effect of in vivo motion on intraoperative HIFU treatments performed in the liver in conjunction with surgery. A speckle tracking method was used on US images to quantify in vivo liver motions occurring intraoperatively during breathing and apnea. A fusion modeling of HIFU treatments was implemented by merging dynamic in vivo motion data in a numerical modeling of HIFU treatments. Two HIFU strategies were studied: a spherical focusing delivering 49 juxtapositions of 5-second HIFU exposures and a toroidal focusing using 1 single 40-second HIFU exposure. Liver motions during breathing were spatially homogenous and could be approximated to a rigid motion mainly encountered in the cranial-caudal direction (f = 0.20Hz, magnitude >13mm). Elastic liver motions due to cardiovascular activity, although negligible, were detectable near millimeter-wide sus-hepatic veins (f = 0.96Hz, magnitude <1mm). The fusion modeling quantified the deleterious effects of respiratory motions on the size and homogeneity of a standard “cigar-shaped” millimetric lesion usually predicted after a 5-second single spherical HIFU exposure in stationary tissues (Dice Similarity Coefficient: DSC<45%). This method assessed the ability to enlarge HIFU ablations during respiration, either by juxtaposing “cigar-shaped” lesions with spherical HIFU exposures, or by generating one large single lesion with toroidal HIFU exposures (DSC>75%). Fusion modeling predictions were preliminarily validated in vivo and showed the potential of using a long-duration toroidal HIFU exposure to accelerate the ablation process during breathing (from 0.5 to 6 cm3·min-1). To improve HIFU treatment control, dynamic fusion modeling may be interesting for assessing numerically focusing strategies and motion compensation techniques in more realistic conditions. PMID:26398366

Plasma Density Effects on Toroidal Flow Stabilization of Edge Localized Modes

NASA Astrophysics Data System (ADS)

Cheng, Shikui; Zhu, Ping; Banerjee, Debabrata

2016-10-01

Recent EAST experiments have demonstrated mitigation and suppression of edge localized modes (ELMs) with toroidal rotation flow in higher collisionality regime, suggesting potential roles of plasma density. In this work, the effects of plasma density on the toroidal flow stabilization of the high- n edge localized modes have been extensively studied in linear calculations for a circular-shaped limiter H-mode tokamak, using the initial-value extended MHD code NIMROD. In the single MHD model, toroidal flow has a weak stabilizing effects on the high- n modes. Such a stabilization, however, can be significantly enhanced with the increase in plasma density. Furthermore, our calculations show that the enhanced stabilization of high- n modes from toroidal flow with higher edge plasma density persists in the 2-fluid MHD model. These findings may explain the ELM mitigation and suppression by toroidal rotation in higher collisionality regime due to the enhancement of plasma density obtained in EAST experiment. Supported by the National Magnetic Confinement Fusion Program of China under Grant Nos. 2014GB124002 and 2015GB101004, the 100 Talent Program and the President International Fellowship Initiative of Chinese Academy of Sciences.

System and method of operating toroidal magnetic confinement devices

DOEpatents

Chance, Morrell S.; Jardin, Stephen C.; Stix, Thomas H.; Grimm, deceased, Ray C.; Manickam, Janardhan; Okabayashi, Michio

1987-01-01

For toroidal magnetic confinement devices the second region of stability against ballooning modes can be accessed with controlled operation. Under certain modes of operation, the first and second stability regions may be joined together. Accessing the second region of stability is accomplished by forming a bean-shaped plasma and increasing the indentation until a critical value of indentation is reached. A pusher coil, located at the inner-major-radius side of the device, is engaged to form a bean-shaped poloidal cross-section in the plasma.

Tokamak foundation in USSR/Russia 1950-1990

NASA Astrophysics Data System (ADS)

Smirnov, V. P.

2010-01-01

In the USSR, nuclear fusion research began in 1950 with the work of I.E. Tamm, A.D. Sakharov and colleagues. They formulated the principles of magnetic confinement of high temperature plasmas, that would allow the development of a thermonuclear reactor. Following this, experimental research on plasma initiation and heating in toroidal systems began in 1951 at the Kurchatov Institute. From the very first devices with vessels made of glass, porcelain or metal with insulating inserts, work progressed to the operation of the first tokamak, T-1, in 1958. More machines followed and the first international collaboration in nuclear fusion, on the T-3 tokamak, established the tokamak as a promising option for magnetic confinement. Experiments continued and specialized machines were developed to test separately improvements to the tokamak concept needed for the production of energy. At the same time, research into plasma physics and tokamak theory was being undertaken which provides the basis for modern theoretical work. Since then, the tokamak concept has been refined by a world-wide effort and today we look forward to the successful operation of ITER.

Plasma-resistivity-induced strong damping of the kinetic resistive wall mode.

PubMed

He, Yuling; Liu, Yueqiang; Liu, Yue; Hao, Guangzhou; Wang, Aike

2014-10-24

An energy-principle-based dispersion relation is derived for the resistive wall mode, which incorporates both the drift kinetic resonance between the mode and energetic particles and the resistive layer physics. The equivalence between the energy-principle approach and the resistive layer matching approach is first demonstrated for the resistive plasma resistive wall mode. As a key new result, it is found that the resistive wall mode, coupled to the favorable average curvature stabilization inside the resistive layer (as well as the toroidal plasma flow), can be substantially more stable than that predicted by drift kinetic theory with fast ion stabilization, but with the ideal fluid assumption. Since the layer stabilization becomes stronger with decreasing plasma resistivity, this regime is favorable for reactor scale, high-temperature fusion devices.

The radiation asymmetry in MGI rapid shutdown on J-TEXT tokamak

NASA Astrophysics Data System (ADS)

Tong, Ruihai; Chen, Zhongyong; Huang, Duwei; Cheng, Zhifeng; Zhang, Xiaolong; Zhuang, Ge; J-TEXT Team

2017-10-01

Disruptions, the sudden termination of tokamak fusion plasmas by instabilities, have the potential to cause severe material wall damage to large tokamaks like ITER. The mitigation of disruption damage is an essential part of any fusion reactor system. Massive gas injection (MGI) rapid shutdown is a technique in which large amounts of noble gas are injected into the plasma in order to safely radiate the plasma energy evenly over the entire plasma-facing first wall. However, the radiated energy during the thermal quench (TQ) in massive gas injection (MGI) induced disruptions is found toroidal asymmetric, and the degrees of asymmetry correlate with the gas penetration and MGI induced magnetohydrodynamics (MHD) activities. A toroidal and poloidal array of ultraviolet photodiodes (AXUV) has been developed to investigate the radiation asymmetry on J-TEXT tokamak. Together with the upgraded mirnov probe arrays, the relation between MGI triggered MHD activities with radiation asymmetry is studied.

A short introduction to the status and motivation for reversed field pinch and compact toroid research

NASA Astrophysics Data System (ADS)

Dreicer, H.

1987-09-01

Potential commercial fusion power systems must be acceptable from a safety and environmental standpoint. They must also promise to be competitive with other sources of energy (i.e., fossil, fission, etc.) when considered from the standpoint of the cost of electricity (COE) and the unit direct cost (UDC) in dollars/kWe. These costs are affected by a host of factors including recirculating power, plant availability, construction time, capital cost, etc., and are influenced by technological complexity. In an attempt to meet these requirements, the emphasis of fusion research in the United States has been moving toward smaller, lower-cost systems. There is increased interest in higher beta tokamaks and stellarators, and in compact alternate concepts such as the Reversed Field Pinch (RFP) and the Compact Toroids (CTs) which are, in part, the subject of this paper.

Classical impurity ion confinement in a toroidal magnetized fusion plasma.

PubMed

Kumar, S T A; Den Hartog, D J; Caspary, K J; Magee, R M; Mirnov, V V; Chapman, B E; Craig, D; Fiksel, G; Sarff, J S

2012-03-23

High-resolution measurements of impurity ion dynamics provide first-time evidence of classical ion confinement in a toroidal, magnetically confined plasma. The density profile evolution of fully stripped carbon is measured in MST reversed-field pinch plasmas with reduced magnetic turbulence to assess Coulomb-collisional transport without the neoclassical enhancement from particle drift effects. The impurity density profile evolves to a hollow shape, consistent with the temperature screening mechanism of classical transport. Corroborating methane pellet injection experiments expose the sensitivity of the impurity particle confinement time to the residual magnetic fluctuation amplitude.

Effect of Gyroviscosity on Tearing Modes in Tokamak Plasmas

NASA Astrophysics Data System (ADS)

White, Ryan; Glasser, Alan

2017-10-01

We present an extension of the Glasser-Greene-Johnson equations, incorporating the Braginskii gyroviscosity. It is found that the dominant terms from the gyroviscous stress are all due to poloidal variation of the equilibrium profile, implying that these physical effects are not captured in a large-aspect-ratio (cylindrical) model. Because these purely toroidal contributions dominate, we conclude that thewell-known ``gyroviscous cancellation'' is a higher-order effect in toroidal confinement systems. We also present preliminary numerical results showing the effect of gyroviscosity on tearing mode stability. ORISE/DOE Fusion Energy Sciences Postdoctoral Fellowship.

Advanced Power Conversion Efficiency in Inventive Plasma for Hybrid Toroidal Reactor

NASA Astrophysics Data System (ADS)

Hançerlioğullari, Aybaba; Cini, Mesut; Güdal, Murat

2013-08-01

Apex hybrid reactor has a good potential to utilize uranium and thorium fuels in the future. This toroidal reactor is a type of system that facilitates the occurrence of the nuclear fusion and fission events together. The most important feature of hybrid reactor is that the first wall surrounding the plasma is liquid. The advantages of utilizing a liquid wall are high power density capacity good power transformation productivity, the magnitude of the reactor's operational duration, low failure percentage, short maintenance time and the inclusion of the system's simple technology and material. The analysis has been made using the MCNP Monte Carlo code and ENDF/B-V-VI nuclear data. Around the fusion chamber, molten salts Flibe (LI2BeF4), lead-lithium (PbLi), Li-Sn, thin-lityum (Li20Sn80) have used as cooling materials. APEX reactor has modeled in the torus form by adding nuclear materials of low significance in the specified percentages between 0 and 12 % to the molten salts. In this study, the neutronic performance of the APEX fusion reactor using various molten salts has been investigated. The nuclear parameters of Apex reactor has been searched for Flibe (LI2BeF4) and Li-Sn, for blanket layers. In case of usage of the Flibe (LI2BeF4), PbLi, and thin-lityum (Li20Sn80) salt solutions at APEX toroidal reactors, fissile material production per source neutron, tritium production speed, total fission rate, energy reproduction factor has been calculated, the results obtained for both salt solutions are compared.

Analysis of Alfvén eigenmode destabilization by energetic particles in Large Helical Device using a Landau-closure model

DOE PAGES

Varela, Jacobo Rodriguez; Spong, D. A.; Garcia, L.

2017-03-06

Here, energetic particle populations in nuclear fusion experiments can destabilize the Alfvén Eigenmodes through inverse Landau damping and couplings with gap modes in the shear Alfvén continua. We use the reduced MHD equations to describe the linear evolution of the poloidal flux and the toroidal component of the vorticity in a full 3D system, coupled with equations of density and parallel velocity moments for the energetic particles. We add the Landau damping and resonant destabilization effects using a closure relation. We apply the model to study the Alfvén mode stability in the inward-shifted configurations of the Large Helical Device (LHD), performing a parametric analysis of the energetic particle β (more » $${{\\beta}_{f}}$$ ) in a range of realistic values, the ratios of the energetic particle thermal/Alfvén velocities ($${{V}_{\\text{th}}}/{{V}_{A0}}$$ ), the magnetic Lundquist numbers (S) and the toroidal modes (n). The n = 1 and n = 2 TAEs are destabilized, although the n = 3 and n = 4 TAEs are weakly perturbed. The most unstable configurations are associated with the density gradients of energetic particles in the plasma core: the TAEs are destabilized, even for small energetic particle populations, if their thermal velocity is lower than 0.4 times the Alfvén velocity. The frequency range of MHD bursts measured in the LHD are 50–70 kHz for the n = 1 and 60–80 kHz for the n = 2 TAE, which is consistent with the model predictions.« less

Effects of the current boundary conditions at the plasma-gun gap on density in SSPX

NASA Astrophysics Data System (ADS)

Kolesnikov, Roman; Lodestro, L. L.; Meyer, W. H.

2012-10-01

The Sustained Spheromak Physics Experiment (SSPX) was a toroidal magnetic-confinement device without toroidal magnetic-field coils or a central transformer but which generated core-plasma currents by dynamo processes driven by coaxial plasma-gun injection into a flux-conserving vessel. Record electron temperatures in a spheromak (Te˜500eV) were achieved, and final results of the SSPX program were reported in [1]. Plasma density, which depended strongly on wall conditions, was an important parameter in SSPX. It was observed that density rises with Igun and that confinement improved as the density was lowered. Shortly after the last experiments, a new feature was added to the Corsica code's solver used to reconstruct SSPX equilibria. Motivated by n=0 fields observed in NIMROD simulations of SSPX, an insulating boundary condition was implemented at the plasma-gun gap. Using this option we will perform new reconstructions of SSPX equilibria and look for correlations between the location of the separatrix (which moves up the gun wall and onto the insulating gap as Igun increases) and plasma density and magnetic-flux amplification [2].[4pt] [1] H. S. McLean, APS, DPP, Dallas, TX, 2008.[0pt] [2] E. B. Hooper et al., Nucl. Fusion 47, 1064 (2007).

Mode control using two electrodes on HBT-EP

NASA Astrophysics Data System (ADS)

Stewart, I. G.; Brooks, J. W.; Levesque, J. P.; Mauel, M. E.; Navratil, G. A.

2017-10-01

Understanding the effects of plasma rotation on magnetohydrodynamic (MHD) modes and tokamak plasma stability is important for performance enhancement of current magnetic confinement experiments and to future fusion devices such as ITER. In order to control plasma rotation, two molybdenum electrodes have been installed on HBT-EP toroidally separated by 144 degrees. This allows independent biasing of the two probes both spatially and temporally. When the bias probes are inserted into the edge of the plasma and a voltage is applied, the probes drive radial currents and produce plasma flow from the torque induced by the currents. If the bias probe voltage is sufficiently positive, the MHD mode rotation transitions into a state with a rapid mode rotation frequency (in excess of 25 kHz) in the direction opposite to mode rotation without bias. The transition into this reversed rotation state occurs when the torque exceeds a threshold, which can depend upon the phase of an applied n = 1 error field. We present recent studies of the two-electrode system on mode rotation, mode stability, and the toroidal symmetry of the radial current through the scrape-off-layer (SOL) during MHD activity and applied magnetic perturbations. Supported by U.S. DOE Grant DE-FG02-86ER53222.

Fusion nuclear science facilities and pilot plants based on the spherical tokamak

NASA Astrophysics Data System (ADS)

Menard, J. E.; Brown, T.; El-Guebaly, L.; Boyer, M.; Canik, J.; Colling, B.; Raman, R.; Wang, Z.; Zhai, Y.; Buxton, P.; Covele, B.; D'Angelo, C.; Davis, A.; Gerhardt, S.; Gryaznevich, M.; Harb, M.; Hender, T. C.; Kaye, S.; Kingham, D.; Kotschenreuther, M.; Mahajan, S.; Maingi, R.; Marriott, E.; Meier, E. T.; Mynsberge, L.; Neumeyer, C.; Ono, M.; Park, J.-K.; Sabbagh, S. A.; Soukhanovskii, V.; Valanju, P.; Woolley, R.

2016-10-01

A fusion nuclear science facility (FNSF) could play an important role in the development of fusion energy by providing the nuclear environment needed to develop fusion materials and components. The spherical torus/tokamak (ST) is a leading candidate for an FNSF due to its potentially high neutron wall loading and modular configuration. A key consideration for the choice of FNSF configuration is the range of achievable missions as a function of device size. Possible missions include: providing high neutron wall loading and fluence, demonstrating tritium self-sufficiency, and demonstrating electrical self-sufficiency. All of these missions must also be compatible with a viable divertor, first-wall, and blanket solution. ST-FNSF configurations have been developed simultaneously incorporating for the first time: (1) a blanket system capable of tritium breeding ratio TBR  ≈  1, (2) a poloidal field coil set supporting high elongation and triangularity for a range of internal inductance and normalized beta values consistent with NSTX/NSTX-U previous/planned operation, (3) a long-legged divertor analogous to the MAST-U divertor which substantially reduces projected peak divertor heat-flux and has all outboard poloidal field coils outside the vacuum chamber and superconducting to reduce power consumption, and (4) a vertical maintenance scheme in which blanket structures and the centerstack can be removed independently. Progress in these ST-FNSF missions versus configuration studies including dependence on plasma major radius R 0 for a range 1 m-2.2 m are described. In particular, it is found the threshold major radius for TBR  =  1 is {{R}0}≥slant 1.7 m, and a smaller R 0  =  1 m ST device has TBR  ≈  0.9 which is below unity but substantially reduces T consumption relative to not breeding. Calculations of neutral beam heating and current drive for non-inductive ramp-up and sustainment are described. An A  =  2, R 0  =  3 m device incorporating high-temperature superconductor toroidal field coil magnets capable of high neutron fluence and both tritium and electrical self-sufficiency is also presented following systematic aspect ratio studies.

Fusion nuclear science facilities and pilot plants based on the spherical tokamak

DOE PAGES

Menard, J. E.; Brown, T.; El-Guebaly, L.; ...

2016-08-16

Here, a fusion nuclear science facility (FNSF) could play an important role in the development of fusion energy by providing the nuclear environment needed to develop fusion materials and components. The spherical torus/tokamak (ST) is a leading candidate for an FNSF due to its potentially high neutron wall loading and modular configuration. A key consideration for the choice of FNSF configuration is the range of achievable missions as a function of device size. Possible missions include: providing high neutron wall loading and fluence, demonstrating tritium self-sufficiency, and demonstrating electrical self-sufficiency. All of these missions must also be compatible with a viable divertor, first-wall, and blanket solution. ST-FNSF configurations have been developed simultaneously incorporating for the first time: (1) a blanket system capable of tritium breeding ratio TBR ≈ 1, (2) a poloidal field coil set supporting high elongation and triangularity for a range of internal inductance and normalized beta values consistent with NSTX/NSTX-U previous/planned operation, (3) a long-legged divertor analogous to the MAST-U divertor which substantially reduces projected peak divertor heat-flux and has all outboard poloidal field coils outside the vacuum chamber and superconducting to reduce power consumption, and (4) a vertical maintenance scheme in which blanket structures and the centerstack can be removed independently. Progress in these ST-FNSF missions versus configuration studies including dependence on plasma major radius R 0 for a range 1 m–2.2 m are described. In particular, it is found the threshold major radius for TBR = 1 ismore » $${{R}_{0}}\\geqslant 1.7$$ m, and a smaller R 0 = 1 m ST device has TBR ≈ 0.9 which is below unity but substantially reduces T consumption relative to not breeding. Calculations of neutral beam heating and current drive for non-inductive ramp-up and sustainment are described. An A = 2, R = 3 m device incorporating high-temperature superconductor toroidal field coil magnets capable of high neutron fluence and both tritium and electrical self-sufficiency is also presented following systematic aspect ratio studies.« less

Full orbit computations of ripple-induced fusion {alpha}-particle losses from burning tokamak plasmas

DOE Office of Scientific and Technical Information (OSTI.GOV)

McClements, K.G.

A full orbit code is used to compute collisionless losses of fusion {alpha} particles from three proposed burning plasma tokamaks: the International Tokamak Experimental Reactor (ITER); a spherical tokamak power plant (STPP) [T. C. Hender, A. Bond, J. Edwards, P. J. Karditsas, K. G. McClements, J. Mustoe, D. V. Sherwood, G. M. Voss, and H. R. Wilson, Fusion Eng. Des. 48, 255 (2000)]; and a spherical tokamak components test facility (CTF) [H. R. Wilson, G. M. Voss, R. J. Akers, L. Appel, A. Dnestrovskij, O. Keating, T. C. Hender, M. J. Hole, G. Huysmans, A. Kirk, P. J. Knight, M.more » Loughlin, K. G. McClements, M. R. O'Brien, and D. Yu. Sychugov, Proceedings of the 20th IAEA Fusion Energy Conference, Invited Paper FT/3-1Ra]. It has been suggested that {alpha} particle transport could be enhanced due to cyclotron resonance with the toroidal magnetic field ripple. However, calculations for inductive operation in ITER yield a loss rate that appears to be broadly consistent with the predictions of guiding center theory, falling monotonically as the number of toroidal field coils N is increased (and hence the ripple amplitude is decreased). For STPP and CTF the loss rate does not decrease monotonically with N, but collisionless losses are generally low in absolute terms. As in the case of ITER, there is no evidence that finite Larmor radius effects would seriously degrade fusion {alpha}-particle confinement.« less

TOPICA/TORIC integration for self-consistent antenna and plasma analysis

NASA Astrophysics Data System (ADS)

Maggiora, Riccardo; Lancellotti, Vito; Milanesio, Daniele; Kyrytsya, Volodymyr; Vecchi, Giuseppe; Bonoli, Paul T.; Wright, John C.

2006-10-01

TOPICA [1] is a numerical suite conceived for prediction and analysis of plasma-facing antennas. It can handle real-life 3D antenna geometries (with housing, Faraday screen, etc.) as well as a realistic plasma model, including measured density and temperature profiles. TORIC [2] solves the finite Larmor radius wave equations in the ICRF regime in arbitrary axisymmetric toroidal plasmas. Due to the approach followed in developing TOPICA (i.e. the formal splitting of the problem in the vacuum region around the antenna and the plasma region inside the toroidal chamber), the code lends itself to handle toroidal plasmas, provided TORIC is run independently to yield the plasma surface admittance tensorsY (m,m',n). The latter enter directly into the integral equations solved by TOPICA, thus allowing a far more accurate plasma description that accounts for curvature effects. TOPICA outputs comprise, among others, the EM fields in front of the plasma: these can in turn be input to TORIC, in order to self-consistently determine the EM field propagation in the plasma. In this work, we report on the theory underlying the TOPICA/TORIC integration and the ongoing evolution of the two codes. [1] V. Lancellotti et al., Nucl. Fusion, 46 (2006) S476 [2] M. Brambilla, Plasma Phys. Contr. Fusion (1999) 41 1

Nuclear resonance tomography with a toroid cavity detector

DOEpatents

Woelk, K.; Rathke, J.W.; Klingler, R.J.

1996-11-12

A toroid cavity detection system is described for determining the spectral properties and distance from a fixed point for a sample using Nuclear Magnetic Resonance. The detection system consists of a toroid with a central conductor oriented along the main axis of the toroidal cylinder and perpendicular to a static uniform magnetic field oriented along the main axis of the toroid. An rf signal is input to the central conductor to produce a magnetic field perpendicular to the central axis of the toroid and whose field strength varies as the inverse of the radius of the toroid. The toroid cavity detection system can be used to encapsulate a sample, or the detection system can be perforated to allow a sample to flow into the detection device or to place the samples in specified sample tubes. The central conductor can also be coated to determine the spectral property of the coating and the coating thickness. The sample is then subjected to the respective magnetic fields and the responses measured to determine the desired properties. 4 figs.

Nuclear resonance tomography with a toroid cavity detector

DOEpatents

Woelk, Klaus; Rathke, Jerome W.; Klingler, Robert J.

1996-01-01

A toroid cavity detection system for determining the spectral properties and distance from a fixed point for a sample using Nuclear Magnetic Resonance. The detection system consists of a toroid with a central conductor oriented along the main axis of the toroidal cylinder and perpendicular to a static uniform magnetic field oriented along the main axis of the toroid. An rf signal is inputted to the central conductor to produce a magnetic field perpendicular to the central axis of the toroid and whose field strength varies as the inverse of the radius of the toroid. The toroid cavity detection system can be used to encapsulate a sample, or the detection system can be perforated to allow a sample to flow into the detection device or to place the samples in specified sample tubes. The central conductor can also be coated to determine the spectral property of the coating and the coating thickness. The sample is then subjected to the respective magnetic fields and the responses measured to determine the desired properties.

Final Technical Report for SBIR entitled Four-Dimensional Finite-Orbit-Width Fokker-Planck Code with Sources, for Neoclassical/Anomalous Transport Simulation of Ion and Electron Distributions

DOE Office of Scientific and Technical Information (OSTI.GOV)

Harvey, R. W.; Petrov, Yu. V.

2013-12-03

Within the US Department of Energy/Office of Fusion Energy magnetic fusion research program, there is an important whole-plasma-modeling need for a radio-frequency/neutral-beam-injection (RF/NBI) transport-oriented finite-difference Fokker-Planck (FP) code with combined capabilities for 4D (2R2V) geometry near the fusion plasma periphery, and computationally less demanding 3D (1R2V) bounce-averaged capabilities for plasma in the core of fusion devices. Demonstration of proof-of-principle achievement of this goal has been carried out in research carried out under Phase I of the SBIR award. Two DOE-sponsored codes, the CQL3D bounce-average Fokker-Planck code in which CompX has specialized, and the COGENT 4D, plasma edge-oriented Fokker-Planck code whichmore » has been constructed by Lawrence Livermore National Laboratory and Lawrence Berkeley Laboratory scientists, where coupled. Coupling was achieved by using CQL3D calculated velocity distributions including an energetic tail resulting from NBI, as boundary conditions for the COGENT code over the two-dimensional velocity space on a spatial interface (flux) surface at a given radius near the plasma periphery. The finite-orbit-width fast ions from the CQL3D distributions penetrated into the peripheral plasma modeled by the COGENT code. This combined code demonstrates the feasibility of the proposed 3D/4D code. By combining these codes, the greatest computational efficiency is achieved subject to present modeling needs in toroidally symmetric magnetic fusion devices. The more efficient 3D code can be used in its regions of applicability, coupled to the more computationally demanding 4D code in higher collisionality edge plasma regions where that extended capability is necessary for accurate representation of the plasma. More efficient code leads to greater use and utility of the model. An ancillary aim of the project is to make the combined 3D/4D code user friendly. Achievement of full-coupling of these two Fokker-Planck codes will advance computational modeling of plasma devices important to the USDOE magnetic fusion energy program, in particular the DIII-D tokamak at General Atomics, San Diego, the NSTX spherical tokamak at Princeton, New Jersey, and the MST reversed-field-pinch Madison, Wisconsin. The validation studies of the code against the experiments will improve understanding of physics important for magnetic fusion, and will increase our design capabilities for achieving the goals of the International Tokamak Experimental Reactor (ITER) project in which the US is a participant and which seeks to demonstrate at least a factor of five in fusion power production divided by input power.« less

Concept of a charged fusion product diagnostic for NSTX.

PubMed

Boeglin, W U; Valenzuela Perez, R; Darrow, D S

2010-10-01

The concept of a new diagnostic for NSTX to determine the time dependent charged fusion product emission profile using an array of semiconductor detectors is presented. The expected time resolution of 1-2 ms should make it possible to study the effect of magnetohydrodynamics and other plasma activities (toroidal Alfvén eigenmodes (TAE), neoclassical tearing modes (NTM), edge localized modes (ELM), etc.) on the radial transport of neutral beam ions. First simulation results of deuterium-deuterium (DD) fusion proton yields for different detector arrangements and methods for inverting the simulated data to obtain the emission profile are discussed.

Toroidal midplane neutral beam armor and plasma limiter

DOEpatents

Kugel, H.W.; Hand, S.W. Jr.; Ksayian, H.

1985-05-31

This invention contemplates an armor shield/plasma limiter positioned upon the inner wall of a toroidal vacuum chamber within which is magnetically confined an energetic plasma in a tokamak nuclear fusion reactor. The armor shield/plasma limiter is thus of a general semi-toroidal shape and is comprised of a plurality of adjacent graphite plates positioned immediately adjacent to each other so as to form a continuous ring upon and around the toroidal chamber's inner wall and the reactor's midplane coil. Each plate has a generally semi-circular outer circumference and a recessed inner portion and is comprised of upper and lower half sections positioned immediately adjacent to one another along the midplane of the plate. With the upper and lower half sections thus joined, a channel or duct is provided within the midplane of the plate in which a magnetic flux loop is positioned. The magnetic flux loop is thus positioned immediately adjacent to the fusing toroidal plasma and serves as a diagnostic sensor with the armor shield/plasma limiter minimizing the amount of power from the energetic plasma as well as from the neutral particle beams heating the plasma incident upon the flux loop.

Collisionality Scaling of Main-ion Toroidal and Poloidal Rotation in Low Torque DIII-D Plasmas

DOE Office of Scientific and Technical Information (OSTI.GOV)

B A Grierson, et al

In tokamak plasmas with low levels of toroidal rotation, the radial electric fi eld Er is a combination of pressure gradient and toroidal and poloidal rotation components, all having similar magnitudes. In order to assess the validity of neoclassical poloidal rotation theory for determining the poloidal rotation contribution to Er , Dα emission from neutral beam heated tokamak discharges in DIII-D [J.L. Luxon, Nucl. Fusion 42 , 614 (2002)] has been evaluated in a sequence of low torque (electron cyclotron resonance heating and balanced diagnostic neutral beam pulse) discharges to determine the local deuterium toroidal rotation velocity. By invoking themore » radial force balance relation the deuterium poloidal rotation can be inferred. It is found that the deuterium poloidal low exceeds the neoclassical value in plasmas with collisionality νi < 0: 1, being more ion diamagnetic, and with a stronger dependence on collisionality than neoclassical theory predicts. At low toroidal rotation, the poloidal rotation contribution to the radial electric fi eld and its shear is signi cant. The eff ect of anomalous levels of poloidal rotation on the radial electric fi eld and cross fi eld heat transport is investigated for ITER parameters.« less

Kinetic electromagnetic instabilities in an ITB plasma with weak magnetic shear

NASA Astrophysics Data System (ADS)

Chen, W.; Yu, D. L.; Ma, R. R.; Shi, P. W.; Li, Y. Y.; Shi, Z. B.; Du, H. R.; Ji, X. Q.; Jiang, M.; Yu, L. M.; Yuan, B. S.; Li, Y. G.; Yang, Z. C.; Zhong, W. L.; Qiu, Z. Y.; Ding, X. T.; Dong, J. Q.; Wang, Z. X.; Wei, H. L.; Cao, J. Y.; Song, S. D.; Song, X. M.; Liu, Yi.; Yang, Q. W.; Xu, M.; Duan, X. R.

2018-05-01

Kinetic Alfvén and pressure gradient driven instabilities are very common in magnetized plasmas, both in space and the laboratory. These instabilities will be easily excited by energetic particles (EPs) and/or pressure gradients in present-day fusion and future burning plasmas. This will not only cause the loss and redistribution of the EPs, but also affect plasma confinement and transport. Alfvénic ion temperature gradient (AITG) instabilities with the frequency ω_BAE<ω<ω_TAE and the toroidal mode numbers n=2{-}8 are found to be unstable in NBI internal transport barrier plasmas with weak shear and low pressure gradients, where ω_BAE and ω_TAE are the frequencies of the beta- and toroidicity-induced Alfvén eigenmodes, respectively. The measured results are consistent with the general fishbone-like dispersion relation and kinetic ballooning mode equation, and the modes become more unstable the smaller the magnetic shear is in low pressure gradient regions. The interaction between AITG activity and EPs also needs to be investigated with greater attention in fusion plasmas, such as ITER (Tomabechi and The ITER Team 1991 Nucl. Fusion 31 1135), since these fluctuations can be enhanced by weak magnetic shear and EPs.

Nondiffusive transport regimes for suprathermal ions in turbulent plasmas

NASA Astrophysics Data System (ADS)

Bovet, A.; Fasoli, A.; Ricci, P.; Furno, I.; Gustafson, K.

2015-04-01

The understanding of the transport of suprathermal ions in the presence of turbulence is important for fusion plasmas in the burning regime that will characterize reactors, and for space plasmas to understand the physics of particle acceleration. Here, three-dimensional measurements of a suprathermal ion beam in the toroidal plasma device TORPEX are presented. These measurements demonstrate, in a turbulent plasma, the existence of subdiffusive and superdiffusive transport of suprathermal ions, depending on their energy. This result stems from the unprecedented combination of uniquely resolved measurements and first-principles numerical simulations that reveal the mechanisms responsible for the nondiffusive transport. The transport regime is determined by the interaction of the suprathermal ion orbits with the turbulent plasma dynamics, and is strongly affected by the ratio of the suprathermal ion energy to the background plasma temperature.

Analytic expression for poloidal flow velocity in the banana regime

DOE Office of Scientific and Technical Information (OSTI.GOV)

Taguchi, M.

The poloidal flow velocity in the banana regime is calculated by improving the l = 1 approximation for the Fokker-Planck collision operator [M. Taguchi, Plasma Phys. Controlled Fusion 30, 1897 (1988)]. The obtained analytic expression for this flow, which can be used for general axisymmetric toroidal plasmas, agrees quite well with the recently calculated numerical results by Parker and Catto [Plasma Phys. Controlled Fusion 54, 085011 (2012)] in the full range of aspect ratio.

Moment free toroidal magnet

DOEpatents

Bonanos, Peter

1983-01-01

A toroidal magnet for confining a high magnetic field for use in fusion reactor research and nuclear particle detection. The magnet includes a series of conductor elements arranged about and fixed at its small major radius portion to the outer surface of a central cylindrical support each conductor element having a geometry such as to maintain the conductor elements in pure tension when a high current flows therein, and a support assembly which redistributes all or part of the tension which would otherwise arise in the small major radius portion of each coil element to the large major radius portion thereof.

Divertor for use in fusion reactors

DOEpatents

Christensen, Uffe R.

1979-01-01

A poloidal divertor for a toroidal plasma column ring having a set of poloidal coils co-axial with the plasma ring for providing a space for a thick shielding blanket close to the plasma along the entire length of the plasma ring cross section and all the way around the axis of rotation of the plasma ring. The poloidal coils of this invention also provide a stagnation point on the inside of the toroidal plasma column ring, gently curving field lines for vertical stability, an initial plasma current, and the shaping of the field lines of a separatrix up and around the shielding blanket.

Measurement of toroidal vessel eddy current during plasma disruption on J-TEXT.

PubMed

Liu, L J; Yu, K X; Zhang, M; Zhuang, G; Li, X; Yuan, T; Rao, B; Zhao, Q

2016-01-01

In this paper, we have employed a thin, printed circuit board eddy current array in order to determine the radial distribution of the azimuthal component of the eddy current density at the surface of a steel plate. The eddy current in the steel plate can be calculated by analytical methods under the simplifying assumptions that the steel plate is infinitely large and the exciting current is of uniform distribution. The measurement on the steel plate shows that this method has high spatial resolution. Then, we extended this methodology to a toroidal geometry with the objective of determining the poloidal distribution of the toroidal component of the eddy current density associated with plasma disruption in a fusion reactor called J-TEXT. The preliminary measured result is consistent with the analysis and calculation results on the J-TEXT vacuum vessel.

Plasma Properties of Microwave Produced Plasma in a Toroidal Device

NASA Astrophysics Data System (ADS)

Singh, Ajay; Edwards, W. F.; Held, Eric

2011-10-01

We have modified a small tokamak, STOR-1M, on loan from University of Saskatchewan, to operate as a low-temperature (~5 eV) toroidal plasma machine with externally induced toroidal magnetic fields ranging from zero to ~50 G. The plasma is produced using microwave discharges at relatively high pressures. Microwaves are produced by a kitchen microwave-oven magnetron operating at 2.45 GHz in continuous operating mode, resulting in pulses ~0.5 s in duration. Initial measurements of plasma formation in this device with and without applied magnetic fields are presented. Plasma density and temperature profiles have been measured using Langmuir probes and the magnetic field profile inside the plasma has been obtained using Hall probes. When the discharge is created with no applied toroidal magnetic field, the plasma does not fill the entire torus due to high background pressure. However, when a toroidal magnetic field is applied, the plasma flows along the applied field, filling the torus. Increasing the applied magnetic field seems to aid plasma formation - the peak density increases and the density gradient becomes steeper. Above a threshold magnetic field, the plasma develops low-frequency density oscillations due to probable excitation of flute modes in the plasma.

Microwave produced plasma in a Toroidal Device

NASA Astrophysics Data System (ADS)

Singh, A. K.; Edwards, W. F.; Held, E. D.

2010-11-01

A currentless toroidal plasma device exhibits a large range of interesting basic plasma physics phenomena. Such a device is not in equilibrium in a strict magneto hydrodynamic sense. There are many sources of free energy in the form of gradients in plasma density, temperature, the background magnetic field and the curvature of the magnetic field. These free energy sources excite waves and instabilities which have been the focus of studies in several devices in last two decades. A full understanding of these simple plasmas is far from complete. At Utah State University we have recently designed and installed a microwave plasma generation system on a small tokamak borrowed from the University of Saskatchewan, Saskatoon, Canada. Microwaves are generated at 2.45 GHz in a pulsed dc mode using a magnetron from a commercial kitchen microwave oven. The device is equipped with horizontal and vertical magnetic fields and a transformer to impose a toroidal electric field for current drive. Plasmas can be obtained over a wide range of pressure with and without magnetic fields. We present some preliminary measurements of plasma density and potential profiles. Measurements of plasma temperature at different operating conditions are also presented.

Self-organized criticality and the dynamics of near-marginal turbulent transport in magnetically confined fusion plasmas

NASA Astrophysics Data System (ADS)

Sanchez, R.; Newman, D. E.

2015-12-01

The high plasma temperatures expected at reactor conditions in magnetic confinement fusion toroidal devices suggest that near-marginal operation could be a reality in future devices and reactors. By near-marginal it is meant that the plasma profiles might wander around the local critical thresholds for the onset of instabilities. Self-organized criticality (SOC) was suggested in the mid 1990s as a more proper paradigm to describe the dynamics of tokamak plasma transport in near-marginal conditions. It advocated that, near marginality, the evolution of mean profiles and fluctuations should be considered simultaneously, in contrast to the more common view of a large separation of scales existing between them. Otherwise, intrinsic features of near-marginal transport would be missed, that are of importance to understand the properties of energy confinement. In the intervening 20 years, the relevance of the idea of SOC for near-marginal transport in fusion plasmas has transitioned from an initial excessive hype to the much more realistic standing of today, which we will attempt to examine critically in this review paper. First, the main theoretical ideas behind SOC will be described. Secondly, how they might relate to the dynamics of near-marginal transport in real magnetically confined plasmas will be discussed. Next, we will review what has been learnt about SOC from various numerical studies and what it has meant for the way in which we do numerical simulation of fusion plasmas today. Then, we will discuss the experimental evidence available from the several experiments that have looked for SOC dynamics in fusion plasmas. Finally, we will conclude by identifying the various problems that still remain open to investigation in this area. Special attention will be given to the discussion of frequent misconceptions and ongoing controversies. The review also contains a description of ongoing efforts that seek effective transport models better suited than traditional equations to capture SOC dynamics. Most of these models, based on the use of fractional transport equations and related concepts, could prove useful both in reactor operation and experiment control and design.

Dynamics of multiple double layers in high pressure glow discharge in a simple torus

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kumar Paul, Manash, E-mail: manashkr@gmail.com; Sharma, P. K.; Thakur, A.

2014-06-15

Parametric characterization of multiple double layers is done during high pressure glow discharge in a toroidal vessel of small aspect ratio. Although glow discharge (without magnetic field) is known to be independent of device geometry, but the toroidal boundary conditions are conducive to plasma growth and eventually the plasma occupy the toroidal volume partially. At higher anode potential, the visibly glowing spots on the body of spatially extended anode transform into multiple intensely luminous spherical plasma blob structures attached to the tip of the positive electrode. Dynamics of multiple double layers are observed in argon glow discharge plasma in presencemore » of toroidal magnetic field. The radial profiles of plasma parameters measured at various toroidal locations show signatures of double layer formation in our system. Parametric dependence of double layer dynamics in presence of toroidal magnetic field is presented here.« less

Close-out report with links to abstracts

DOE Office of Scientific and Technical Information (OSTI.GOV)

Marmar, Earl S.

This grant provided A/V support for two technical meetings of the Edge Coordinating Committee: (1) Nov 13, 2013 (co-located with the APS-DPP meeting in Denver, CO) https://ecc.mit.edu/fall-2013-technical-meeting#overlay-context=ecc-meetings; (2) April 28-May 1, 2015 (embedded sessions in the Transport Task Force Meeting, Salem, MA) http://www-internal.psfc.mit.edu/TTF2015/index.html. The ultimate goal of the U.S. Transport Task Force is to develop a physics-based understanding of particle, momentum and heat transport in magnetic fusion devices. This understanding should be of sufficient depth that it allows the development of predictive models of plasma transport that can be validated against experiment, and then used to anticipate the future performancemore » of burning plasmas in ITER, as well as to provide guidance for the design of next-step fusion nuclear science facilities. To achieve success in transport science, it is essential to characterize local fluctuations and transport in toroidal plasmas, to understand the basic mechanisms responsible for transport, and ultimately to control these transport processes. These goals must be pursued in multiple areas, and these topics evolve in order to reflect current interests.« less

Multi-device studies of pedestal physics and confinement in the I-mode regime

DOE PAGES

Hubbard, A. E.; Osborne, T.; Ryter, F.; ...

2016-07-05

This paper describes joint ITPA studies of the I-mode regime, which features an edge thermal barrier together with L-mode-like particle and impurity transport and no Edge Localized Modes (ELMs). The regime has been demonstrated on the Alcator C-Mod, ASDEX Upgrade and DIII-D tokamaks, over a wide range of device parameters and pedestal conditions. Dimensionless parameters at the pedestal show overlap across devices and extend to low collisionality. When they are matched, pedestal temperature profiles are also similar. Pedestals are stable to peeling ballooning modes, consistent with lack of ELMs. Access to Imode is independent of heating method (neutral beam injection,more » Ion Cyclotron and/or Electron Cyclotron Resonance Heating). Normalized energy confinement H 98,y2 ≥ 1 has been achieved for a range of 3 ≤ q 95 ≤ 4.9 and scales favourably with power. Changes in turbulence in the pedestal region accompany the transition from L-mode to I-mode. The L-I threshold increases with plasma density and current, and with device size, but has a weak dependence on toroidal magnetic field B T. The upper limit of power for I-modes, which is set by I-H transitions, increases with B T and the power range is largest on Alcator C-Mod at B > 5 T. Finally, issues for extrapolation to ITER and other future fusion devices are discussed.« less

Feedback stabilization of resistive wall modes in a reversed-field pinch

NASA Astrophysics Data System (ADS)

Brunsell, P. R.; Yadikin, D.; Gregoratto, D.; Paccagnella, R.; Liu, Y. Q.; Cecconello, M.; Drake, J. R.; Manduchi, G.; Marchiori, G.

2005-09-01

An array of saddle coils having Nc=16 equally spaced positions along the toroidal direction has been installed for feedback control of resistive wall modes (RWMs) on the EXTRAP T2R reversed-field pinch [P. R. Brunsell, H. Bergsaker, M. Cecconello et al., Plasma Phys. Controlled Fusion 43, 1457 (2001)]. Using feedback, multiple nonresonant RWMs are simultaneously suppressed for three to four wall times. Feedback stabilization of RWMs results in a significant prolongation of the discharge duration. This is linked to a better sustainment of the plasma and tearing mode toroidal rotation with feedback. Due to the limited number of coils in the toroidal direction, pairs of modes with toroidal mode numbers n ,n' that fulfill the condition ∣n-n'∣=Nc are coupled by the feedback action from the discrete coil array. With only one unstable mode in a pair of coupled modes, the suppression of the unstable mode is successful. If two modes are unstable in a coupled pair, two possibilities exist: partial suppression of both modes or, alternatively, complete stabilization of one target mode while the other is left unstable.

Anomalous-viscosity current drive

DOEpatents

Stix, T.H.; Ono, M.

1986-04-25

The present invention relates to a method and apparatus for maintaining a steady-state current for magnetically confining the plasma in a toroidal magnetic confinement device using anomalous viscosity current drive. A second aspect of this invention relates to an apparatus and method for the start-up of a magnetically confined toroidal plasma.

Role of poloidal flows on the particle confinement time in a simple toroidal device : an experimental study

NASA Astrophysics Data System (ADS)

Kumar, Umesh; Ganesh, R.; Saxena, Y. C.; Thatipamula, Shekar G.; Sathyanarayana, K.; Raju, Daniel

2017-10-01

In magnetized toroidal devices without rotational transform also known as Simple Magnetized Torus (SMT). The device BETA at the IPR is one such SMT with a major radius of 45 cm, minor radius of 15 cm and a maximum toroidal field of 0.1 Tesla. Understanding confinement in such helical configurations is an important problem both for fundamental plasma physics and for Tokamak edge physics. In a recent series of experiments it was demonstrated experimentally that the mean plasma profiles, fluctuation, flow and turbulence depend crucially on the parallel connection length, which was controlled by external vertical field. In the present work, we report our experimental findings, wherein we measure the particle confinement time for hot cathode discharge and ECRH discharge, with variation in parallel connection length. As ECRH plasma don't have mean electric field and hence the poloidal rotation of plasma is absent. However, in hot cathode discharge, there exist strong poloidal flows due to mean electric field. An experimental comparison of these along with theoretical model with variation in connection length will be presented. We also present experimental measurements of variation of plasma confinement time with mass as well as the ratio of vertical field to toroidal magnetic field.

Rf-assisted current startup in the Fusion Engineering Device (FED)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Borowski, S.K.; Peng, Y.K.M.; Kammash, T.

1982-08-01

Auxiliary rf heating of electrons before and during the current rise phase in the Fusion Engineering Device is examined as a means of reducing both the initiation loop voltage and resistive flux expenditure during startup. Prior to current initiation 1 to 2 MW of electron cyclotron resonance heating (ECRH) power at approx. 90 GHz is used to create a small volume of high conductivity plasma (T/sub e/ approx. = 100 eV, n/sub e/ approx. = 10/sup 13/ cm/sup -3/) near the upper hybrid resonance (UHR) region. This plasma conditioning permits a small radius (a/sub o/ approx. = 0.2 to 0.4more » m) current channel to be established with a relatively low initial loop voltage (less than or equal to 25 V). During the subsequent plasma expansion and current ramp phase, additional rf power is introduced to reduce volt-second consumption due to plasma resistance. A near classical particle and energy transport model has been developed to estimate the efficiency of electron heating in a currentless toroidal plasma. The model assumes that preferential electron heating at the UHR leads to the formation of an ambipolar sheath potential between the neutral plasma and the conducting vacuum vessel and limiter. The ambipolar electric field (E/sub AMB/) enables the plasma to neutralize itself via poloidal E vector/sub AMB/ x B vector drift. This form of effective rotational transform short-circuits the vertical charge separation and improves particle confinement.« less

Amphiphilic gold nanoparticles as modulators of lipid membrane fusion

NASA Astrophysics Data System (ADS)

Tahir, Mukarram; Alexander-Katz, Alfredo

The fusion of lipid membranes is central to biological functions like inter-cellular transport and signaling and is coordinated by proteins of the soluble N-ethylmaleimide-sensitive factor attachment protein receptors (SNAREs) superfamily. We utilize molecular dynamics simulations to demonstrate that gold nanoparticles functionalized with a mixed-monolayer of hydrophobic and hydrophilic alkanethiol ligands can act as synthetic analogues of these fusion proteins and mediate lipid membrane fusion by catalyzing the formation of a toroidal stalk between adjacent membranes and enabling the formation of a fusion pore upon influx of Ca2+ into the exterior solvent. The fusion pathway enabled by these synthetic nanostructures is analogous to the regulated fast fusion pathway observed during synaptic vesicle fusion; it therefore provides novel physical insights into this important biological process while also being relevant in a number of single-cell therapeutic applications. Computational resources from NSF XSEDE contract TG-DMR130042. Financial support from DOE CSGF fellowship DE-FG02-97ER25308.

Analysis of Alfvén eigenmode destabilization by energetic particles in Large Helical Device using a Landau-closure model

NASA Astrophysics Data System (ADS)

Varela, J.; Spong, D. A.; Garcia, L.

2017-04-01

Energetic particle populations in nuclear fusion experiments can destabilize the Alfvén Eigenmodes through inverse Landau damping and couplings with gap modes in the shear Alfvén continua. We use the reduced MHD equations to describe the linear evolution of the poloidal flux and the toroidal component of the vorticity in a full 3D system, coupled with equations of density and parallel velocity moments for the energetic particles. We add the Landau damping and resonant destabilization effects using a closure relation. We apply the model to study the Alfvén mode stability in the inward-shifted configurations of the Large Helical Device (LHD), performing a parametric analysis of the energetic particle β ({βf} ) in a range of realistic values, the ratios of the energetic particle thermal/Alfvén velocities ({{V}\\text{th}}/{{V}A0} ), the magnetic Lundquist numbers (S) and the toroidal modes (n). The n  =  1 and n  =  2 TAEs are destabilized, although the n  =  3 and n  =  4 TAEs are weakly perturbed. The most unstable configurations are associated with the density gradients of energetic particles in the plasma core: the TAEs are destabilized, even for small energetic particle populations, if their thermal velocity is lower than 0.4 times the Alfvén velocity. The frequency range of MHD bursts measured in the LHD are 50-70 kHz for the n  =  1 and 60-80 kHz for the n  =  2 TAE, which is consistent with the model predictions. ).

Avoidance of tearing mode locking with electro-magnetic torque introduced by feedback-based mode rotation control in DIII-D and RFX-mod

DOE PAGES

Okabayashi, M.; Zanca, P.; Strait, E. J.; ...

2016-11-25

Disruptions caused by tearing modes (TMs) are considered to be one of the most critical roadblocks to achieving reliable, steady-state operation of tokamak fusion reactors. We have demonstrated a promising scheme to avoid mode locking by utilizing the electro-magnetic (EM) torque produced with 3D coils that are available in many tokamaks. In this scheme, the EM torque is delivered to the modes by a toroidal phase shift between the externally applied field and the excited TM fields, compensating for the mode momentum loss through the interaction with the resistive wall and uncorrected error fields. Fine control of torque balance ismore » provided by a feedback scheme. We have explored this approach in two widely different devices and plasma conditions: DIII-D and RFX-mod operated in tokamak mode. In DIII-D, the plasma target was high β N in a non-circular divertor tokamak. We define β N as β N = β/(I p /aB t) (%Tm/MA), where β, I p, a, B t are the total stored plasma pressure normalized by the magnetic pressure, plasma current, plasma minor radius and toroidal magnetic field at the plasma center, respectively. The RFX-mod plasma was ohmically-heated with ultra-low safety factor in a circular limiter discharge with active feedback coils outside the thick resistive shell. The DIII-D and RFX-mod experiments showed remarkable consistency with theoretical predictions of torque balance. The application to ignition-oriented devices such as the International Thermonuclear Experimental Reactor (ITER) would expand the horizon of its operational regime. Finally, the internal 3D coil set currently under consideration for edge localized mode suppression in ITER would be well suited for this purpose.« less

Core transport properties in JT-60U and JET identity plasmas

NASA Astrophysics Data System (ADS)

Litaudon, X.; Sakamoto, Y.; de Vries, P. C.; Salmi, A.; Tala, T.; Angioni, C.; Benkadda, S.; Beurskens, M. N. A.; Bourdelle, C.; Brix, M.; Crombé, K.; Fujita, T.; Futatani, S.; Garbet, X.; Giroud, C.; Hawkes, N. C.; Hayashi, N.; Hoang, G. T.; Hogeweij, G. M. D.; Matsunaga, G.; Nakano, T.; Oyama, N.; Parail, V.; Shinohara, K.; Suzuki, T.; Takechi, M.; Takenaga, H.; Takizuka, T.; Urano, H.; Voitsekhovitch, I.; Yoshida, M.; ITPA Transport Group; JT-60 Team; EFDA contributors, JET

2011-07-01

The paper compares the transport properties of a set of dimensionless identity experiments performed between JET and JT-60U in the advanced tokamak regime with internal transport barrier, ITB. These International Tokamak Physics Activity, ITPA, joint experiments were carried out with the same plasma shape, toroidal magnetic field ripple and dimensionless profiles as close as possible during the ITB triggering phase in terms of safety factor, normalized Larmor radius, normalized collision frequency, thermal beta, ratio of ion to electron temperatures. Similarities in the ITB triggering mechanisms and sustainment were observed when a good match was achieved of the most relevant normalized profiles except the toroidal Mach number. Similar thermal ion transport levels in the two devices have been measured in either monotonic or non-monotonic q-profiles. In contrast, differences between JET and JT-60U were observed on the electron thermal and particle confinement in reversed magnetic shear configurations. It was found that the larger shear reversal in the very centre (inside normalized radius of 0.2) of JT-60U plasmas allowed the sustainment of stronger electron density ITBs compared with JET. As a consequence of peaked density profile, the core bootstrap current density is more than five times higher in JT-60U compared with JET. Thanks to the bootstrap effect and the slightly broader neutral beam deposition, reversed magnetic shear configurations are self-sustained in JT-60U scenarios. Analyses of similarities and differences between the two devices address key questions on the validity of the usual assumptions made in ITER steady scenario modelling, e.g. a flat density profile in the core with thermal transport barrier? Such assumptions have consequences on the prediction of fusion performance, bootstrap current and on the sustainment of the scenario.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Okabayashi, M.; Zanca, P.; Strait, E. J.

Disruptions caused by tearing modes (TMs) are considered to be one of the most critical roadblocks to achieving reliable, steady-state operation of tokamak fusion reactors. We have demonstrated a promising scheme to avoid mode locking by utilizing the electro-magnetic (EM) torque produced with 3D coils that are available in many tokamaks. In this scheme, the EM torque is delivered to the modes by a toroidal phase shift between the externally applied field and the excited TM fields, compensating for the mode momentum loss through the interaction with the resistive wall and uncorrected error fields. Fine control of torque balance ismore » provided by a feedback scheme. We have explored this approach in two widely different devices and plasma conditions: DIII-D and RFX-mod operated in tokamak mode. In DIII-D, the plasma target was high β N in a non-circular divertor tokamak. We define β N as β N = β/(I p /aB t) (%Tm/MA), where β, I p, a, B t are the total stored plasma pressure normalized by the magnetic pressure, plasma current, plasma minor radius and toroidal magnetic field at the plasma center, respectively. The RFX-mod plasma was ohmically-heated with ultra-low safety factor in a circular limiter discharge with active feedback coils outside the thick resistive shell. The DIII-D and RFX-mod experiments showed remarkable consistency with theoretical predictions of torque balance. The application to ignition-oriented devices such as the International Thermonuclear Experimental Reactor (ITER) would expand the horizon of its operational regime. Finally, the internal 3D coil set currently under consideration for edge localized mode suppression in ITER would be well suited for this purpose.« less

Avoidance of tearing mode locking with electro-magnetic torque introduced by feedback-based mode rotation control in DIII-D and RFX-mod

NASA Astrophysics Data System (ADS)

Okabayashi, M.; Zanca, P.; Strait, E. J.; Garofalo, A. M.; Hanson, J. M.; In, Y.; La Haye, R. J.; Marrelli, L.; Martin, P.; Paccagnella, R.; Paz-Soldan, C.; Piovesan, P.; Piron, C.; Piron, L.; Shiraki, D.; Volpe, F. A.; DIII-D, The; RFX-mod Teams

2017-01-01

Disruptions caused by tearing modes (TMs) are considered to be one of the most critical roadblocks to achieving reliable, steady-state operation of tokamak fusion reactors. Here we have demonstrated a promising scheme to avoid mode locking by utilizing the electro-magnetic (EM) torque produced with 3D coils that are available in many tokamaks. In this scheme, the EM torque is delivered to the modes by a toroidal phase shift between the externally applied field and the excited TM fields, compensating for the mode momentum loss through the interaction with the resistive wall and uncorrected error fields. Fine control of torque balance is provided by a feedback scheme. We have explored this approach in two widely different devices and plasma conditions: DIII-D and RFX-mod operated in tokamak mode. In DIII-D, the plasma target was high β N in a non-circular divertor tokamak. Here β N is defined as β N  =  β/(I p /aB t) (%Tm/MA), where β, I p, a, B t are the total stored plasma pressure normalized by the magnetic pressure, plasma current, plasma minor radius and toroidal magnetic field at the plasma center, respectively. The RFX-mod plasma was ohmically-heated with ultra-low safety factor in a circular limiter discharge with active feedback coils outside the thick resistive shell. The DIII-D and RFX-mod experiments showed remarkable consistency with theoretical predictions of torque balance. The application to ignition-oriented devices such as the International Thermonuclear Experimental Reactor (ITER) would expand the horizon of its operational regime. The internal 3D coil set currently under consideration for edge localized mode suppression in ITER would be well suited for this purpose.

Measurement of toroidal vessel eddy current during plasma disruption on J-TEXT

DOE Office of Scientific and Technical Information (OSTI.GOV)

Liu, L. J.; Yu, K. X.; Zhang, M., E-mail: zhangming@hust.edu.cn

2016-01-15

In this paper, we have employed a thin, printed circuit board eddy current array in order to determine the radial distribution of the azimuthal component of the eddy current density at the surface of a steel plate. The eddy current in the steel plate can be calculated by analytical methods under the simplifying assumptions that the steel plate is infinitely large and the exciting current is of uniform distribution. The measurement on the steel plate shows that this method has high spatial resolution. Then, we extended this methodology to a toroidal geometry with the objective of determining the poloidal distributionmore » of the toroidal component of the eddy current density associated with plasma disruption in a fusion reactor called J-TEXT. The preliminary measured result is consistent with the analysis and calculation results on the J-TEXT vacuum vessel.« less

Nearly axisymmetric hot plasmas in a highly rippled tokamak

NASA Astrophysics Data System (ADS)

Bellan, Paul

2002-11-01

Tokamak ohmic heating current flowing along toroidally rippled flux surfaces results in a poloidal torque. Since pressure gradients cannot offset torques, the torque drives plasma flows which convect plasma toroidally from ripple necks (high B_pol^2) to ripple bulges (low B_pol^2). Stagnation of the oppositely directed toroidal flows at the ripple bulges thermalizes the directed flow velocity ˜ B_pol/μ_0ρ , giving β _pol ˜1. These flows also convect frozen-in poloidal field lines which accumulate at the bulges enhancing the pinch force there and so reducing the bulge. Thus, a nearly axisymmetric β_pol ˜1 equilibrium is achieved using only a few TF coils. Particles bouncing in step between approaching flows will be Fermi accelerated to form a high energy tail. The ST tokamak magnetic mountain experiment [1] showed that, compared to a 1.8% ripple configuration, a 28% ripple configuration had four times the neutron production, and only a modest degradation of overall confinement; the former is consistent with the notion of Fermi acceleration of particles bouncing between colliding toroidal flows and the latter is consistent with ripple reduction due to toroidal convection of poloidal field lines. [1] W. Stodiek et al, Proc. 4th Intl. Conf. Plasma Phys. and Contr. Nuc. Fusion Res., (Madison, 1971), Vol. 1, p. 465

Modelling of plasma-wall interaction and impurity transport in fusion devices and prompt deposition of tungsten as application

NASA Astrophysics Data System (ADS)

Kirschner, A.; Tskhakaya, D.; Brezinsek, S.; Borodin, D.; Romazanov, J.; Ding, R.; Eksaeva, A.; Linsmeier, Ch

2018-01-01

Main processes of plasma-wall interaction and impurity transport in fusion devices and their impact on the availability of the devices are presented and modelling tools, in particular the three-dimensional Monte-Carlo code ERO, are introduced. The capability of ERO is demonstrated on the example of tungsten erosion and deposition modelling. The dependence of tungsten deposition on plasma temperature and density is studied by simulations with a simplified geometry assuming (almost) constant plasma parameters. The amount of deposition increases with increasing electron temperature and density. Up to 100% of eroded tungsten can be promptly deposited near to the location of erosion at very high densities (˜1 × 1014 cm-3 expected e.g. in the divertor of ITER). The effect of the sheath characteristics on tungsten prompt deposition is investigated by using particle-in-cell (PIC) simulations to spatially resolve the plasma parameters inside the sheath. Applying PIC data instead of non-resolved sheath leads in general to smaller tungsten deposition, which is mainly due to a density and temperature decrease towards the surface within the sheath. Two-dimensional tungsten erosion/deposition simulations, assuming symmetry in toroidal direction but poloidally spatially varying plasma parameter profiles, have been carried out for the JET divertor. The simulations reveal, similar to experimental findings, that tungsten gross erosion is dominated in H-mode plasmas by the intra-ELM phases. However, due to deposition, the net tungsten erosion can be similar within intra- and inter-ELM phases if the inter-ELM electron temperature is high enough. Also, the simulated deposition fraction of about 84% in between ELMs is in line with spectroscopic observations from which a lower limit of 50% has been estimated.

Results of subscale MTF compression experiments

NASA Astrophysics Data System (ADS)

Howard, Stephen; Mossman, A.; Donaldson, M.; Fusion Team, General

2016-10-01

In magnetized target fusion (MTF) a magnetized plasma torus is compressed in a time shorter than its own energy confinement time, thereby heating to fusion conditions. Understanding plasma behavior and scaling laws is needed to advance toward a reactor-scale demonstration. General Fusion is conducting a sequence of subscale experiments of compact toroid (CT) plasmas being compressed by chemically driven implosion of an aluminum liner, providing data on several key questions. CT plasmas are formed by a coaxial Marshall gun, with magnetic fields supported by internal plasma currents and eddy currents in the wall. Configurations that have been compressed so far include decaying and sustained spheromaks and an ST that is formed into a pre-existing toroidal field. Diagnostics measure B, ne, visible and x-ray emission, Ti and Te. Before compression the CT has an energy of 10kJ magnetic, 1 kJ thermal, with Te of 100 - 200 eV, ne 5x1020 m-3. Plasma was stable during a compression factor R0/R >3 on best shots. A reactor scale demonstration would require 10x higher initial B and ne but similar Te. Liner improvements have minimized ripple, tearing and ejection of micro-debris. Plasma facing surfaces have included plasma-sprayed tungsten, bare Cu and Al, and gettering with Ti and Li.

High-performance superconductors for Fusion Nuclear Science Facility

DOE PAGES

Zhai, Yuhu; Kessel, Chuck; Barth, Christian; ...

2016-11-09

High-performance superconducting magnets play an important role in the design of the next step large-scale, high-field fusion reactors such as the fusion nuclear science facility (FNSF) and the spherical tokamak (ST) pilot plant beyond ITER. Here, Princeton Plasma Physics Laboratory is currently leading the design studies of the FNSF and the ST pilot plant study. ITER, which is under construction in the south of France, utilizes the state-of-the-art low temperature superconducting magnet technology based on the cable-in-conduit conductor design, where over a thousand multifilament Nb 3Sn superconducting strands are twisted together to form a high-current-carrying cable inserted into a steelmore » jacket for coil windings. We present design options of the high-performance superconductors in the winding pack for the FNSF toroidal field magnet system based on the toroidal field radial build from the system code. For the low temperature superconductor options, the advanced J cNb 3Sn RRP strands (J c > 1000 A/mm 2 at 16 T, 4 K) from Oxford Superconducting Technology are under consideration. For the high-temperature superconductor options, the rectangular-shaped high-current HTS cable made of stacked YBCO tapes will be considered to validate feasibility of TF coil winding pack design for the ST-FNSF magnets.« less

High-performance superconductors for Fusion Nuclear Science Facility

DOE Office of Scientific and Technical Information (OSTI.GOV)

Zhai, Yuhu; Kessel, Chuck; Barth, Christian

High-performance superconducting magnets play an important role in the design of the next step large-scale, high-field fusion reactors such as the fusion nuclear science facility (FNSF) and the spherical tokamak (ST) pilot plant beyond ITER. Here, Princeton Plasma Physics Laboratory is currently leading the design studies of the FNSF and the ST pilot plant study. ITER, which is under construction in the south of France, utilizes the state-of-the-art low temperature superconducting magnet technology based on the cable-in-conduit conductor design, where over a thousand multifilament Nb 3Sn superconducting strands are twisted together to form a high-current-carrying cable inserted into a steelmore » jacket for coil windings. We present design options of the high-performance superconductors in the winding pack for the FNSF toroidal field magnet system based on the toroidal field radial build from the system code. For the low temperature superconductor options, the advanced J cNb 3Sn RRP strands (J c > 1000 A/mm 2 at 16 T, 4 K) from Oxford Superconducting Technology are under consideration. For the high-temperature superconductor options, the rectangular-shaped high-current HTS cable made of stacked YBCO tapes will be considered to validate feasibility of TF coil winding pack design for the ST-FNSF magnets.« less

Suppressed ion-scale turbulence in a hot high-β plasma

NASA Astrophysics Data System (ADS)

Schmitz, L.; Fulton, D. P.; Ruskov, E.; Lau, C.; Deng, B. H.; Tajima, T.; Binderbauer, M. W.; Holod, I.; Lin, Z.; Gota, H.; Tuszewski, M.; Dettrick, S. A.; Steinhauer, L. C.

2016-12-01

An economic magnetic fusion reactor favours a high ratio of plasma kinetic pressure to magnetic pressure in a well-confined, hot plasma with low thermal losses across the confining magnetic field. Field-reversed configuration (FRC) plasmas are potentially attractive as a reactor concept, achieving high plasma pressure in a simple axisymmetric geometry. Here, we show that FRC plasmas have unique, beneficial microstability properties that differ from typical regimes in toroidal confinement devices. Ion-scale fluctuations are found to be absent or strongly suppressed in the plasma core, mainly due to the large FRC ion orbits, resulting in near-classical thermal ion confinement. In the surrounding boundary layer plasma, ion- and electron-scale turbulence is observed once a critical pressure gradient is exceeded. The critical gradient increases in the presence of sheared plasma flow induced via electrostatic biasing, opening the prospect of active boundary and transport control in view of reactor requirements.

Suppressed ion-scale turbulence in a hot high-β plasma

PubMed Central

Schmitz, L.; Fulton, D. P.; Ruskov, E.; Lau, C.; Deng, B. H.; Tajima, T.; Binderbauer, M. W.; Holod, I.; Lin, Z.; Gota, H.; Tuszewski, M.; Dettrick, S. A.; Steinhauer, L. C.

2016-01-01

An economic magnetic fusion reactor favours a high ratio of plasma kinetic pressure to magnetic pressure in a well-confined, hot plasma with low thermal losses across the confining magnetic field. Field-reversed configuration (FRC) plasmas are potentially attractive as a reactor concept, achieving high plasma pressure in a simple axisymmetric geometry. Here, we show that FRC plasmas have unique, beneficial microstability properties that differ from typical regimes in toroidal confinement devices. Ion-scale fluctuations are found to be absent or strongly suppressed in the plasma core, mainly due to the large FRC ion orbits, resulting in near-classical thermal ion confinement. In the surrounding boundary layer plasma, ion- and electron-scale turbulence is observed once a critical pressure gradient is exceeded. The critical gradient increases in the presence of sheared plasma flow induced via electrostatic biasing, opening the prospect of active boundary and transport control in view of reactor requirements. PMID:28000675

Development of Numerical Methods to Estimate the Ohmic Breakdown Scenarios of a Tokamak

NASA Astrophysics Data System (ADS)

Yoo, Min-Gu; Kim, Jayhyun; An, Younghwa; Hwang, Yong-Seok; Shim, Seung Bo; Lee, Hae June; Na, Yong-Su

2011-10-01

The ohmic breakdown is a fundamental method to initiate the plasma in a tokamak. For the robust breakdown, ohmic breakdown scenarios have to be carefully designed by optimizing the magnetic field configurations to minimize the stray magnetic fields. This research focuses on development of numerical methods to estimate the ohmic breakdown scenarios by precise analysis of the magnetic field configurations. This is essential for the robust and optimal breakdown and start-up of fusion devices especially for ITER and its beyond equipped with low toroidal electric field (ET <= 0.3 V/m). A field-line-following analysis code based on the Townsend avalanche theory and a particle simulation code are developed to analyze the breakdown characteristics of actual complex magnetic field configurations including the stray magnetic fields in tokamaks. They are applied to the ohmic breakdown scenarios of tokamaks such as KSTAR and VEST and compared with experiments.

Gyrokinetic modeling of impurity peaking in JET H-mode plasmas

NASA Astrophysics Data System (ADS)

Manas, P.; Camenen, Y.; Benkadda, S.; Weisen, H.; Angioni, C.; Casson, F. J.; Giroud, C.; Gelfusa, M.; Maslov, M.

2017-06-01

Quantitative comparisons are presented between gyrokinetic simulations and experimental values of the carbon impurity peaking factor in a database of JET H-modes during the carbon wall era. These plasmas feature strong NBI heating and hence high values of toroidal rotation and corresponding gradient. Furthermore, the carbon profiles present particularly interesting shapes for fusion devices, i.e., hollow in the core and peaked near the edge. Dependencies of the experimental carbon peaking factor ( R / L nC ) on plasma parameters are investigated via multilinear regressions. A marked correlation between R / L nC and the normalised toroidal rotation gradient is observed in the core, which suggests an important role of the rotation in establishing hollow carbon profiles. The carbon peaking factor is then computed with the gyrokinetic code GKW, using a quasi-linear approach, supported by a few non-linear simulations. The comparison of the quasi-linear predictions to the experimental values at mid-radius reveals two main regimes. At low normalised collisionality, ν * , and T e / T i < 1 , the gyrokinetic simulations quantitatively recover experimental carbon density profiles, provided that rotodiffusion is taken into account. In contrast, at higher ν * and T e / T i > 1 , the very hollow experimental carbon density profiles are never predicted by the simulations and the carbon density peaking is systematically over estimated. This points to a possible missing ingredient in this regime.

Ideal relaxation of the Hopf fibration

NASA Astrophysics Data System (ADS)

Smiet, Christopher Berg; Candelaresi, Simon; Bouwmeester, Dirk

2017-07-01

Ideal magnetohydrodynamics relaxation is the topology-conserving reconfiguration of a magnetic field into a lower energy state where the net force is zero. This is achieved by modeling the plasma as perfectly conducting viscous fluid. It is an important tool for investigating plasma equilibria and is often used to study the magnetic configurations in fusion devices and astrophysical plasmas. We study the equilibrium reached by a localized magnetic field through the topology conserving relaxation of a magnetic field based on the Hopf fibration in which magnetic field lines are closed circles that are all linked with one another. Magnetic fields with this topology have recently been shown to occur in non-ideal numerical simulations. Our results show that any localized field can only attain equilibrium if there is a finite external pressure, and that for such a field a Taylor state is unattainable. We find an equilibrium plasma configuration that is characterized by a lowered pressure in a toroidal region, with field lines lying on surfaces of constant pressure. Therefore, the field is in a Grad-Shafranov equilibrium. Localized helical magnetic fields are found when plasma is ejected from astrophysical bodies and subsequently relaxes against the background plasma, as well as on earth in plasmoids generated by, e.g., a Marshall gun. This work shows under which conditions an equilibrium can be reached and identifies a toroidal depression as the characteristic feature of such a configuration.

High-beta spherical tokamak startup in TS-4 merging experiment by use of toroidal field ramp-up

NASA Astrophysics Data System (ADS)

Kaminou, Yasuhiro; , Toru, II; Kato, Joji; Inomoto, Michiaki; Ono, Yasushi; TS Group Team; National InstituteFusion Science Collaboration

2014-10-01

We demonstrated the formation method of an ultrahigh-beta spherical tokamak by use of a field-reversed configuration and a spheromak in TS-4 device (R ~ 0.5 m, A ~ 1.5, Ip ~ 30-100 kA, B ~ 100 mT). This method is composed of the following steps: 1. Two spheromaks are merged together and a high-beta spheromak or FRC is formed by reconnection heating. 2. External toroidal magnetic field is added (current rising time ~50 μs), and spherical tokamak-like configuration is formed. In this way, the ultrahigh-beta ST is formed. The ultrahigh-beta ST formed by FRC has a diamagnetic toroidal field, and it presumed to be in a second-stable state for ballooning stability, and the one formed by spheromak has a weak paramagnetic toroidal magnetic field, while a spheormak has a strong paramagnetic toroidal magnetic field. This diamagnetic current derives from inductive electric field by ramping up the external toroidal magnetic field, and the diamagnetic current sustains high thermal pressure of the ultrahigh-beta spherical tokamak. And the beta of the ultrahigh-beta ST formed by FRC reaches about 50%. To sustain the high-beta state, 0.6 MW neutral beam injection and center solenoid coils are installed to the TS-4 device. In the poster, we report the experimental results of ultrahigh-beta spherical tokamak startup and sustainment by NBI and CS current driving experiment.

The NASA-Lewis program on fusion energy for space power and propulsion, 1958-1978

NASA Technical Reports Server (NTRS)

Schulze, Norman R.; Roth, J. Reece

1990-01-01

An historical synopsis is provided of the NASA-Lewis research program on fusion energy for space power and propulsion systems. It was initiated to explore the potential applications of fusion energy to space power and propulsion systems. Some fusion related accomplishments and program areas covered include: basic research on the Electric Field Bumpy Torus (EFBT) magnetoelectric fusion containment concept, including identification of its radial transport mechanism and confinement time scaling; operation of the Pilot Rig mirror machine, the first superconducting magnet facility to be used in plasma physics or fusion research; operation of the Superconducting Bumpy Torus magnet facility, first used to generate a toroidal magnetic field; steady state production of neutrons from DD reactions; studies of the direct conversion of plasma enthalpy to thrust by a direct fusion rocket via propellant addition and magnetic nozzles; power and propulsion system studies, including D(3)He power balance, neutron shielding, and refrigeration requirements; and development of large volume, high field superconducting and cryogenic magnet technology.

The design of a low-cost Thomson Scattering system for use on the ORNL PhIX device

NASA Astrophysics Data System (ADS)

Biewer, T. M.; Lore, J.; Goulding, R. H.; Hillis, D. L.; Owen, L.; Rapp, J.

2012-10-01

Study of the plasma-material interface (PMI) under high power and particle flux on linear plasma devices is an active area of research that is relevant to fusion-grade toroidal devices such as ITER and DEMO. ORNL is assembling a 15 cm diameter, ˜3 m long linear machine, called the Physics Integration eXperiment (PhIX), which incorporates a helicon plasma source, electron heating, and a material target. The helicon source has demonstrated coupling of up to 100 kW of rf power, and produced ne >= 4 x 10^19 m-3 in D, and He fueled plasmas, measured with interferometry and Langmuir probes (LP). Optical emission spectroscopy was used to confirm LP measurements that Te is about 10 eV in helicon heated plasmas, which will presumably increase when electron heating is applied. Plasma parameters (ne, Te, n0) of the PhIX device will be measured with a novel, low-cost Thomson Scattering (TS) system. The data will be used to characterize the PMI regime with multiple profile measurements in front of the target. Profiles near the source and target will be used to determine the parallel transport regime via comparison to 2D fluid plasma simulations. This work was supported by the US. D.O.E. contract DE-AC05-00OR22725.

Electron cyclotron emission imaging and applications in magnetic fusion energy

NASA Astrophysics Data System (ADS)

Tobias, Benjamin John

Energy production through the burning of fossil fuels is an unsustainable practice. Exponentially increasing energy consumption and dwindling natural resources ensure that coal and gas fueled power plants will someday be a thing of the past. However, even before fuel reserves are depleted, our planet may well succumb to disastrous side effects, namely the build up of carbon emissions in the environment triggering world-wide climate change and the countless industrial spills of pollutants that continue to this day. Many alternatives are currently being developed, but none has so much promise as fusion nuclear energy, the energy of the sun. The confinement of hot plasma at temperatures in excess of 100 million Kelvin by a carefully arranged magnetic field for the realization of a self-sustaining fusion power plant requires new technologies and improved understanding of fundamental physical phenomena. Imaging of electron cyclotron radiation lends insight into the spatial and temporal behavior of electron temperature fluctuations and instabilities, providing a powerful diagnostic for investigations into basic plasma physics and nuclear fusion reactor operation. This dissertation presents the design and implementation of a new generation of Electron Cyclotron Emission Imaging (ECEI) diagnostics on toroidal magnetic fusion confinement devices, or tokamaks, around the world. The underlying physics of cyclotron radiation in fusion plasmas is reviewed, and a thorough discussion of millimeter wave imaging techniques and heterodyne radiometry in ECEI follows. The imaging of turbulence and fluid flows has evolved over half a millennium since Leonardo da Vinci's first sketches of cascading water, and applications for ECEI in fusion research are broad ranging. Two areas of physical investigation are discussed in this dissertation: the identification of poloidal shearing in Alfven eigenmode structures predicted by hybrid gyrofluid-magnetohydrodynamic (gyrofluid-MHD) modeling, and magnetic field line displacement during precursor oscillations associated with the sawtooth crash, a disruptive instability observed both in tokamak plasmas with high core current and in the magnetized plasmas of solar flares and other interstellar plasmas. Understanding both of these phenomena is essential for the future of magnetic fusion energy, and important new observations described herein underscore the advantages of imaging techniques in experimental physics.

Direct current stimulation of titanium interbody fusion devices in primates.

PubMed

Cook, Stephen D; Patron, Laura P; Christakis, Petros M; Bailey, Kirk J; Banta, Charles; Glazer, Paul A

2004-01-01

The fusion rate for anterior lumbar interbody fusion (ALIF) varies widely with the use of different interbody devices and bone graft options. Adjunctive techniques such as electrical stimulation may improve the rate of bony fusion. To determine if direct current (DC) electrical stimulation of a metallic interbody fusion device enhanced the incidence or extent of anterior bony fusion. ALIF was performed using titanium alloy interbody fusion devices with and without adjunctive DC electrical stimulation in nonhuman primates. ALIF was performed through an anterolateral approach in 35 macaques with autogenous bone graft and either a titanium alloy (Ti-6Al-4V) fusion device or femoral allograft ring. The fusion devices of 19 animals received high (current density 19.6 microA/cm2) or low (current density 5.4 microA/cm2) DC electrical stimulation using an implanted generator for a 12- or 26-week evaluation period. Fusion sites were studied using serial radiographs, computed tomography imaging, nondestructive mechanical testing and qualitative and semiquantitative histology. Fusion was achieved with the titanium fusion device and autogenous bone graft. At 12 weeks, the graft was consolidating and early to moderate bridging callus was observed in and around the device. By 26 weeks, the anterior callus formation was more advanced with increased evidence of bridging trabeculations and early bone remodeling. The callus formation was not as advanced or abundant for the allograft ring group. Histology revealed the spinal fusion device had an 86% incidence of bony fusion at 26 weeks compared with a 50% fusion rate for the allograft rings. DC electrical stimulation of the fusion device had a positive effect on anterior interbody fusion by increasing both the presence and extent of bony fusion in a current density-dependent manner. Adjunctive DC electrical stimulation of the fusion device improved the rate and extent of bony fusion compared with a nonstimulated device. The fusion device was equivalent to or better than the femoral allograft ring in all evaluations. The use of adjunctive direct current electrical stimulation may provide a means of improving anterior interbody fusion.

Physics Basis for the Advanced Tokamak Fusion Power Plant ARIES-AT

DOE Office of Scientific and Technical Information (OSTI.GOV)

S.C. Jardin; C.E. Kessel; T.K. Mau

2003-10-07

The advanced tokamak is considered as the basis for a fusion power plant. The ARIES-AT design has an aspect ratio of A always equal to R/a = 4.0, an elongation and triangularity of kappa = 2.20, delta = 0.90 (evaluated at the separatrix surface), a toroidal beta of beta = 9.1% (normalized to the vacuum toroidal field at the plasma center), which corresponds to a normalized beta of bN * 100 x b/(I(sub)P(MA)/a(m)B(T)) = 5.4. These beta values are chosen to be 10% below the ideal-MHD stability limit. The bootstrap-current fraction is fBS * I(sub)BS/I(sub)P = 0.91. This leads tomore » a design with total plasma current I(sub)P = 12.8 MA, and toroidal field of 11.1 T (at the coil edge) and 5.8 T (at the plasma center). The major and minor radii are 5.2 and 1.3 m, respectively. The effects of H-mode edge gradients and the stability of this configuration to non-ideal modes is analyzed. The current-drive system consists of ICRF/FW for on-axis current drive and a lower-hybrid system for off-axis. Tran sport projections are presented using the drift-wave based GLF23 model. The approach to power and particle exhaust using both plasma core and scrape-off-layer radiation is presented.« less

Kinetic description of rotating Tokamak plasmas with anisotropic temperatures in the collisionless regime

NASA Astrophysics Data System (ADS)

Cremaschini, Claudio; Tessarotto, Massimo

2011-11-01

A largely unsolved theoretical issue in controlled fusion research is the consistent kinetic treatment of slowly-time varying plasma states occurring in collisionless and magnetized axisymmetric plasmas. The phenomenology may include finite pressure anisotropies as well as strong toroidal and poloidal differential rotation, characteristic of Tokamak plasmas. Despite the fact that physical phenomena occurring in fusion plasmas depend fundamentally on the microscopic particle phase-space dynamics, their consistent kinetic treatment remains still essentially unchallenged to date. The goal of this paper is to address the problem within the framework of Vlasov-Maxwell description. The gyrokinetic treatment of charged particles dynamics is adopted for the construction of asymptotic solutions for the quasi-stationary species kinetic distribution functions. These are expressed in terms of the particle exact and adiabatic invariants. The theory relies on a perturbative approach, which permits to construct asymptotic analytical solutions of the Vlasov-Maxwell system. In this way, both diamagnetic and energy corrections are included consistently into the theory. In particular, by imposing suitable kinetic constraints, the existence of generalized bi-Maxwellian asymptotic kinetic equilibria is pointed out. The theory applies for toroidal rotation velocity of the order of the ion thermal speed. These solutions satisfy identically also the constraints imposed by the Maxwell equations, i.e., quasi-neutrality and Ampere's law. As a result, it is shown that, in the presence of nonuniform fluid and EM fields, these kinetic equilibria can sustain simultaneously toroidal differential rotation, quasi-stationary finite poloidal flows and temperature anisotropy.

Evaluation of the toroidal torque driven by external non-resonant non-axisymmetric magnetic field perturbations in a tokamak

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kasilov, Sergei V.; Institute of Plasma Physics National Science Center “Kharkov Institute of Physics and Technology” ul. Akademicheskaya 1, 61108 Kharkov; Kernbichler, Winfried

2014-09-15

The toroidal torque driven by external non-resonant magnetic perturbations (neoclassical toroidal viscosity) is an important momentum source affecting the toroidal plasma rotation in tokamaks. The well-known force-flux relation directly links this torque to the non-ambipolar neoclassical particle fluxes arising due to the violation of the toroidal symmetry of the magnetic field. Here, a quasilinear approach for the numerical computation of these fluxes is described, which reduces the dimension of a standard neoclassical transport problem by one without model simplifications of the linearized drift kinetic equation. The only limiting condition is that the non-axisymmetric perturbation field is small enough such thatmore » the effect of the perturbation field on particle motion within the flux surface is negligible. Therefore, in addition to most of the transport regimes described by the banana (bounce averaged) kinetic equation also such regimes as, e.g., ripple-plateau and resonant diffusion regimes are naturally included in this approach. Based on this approach, a quasilinear version of the code NEO-2 [W. Kernbichler et al., Plasma Fusion Res. 3, S1061 (2008).] has been developed and benchmarked against a few analytical and numerical models. Results from NEO-2 stay in good agreement with results from these models in their pertinent range of validity.« less

Microinstability properties of negative magnetic shear discharges in the Tokamak Fusion Test Reactor and DIII-D

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rewoldt, G.; Tang, W.M.; Lao, L.L.

1997-03-01

The microinstability properties of discharges with negative (reversed) magnetic shear in the Tokamak Fusion Test Reactor (TFTR) and DIII-D experiments with and without confinement transitions are investigated. A comprehensive kinetic linear eigenmode calculation employing the ballooning representation is employed with experimentally measured profile data, and using the corresponding numerically computed magnetohydrodynamic (MHD) equilibria. The instability considered is the toroidal drift mode (trapped-electron-{eta}{sub i} mode). A variety of physical effects associated with differing q-profiles are explained. In addition, different negative magnetic shear discharges at different times in the discharge for TFTR and DIII-D are analyzed. The effects of sheared toroidal rotation,more » using data from direct spectroscopic measurements for carbon, are analyzed using comparisons with results from a two-dimensional calculation. Comparisons are also made for nonlinear stabilization associated with shear in E{sub r}/RB{sub {theta}}. The relative importance of changes in different profiles (density, temperature, q, rotation, etc.) on the linear growth rates is considered.« less

Bidirectional tornado modes on the Joint European Torus

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sandquist, P.; Sharapov, S. E.; Lisak, M.

In discharges on the Joint European Torus [P. H. Rebut and B. E. Keen, Fusion Technol. 11, 13 (1987)] with safety factor q(0)<1 and high-power ion cyclotron resonance heating (ICRH), monster sawtooth crashes are preceded by frequency sweeping 'tornado modes' in the toroidal Alfven eigenmode frequency range. A suite of equilibrium and spectral magnetohydrodynamical codes is used for explaining the observed evolution of the tornado mode frequency and for identifying temporal evolution of the safety factor inside the q=1 radius just before sawtooth crashes. In some cases, the tornado modes are observed simultaneously with both positive and negative toroidal modemore » numbers. Hence, a free energy source other than the radial gradient of the energetic ion pressure exciting these modes is sought. The distribution function of the ICRH-accelerated ions is assessed with the SELFO code [J. Hedin et al., Nucl. Fusion 42, 527 (2002)] and energetic particle drive due to the velocity space anisotropy of ICRH-accelerated ions is considered analytically as the possible source for excitation of bidirectional tornado modes.« less

An Imposed Dynamo Current Drive Experiment: Demonstration of Confinement

NASA Astrophysics Data System (ADS)

Jarboe, Thomas; Hansen, Chris; Hossack, Aaron; Marklin, George; Morgan, Kyle; Nelson, Brian; Sutherland, Derek; Victor, Brian

2014-10-01

An experiment for studying and developing the efficient sustainment of a spheromak with sufficient confinement (current-drive power heats the plasma to its stability β-limit) and in the keV temperature range is discussed. A high- β spheromak sustained by imposed dynamo current drive (IDCD) is justified because: previous transient experiments showed sufficient confinement in the keV range with no external toroidal field coil; recent results on HIT-SI show sustainment with sufficient confinement at low temperature; the potential of IDCD of solving other fusion issues; a very attractive reactor concept; and the general need for efficient current drive in magnetic fusion. The design of a 0.55 m minor radius machine with the required density control, wall loading, and neutral shielding for a 2 s pulse is presented. Peak temperatures of 1 keV and toroidal currents of 1.35 MA and 16% wall-normalized plasma beta are envisioned. The experiment is large enough to address the key issues yet small enough for rapid modification and for extended MHD modeling of startup and code validation.

Comparison of Observed Toroidal Rotation with Neoclassical Transport Theory

NASA Astrophysics Data System (ADS)

Wong, S. K.; Chan, V. S.; Hinton, F. L.

2000-10-01

Toroidal rotations have been observed in Ohmic and ICRF discharges(J.E. Rice et al.), Nucl. Fusion 39 (1999) 1175. which have little overall momentum input. They are found to correlate with the thermal energy content and the magnitude of the plasma current and change sign relative to the plasma current in different conditions. Existing comparisons with neoclassical transport theory either focus on the relation of the rotation with the radial electric field or fail to use the full expression of the angular momentum flux. We seek to remedy this by invoking the correct expressions(M.N. Rosenbluth et al.), Plasma Phys. Contr. Nucl. Fusion Research (IAEA, Vienna, 1971), Vol. 1, p. 495.^,(R.D. Hazeltine, Phys. Fluids 17) (1974) 961.^,(F.L. Hinton and S.K. Wong, Phys. Fluids 28) (1985) 3082. which contain both diffusive and non-diffusive terms. Developmental work is performed to consider such issues as the presence of impurity ions, the occurrence of near-sonic flows, and the lack of up-down symmetry of flux surfaces. Comparison with experiments will be presented.

Disruption forces on the tokamak wall with and without poloidal currents

NASA Astrophysics Data System (ADS)

Pustovitov, V. D.

2017-05-01

The contributions into the disruption radial force on the tokamak vacuum vessel wall are calculated and analyzed. One is due to the induced toroidal current in the wall, and another is due to the poloidal current. The latter is not accounted for in the models that represent the wall as a set of isolated toroidal filaments. It is shown that such modeling must lead to significant errors in the evaluation of the force during either thermal or current quench. The analytical derivations are performed here for an arbitrary tokamak configuration with final estimates for a circular large-aspect-ratio plasma and a coaxial wall reacting on perturbations as a perfect conductor. The results are compared with those recently obtained numerically by the codes DINA, MAXFEA and CarMa0NL. The discrepancies between the DINA simulations (Khayrutdinov et al 2016 Plasma Phys. Control. Fusion 58 115012) and earlier analytical predictions are explained. The recent conclusion (Villone et al 2015 Fusion Eng. Des. 93 57) on the role of the disruption-induced poloidal current in the wall is confirmed and extended to a wider area.

Status and problems of fusion reactor development.

PubMed

Schumacher, U

2001-03-01

Thermonuclear fusion of deuterium and tritium constitutes an enormous potential for a safe, environmentally compatible and sustainable energy supply. The fuel source is practically inexhaustible. Further, the safety prospects of a fusion reactor are quite favourable due to the inherently self-limiting fusion process, the limited radiologic toxicity and the passive cooling property. Among a small number of approaches, the concept of toroidal magnetic confinement of fusion plasmas has achieved most impressive scientific and technical progress towards energy release by thermonuclear burn of deuterium-tritium fuels. The status of thermonuclear fusion research activity world-wide is reviewed and present solutions to the complicated physical and technological problems are presented. These problems comprise plasma heating, confinement and exhaust of energy and particles, plasma stability, alpha particle heating, fusion reactor materials, reactor safety and environmental compatibility. The results and the high scientific level of this international research activity provide a sound basis for the realisation of the International Thermonuclear Experimental Reactor (ITER), whose goal is to demonstrate the scientific and technological feasibility of a fusion energy source for peaceful purposes.

Heating and current drive on NSTX

NASA Astrophysics Data System (ADS)

Wilson, J. R.; Batchelor, D.; Carter, M.; Hosea, J.; Ignat, D.; LeBlanc, B.; Majeski, R.; Ono, M.; Phillips, C. K.; Rogers, J. H.; Schilling, G.

1997-04-01

Low aspect ratio tokamaks pose interesting new challenges for heating and current drive. The NSTX (National Spherical Tokamak Experiment) device to be built at Princeton is a low aspect ratio toroidal device that has the achievement of high toroidal beta (˜45%) and non-inductive operation as two of its main research goals. To achieve these goals significant auxiliary heating and current drive systems are required. Present plans include ECH (Electron cyclotron heating) for pre-ionization and start-up assist, HHFW (high harmonic fast wave) for heating and current drive and eventually NBI (neutral beam injection) for heating, current drive and plasma rotation.

Stellarator Research Opportunities: A Report of the National Stellarator Coordinating Committee

NASA Astrophysics Data System (ADS)

Gates, D. A.; Anderson, D.; Anderson, S.; Zarnstorff, M.; Spong, D. A.; Weitzner, H.; Neilson, G. H.; Ruzic, D.; Andruczyk, D.; Harris, J. H.; Mynick, H.; Hegna, C. C.; Schmitz, O.; Talmadge, J. N.; Curreli, D.; Maurer, D.; Boozer, A. H.; Knowlton, S.; Allain, J. P.; Ennis, D.; Wurden, G.; Reiman, A.; Lore, J. D.; Landreman, M.; Freidberg, J. P.; Hudson, S. R.; Porkolab, M.; Demers, D.; Terry, J.; Edlund, E.; Lazerson, S. A.; Pablant, N.; Fonck, R.; Volpe, F.; Canik, J.; Granetz, R.; Ware, A.; Hanson, J. D.; Kumar, S.; Deng, C.; Likin, K.; Cerfon, A.; Ram, A.; Hassam, A.; Prager, S.; Paz-Soldan, C.; Pueschel, M. J.; Joseph, I.; Glasser, A. H.

2018-02-01

This document is the product of a stellarator community workshop, organized by the National Stellarator Coordinating Committee and referred to as Stellcon, that was held in Cambridge, Massachusetts in February 2016, hosted by MIT. The workshop was widely advertised, and was attended by 40 scientists from 12 different institutions including national labs, universities and private industry, as well as a representative from the Department of Energy. The final section of this document describes areas of community wide consensus that were developed as a result of the discussions held at that workshop. Areas where further study would be helpful to generate a consensus path forward for the US stellarator program are also discussed. The program outlined in this document is directly responsive to many of the strategic priorities of FES as articulated in "Fusion Energy Sciences: A Ten-Year Perspective (2015-2025)" [1]. The natural disruption immunity of the stellarator directly addresses "Elimination of transient events that can be deleterious to toroidal fusion plasma confinement devices" an area of critical importance for the US fusion energy sciences enterprise over the next decade. Another critical area of research "Strengthening our partnerships with international research facilities," is being significantly advanced on the W7-X stellarator in Germany and serves as a test-bed for development of successful international collaboration on ITER. This report also outlines how materials science as it relates to plasma and fusion sciences, another critical research area, can be carried out effectively in a stellarator. Additionally, significant advances along two of the Research Directions outlined in the report; "Burning Plasma Science: Foundations - Next-generation research capabilities", and "Burning Plasma Science: Long pulse - Sustainment of Long-Pulse Plasma Equilibria" are proposed.

Nuclear design of a very-low-activation fusion reactor

NASA Astrophysics Data System (ADS)

Cheng, E. T.; Hopkins, G. R.

1983-06-01

The nuclear design aspects of using very-low-activation materials, such as SiC, MgO, and aluminum for fusion-reactor first wall, blanket, and shield applications were investigated. In addition to the advantage of very-low radioactive inventory, it was found that the very-low-activation fusion reactor can also offer an adequate tritium-breeding ratio and substantial amount of blanket nuclear heating as a conventional-material-structured reactor does. The most-stringent design constraint found in a very-low-activation fusion reactor is the limited space available in the inboard region of a Tokamak concept for shielding to protect the superconducting toroidal field coil. A reference design was developed which mitigates the constraint by adopting a removable tungsten shield design that retains the inboard dimensions and gives the same shield performance as the reference STARFIRE Tokamak reactor design.

Commissioning and Plans for the NSTX-U Facility

NASA Astrophysics Data System (ADS)

Ono, Masayuki; NSTX-U Team

2016-10-01

The National Spherical Torus Experiment - Upgrade (NSTX-U) has started its first year of plasma operations after the successful completion of the CD-4 milestones. The unique operating regimes of NSTX-U can contribute to several important issues in the physics of burning plasmas to optimize the performance of ITER. The major mission of NSTX-U is also to develop the physics and technology basis for an ST-based Fusion Nuclear Science Facility (FNSF). The new center stack will provide toroidal field of 1 Tesla at a major radius of 0.93 m which should enable a plasma current of up to 2 mega-Amp for 5 sec. A much more tangential 2nd NBI system, with 2-3 times higher current drive efficiency compared to the 1st NBI system, is installed. NSTX-U is designed to attain the 100% non-inductive operation needed for a compact FNSF design. With higher fields and heating powers of 14 MW, the NSTX-U plasma collisionality will be reduced by a factor of 3-6 to help explore the trend in transport towards the low collisionality FNSF regime. If the favorable trends observed on NSTX holds at low collisionality, high fusion neutron fluences could be achievable in very compact ST devices.

Coaxial ion trap mass spectrometer: concentric toroidal and quadrupolar trapping regions.

PubMed

Peng, Ying; Hansen, Brett J; Quist, Hannah; Zhang, Zhiping; Wang, Miao; Hawkins, Aaron R; Austin, Daniel E

2011-07-15

We present the design and results for a new radio-frequency ion trap mass analyzer, the coaxial ion trap, in which both toroidal and quadrupolar trapping regions are created simultaneously. The device is composed of two parallel ceramic plates, the facing surfaces of which are lithographically patterned with concentric metal rings and covered with a thin film of germanium. Experiments demonstrate that ions can be trapped in either region, transferred from the toroidal to the quadrupolar region, and mass-selectively ejected from the quadrupolar region to a detector. Ions trapped in the toroidal region can be transferred to the quadrupole region using an applied ac signal in the radial direction, although it appears that the mechanism of this transfer does not involve resonance with the ion secular frequency, and the process is not mass selective. Ions in the quadrupole trapping region are mass analyzed using dipole resonant ejection. Multiple transfer steps and mass analysis scans are possible on a single population of ions, as from a single ionization/trapping event. The device demonstrates better mass resolving power than the radially ejecting halo ion trap and better sensitivity than the planar quadrupole ion trap.

Saturation of a toroidal Alfvén eigenmode due to enhanced damping of nonlinear sidebands

NASA Astrophysics Data System (ADS)

Todo, Y.; Berk, H. L.; Breizman, B. N.

2012-09-01

This paper examines nonlinear magneto-hydrodynamic effects on the energetic particle driven toroidal Alfvén eigenmode (TAE) for lower dissipation coefficients and with higher numerical resolution than in the previous simulations (Todo et al 2010 Nucl. Fusion 50 084016). The investigation is focused on a TAE mode with toroidal mode number n = 4. It is demonstrated that the mechanism of mode saturation involves generation of zonal (n = 0) and higher-n (n ⩾ 8) sidebands, and that the sidebands effectively increase the mode damping rate via continuum damping. The n = 0 sideband includes the zonal flow peaks at the TAE gap locations. It is also found that the n = 0 poloidal flow represents a balance between the nonlinear driving force from the n = 4 components and the equilibrium plasma response to the n = 0 fluctuations. The spatial profile of the n = 8 sideband peaks at the n = 8 Alfvén continuum, indicating enhanced dissipation due to continuum damping.

Determination of broken KAM surfaces for particle orbits in toroidal confinement systems

DOE Office of Scientific and Technical Information (OSTI.GOV)

White, R. B.

2015-10-05

Here, the destruction of Kolmogorov–Arnold–Moser surfaces in a Hamiltonian system is an important topic in nonlinear dynamics, and in particular in the theory of particle orbits in toroidal magnetic confinement systems. Analytic models for transport due to mode-particle resonances are not sufficiently correct to give the effect of these resonances on transport. In this paper we compare three different methods for the detection of the loss of stability of orbits in the dynamics of charged particles in a toroidal magnetic confinement device in the presence of time dependent magnetic perturbations.

Experimental results of 40-kA Nb[sub 3]Al cable-in-conduit conductor for fusion machines

DOE Office of Scientific and Technical Information (OSTI.GOV)

Takahashi, Y.; Sugimoto, M.; Isono, T.

1994-07-01

A 40-kA Nb[sub 3]Al cable-in-conduit conductor has been developed for the toroidal field coils of fusion reactors, because Nb[sub 3]Al has excellent mechanical performance. This conductor consists of 405 copper-stabilized multifilamentary strands inserted into a CuNi case circular conduit. The Nb[sub 3]Al strands are fabricated by the Jelly-roll process with a diameter of 1.22 mm. This conductor could be operated up to a current of 46 kA at an external field of 11.2 T. Accordingly, Nb[sub 3]Al promises to soon become a useful superconductor for large-scale high-field applications, such as fusion machines.

Divertor-localized fluctuations in NSTX-U L-mode discharges

NASA Astrophysics Data System (ADS)

Scotti, Filippo; Soukhanovskii, V. A.; Zweben, S.; Myra, J.; Baver, D.; Sabbagh, S. A.

2017-10-01

The 3-D structure of divertor turbulence is characterized in NSTX-U by means of fast camera imaging. Edge and divertor turbulence can be important in determining the heat flux width in fusion devices. Field-aligned filaments are found on the divertor legs via imaging of C III and D- α emission in NBI-heated diverted L-mode discharges, similar to observations in Alcator C-Mod and MAST. These flute-like fluctuations of up to 10-20% in RMS/mean are radially localized around the separatrix and limited to the region below the X-point. Poloidal and parallel correlation lengths are a few cm (10-50ρi) and several meters, respectively. For the outer leg filaments, poloidal correlation lengths decrease along the leg away from the strike point and typical effective toroidal mode numbers are in the range of 10-20. Opposite toroidal rotation is observed for inner (co-current rotation) and outer leg (counter-current rotation) filaments with apparent poloidal propagation of 1 km/s. The poloidal motion of outer leg filaments is opposite to the one typically observed for NSTX upstream blobs in the scrape-off layer. The shape, dynamics and absence of correlation with upstream turbulence suggest that these fluctuations are generated and localized in the divertor region. Supported by US DOE DE-AC52-07NA27344, DE-AC02-09CH11466, DE-FG02- 02ER54678, DE-FG02-99ER54524.

Modular coils and finite-β operation of a quasi-axially symmetric tokamak

NASA Astrophysics Data System (ADS)

Drevlak, M.

1998-09-01

Quasi-axially symmetric tokamaks (QA tokamaks) are an extension of the conventional tokamak concept. In these devices the magnetic field strength is independent of the generalized toroidal magnetic co-ordinate even though the cross-sectional shape changes. An optimized plasma equilibrium belonging to the class of QA tokamaks has been proposed by Nührenberg. It features the small aspect ratio of a tokamak while allowing part of the rotational transform to be generated by the external field. In this article, two particular aspects of the viability of QA tokamaks are explored, namely the feasibility of modular coils and the possibility of maintaining quasi-axial symmetry in the free-boundary equilibria obtained with the coils found. A set of easily feasible modular coils for the configuration is presented. It was designed using the extended version of the NESCOIL code (Merkel, P., Nucl. Fusion 27 (1987) 867). Using this coil system, free-boundary calculations of the plasma equilibrium were carried out using the NEMEC code (Hirshman, S.P., Van Rij, W.I., Merkel, P., Comput. Phys. Commun. 43 (1986) 143). It is observed that the effects of finite β and net toroidal plasma current can be compensated for with good precision by applying a vertical magnetic field and by separately adjusting the currents of the modular coils. A set of fully three dimensional (3-D) auxiliary coils is proposed to exert control on the rotational transform in the plasma. Deterioration of the quasi-axial symmetry induced by the auxiliary coils can be avoided by adequate adjustment of the currents in the primary coils. Finally, the neoclassical transport properties of the configuration are examined. It is observed that optimization with respect to confinement of the alpha particles can be maintained at operation with finite toroidal current if the aforementioned corrective measures are used. In this case, the neoclassical behaviour is shown to be very similar to that of a conventional tokamak.

Fusion Research in Ioffe Institute

NASA Astrophysics Data System (ADS)

Askinazi, L. G.; Afanasyev, V. I.; Altukhov, A. B.; Bakharev, N. N.; Belokurov, A. A.; Bulanin, V. V.; Bykov, A. S.; Chernyshev, F. V.; Chugunov, I.; Dyachenko, V. V.; Esipov, L. A.; Gin, D.; Goncharov, P. R.; Gurchenko, A. D.; Gusakov, E. Z.; Gusev, V. K.; Heuraux, S.; Iblyaminova, A. D.; Irzak, M. A.; Kantor, M. Yu.; Kaveeva, E. G.; Kiviniemi, T.; Khilkevitch, E. M.; Khitrov, S. A.; Khromov, N. A.; Kornev, V. A.; Kouprienko, D. V.; Kurskiev, G. S.; Lashkul, S. I.; Lebedev, S. V.; Leerink, S.; Melnik, A. D.; Minaev, V. B.; Mironov, M. I.; Miroshnikov, I. V.; Mukhin, E. E.; Nesenevich, V. G.; Niskala, P.; Novokhatsky, A. N.; Patrov, M. I.; Perevalov, A. A.; Petrov, M. P.; Petrov, A. V.; Petrov, Yu. V.; Popov, A. Yu.; Rozhansky, V. A.; Rozhdestvenskiy, V. V.; Sakharov, N. V.; Saveliev, A. N.; Senichenkov, I. Yu.; Sergeev, V. Yu.; Shchegolev, P. B.; Shatalin, S. V.; Shcherbinin, O. N.; Shevelev, A. E.; Sidorov, A. V.; da Silva, F.; Smirnov, A. I.; Stepanov, A. Yu.; Sysoeva, E. V.; Teplova, N. V.; Tolstyakov, S. Yu.; Tukachinsky, A. S.; Varfolomeev, V. I.; Vekshina, E. O.; Vildjunas, M. I.; Voronin, A. V.; Voskoboinikov, S. P.; Wagner, F.; Yashin, A. Yu.; Zhubr, N. A.

2015-10-01

In this paper we present the fusion related activities of the Plasma Physics Division at the Ioffe Institute. The first experiments on lower hybrid current drive (LHCD) in a spherical tokamak performed at the Globus-M tokamak (R = 0.36 m, a = 0.24 m, Bt = 0.4 T, Ip = 200 kA) with a novel poloidally oriented grill resulted in an RF driven current of up to 30 kA at (100 kW, 2.5 GHz), exceeding the modelling predictions. At the FT-2 tokamak (R = 0.56 m, a = 0.08 m, Bt = 3 T, Ip = 30 kA) experiments with a traditional toroidally oriented grill revealed no strong dependence of the LHCD density limit on the H/D ratio in spite of LH resonance densities differing by a factor of 3. Microwave Doppler reflectometry (DR) at the Globus-M, and DR and heavy ion beam probe measurements at the tokamak TUMAN-3M (R = 0.53 m, a = 0.24 m, Bt = 1.0 T, Ip = 190 kA) demonstrated geodesic acoustic mode (GAM) suppression at the L to H transition. Observations at FT-2 using Doppler Enhanced Scattering showed that the GAM amplitude is anti-correlated both spatially and temporally to the drift turbulence level and electron thermal diffusivity. For the first time turbulence amplitude modulation at the GAM frequency was found both experimentally and in global gyrokinetic modelling. A model of the L-H transition is proposed based on this effect. The loss mechanisms of energetic ions' (EI) were investigated in the neutral beam injection (NBI) experiments on Globus-M and TUMAN-3M. Empirical scaling of the 2.45 MeV DD neutron rate for the two devices shows a strong dependence on toroidal field Bt1.29 and plasma current Ip1.34 justifying the Bt and Ip increase by a factor of 2.5 for the proposed upgrade of Globus-M. Bursts of ˜1 MHz Alfvenic type oscillations correlating with sawtooth crashes were observed in ohmic TUMAN-3M discharges. The possibility of low threshold parametric excitation of Bernstein and upper hybrid waves trapped in drift-wave eddies resulting in anomalous absorption in electron cyclotron resonance heating (ECRH) experiments in toroidal plasmas was identified 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. On the progress in design and fabrication of three diagnostics for ITER developed in the Ioffe institute is reported: neutral particle analysis, divertor Thomson scattering and gamma spectroscopy.

New design of cable-in-conduit conductor for application in future fusion reactors

NASA Astrophysics Data System (ADS)

Qin, Jinggang; Wu, Yu; Li, Jiangang; Liu, Fang; Dai, Chao; Shi, Yi; Liu, Huajun; Mao, Zhehua; Nijhuis, Arend; Zhou, Chao; Yagotintsev, Konstantin A.; Lubkemann, Ruben; Anvar, V. A.; Devred, Arnaud

2017-11-01

The China Fusion Engineering Test Reactor (CFETR) is a new tokamak device whose magnet system includes toroidal field, central solenoid (CS) and poloidal field coils. The main goal is to build a fusion engineering tokamak reactor with about 1 GW fusion power and self-sufficiency by blanket. In order to reach this high performance, the magnet field target is 15 T. However, the huge electromagnetic load caused by high field and current is a threat for conductor degradation under cycling. The conductor with a short-twist-pitch (STP) design has large stiffness, which enables a significant performance improvement in view of load and thermal cycling. But the conductor with STP design has a remarkable disadvantage: it can easily cause severe strand indentation during cabling. The indentation can reduce the strand performance, especially under high load cycling. In order to overcome this disadvantage, a new design is proposed. The main characteristic of this new design is an updated layout in the triplet. The triplet is made of two Nb3Sn strands and one soft copper strand. The twist pitch of the two Nb3Sn strands is large and cabled first. The copper strand is then wound around the two superconducting strands (CWS) with a shorter twist pitch. The following cable stages layout and twist pitches are similar to the ITER CS conductor with STP design. One short conductor sample with a similar scale to the ITER CS was manufactured and tested with the Twente Cable Press to investigate the mechanical properties, AC loss and internal inspection by destructive examination. The results are compared to the STP conductor (ITER CS and CFETR CSMC) tests. The results show that the new conductor design has similar stiffness, but much lower strand indentation than the STP design. The new design shows potential for application in future fusion reactors.

MHD Studies of Advanced Tokamak Equilibria

NASA Astrophysics Data System (ADS)

Strumberger, E.

2005-10-01

Advanced tokamak scenarios are often characterized by an extremely reversed profile of the safety factor, q, and a fast toroidal rotation. ASDEX Upgrade type equilibria with toroidal flow are computed up to a toroidal Mach number of Mta= 0.5, and compared with the static solution. Using these equilibria, the stabilizing effect of differential toroidal rotation on double tearing modes (DTMs) is investigated. These studies show that the computation of equilibria with flow is necessary for toroidally rotating plasma with Mta>=0.2. The use of ρtor instead of ρpol as radial coordinate enables us also to investigate the stability of equilibria with current holes. For numerical reasons, the rotational transform, = 1/q, has to be unequal zero in the CASTOR$FLOW code, but values of a>=0.001 (qa<=1000) can be easily handled. Stability studies of DTMs in the presence of a current hole are presented. Tokamak equilibria are only approximately axisymmetric. The finite number of toroidal field coils destroys the perfect axisymmetry of the device, and the coils produce a short wavelength ripple in the magnetic field strength. This toroidal field ripple plays a crucial role for the loss of high energy particles. Therefore, three-dimensional tokamak equilibria with and without current holes are computed for various plasma beta values. In addition the influence of the plasma beta on the toroidal field ripple is investigated.

Extending SIESTA capabilities: removing field-periodic and stellarator symmetric limitations

NASA Astrophysics Data System (ADS)

Cook, C. R.; Hirshman, S. P.; Sanchez, R.; Anderson, D. T.

2011-10-01

SIESTA is a three-dimensional magnetohydrodynamics equilibrium code capable of resolving magnetic islands in toroidal plasma confinement devices. Currently SIESTA assumes that plasma perturbations, and thus also magnetic islands, are field-periodic. This limitation is being removed from the code by allowing the displacement toroidal mode number to not be restricted to multiples of the number of field periods. Extending SIESTA in this manner will allow larger, lower-order resonant islands to form in devices such as CTH. An example of a non-field-periodic perturbation in CTH will be demonstrated. Currently the code also operates in a stellarator-symmetric fashion in which an ``up-down'' symmetry is present at some toroidal angle. Nearly all of the current tokamaks (and ITER in the future) operate with a divertor and as such do not possess stellarator symmetry. Removal of this symmetry restriction requires including both sine and cosine terms in the Fourier expansion for the geometry of the device and the fields contained within. The current status of this extension of the code will be discussed, along with the method of implementation. U.S. DOE Contract No. DE-AC05-00OR22725 with UT-Battelle, LLC.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Shaing, K. C.; Peng, Yueng Kay Martin

Transport theory for potato orbits in the region near the magnetic axis in an axisymmetric torus such as tokamaks and spherical tori is extended to the situation where the toroidal flow speed is of the order of the sonic speed as observed in National Spherical Torus Experiment [E. J. Synakowski, M. G. Bell, R. E. Bell et al., Nucl. Fusion 43, 1653 (2003)]. It is found that transport fluxes such as ion radial heat flux, and bootstrap current density are modified by a factor of the order of the square of the toroidal Mach number. The consequences of the orbitmore » squeezing are also presented. The theory is developed for parabolic (in radius r) plasma profiles. A method to apply the results of the theory for the transport modeling is discussed.« less

Initial operation of the NSTX-Upgrade real-time velocity diagnostic

DOE PAGES

Podestà, M.; Bell, R. E.

2016-11-03

A real-time velocity (RTV) diagnostic based on active charge-exchange recombination spectroscopy is now operational on the National Spherical Torus Experiment-Upgrade (NSTX-U) spherical torus (Menard et al 2012 Nucl. Fusion 52 083015). We designed the system in order to supply plasma velocity data in real time to the NSTX-U plasma control system, as required for the implementation of toroidal rotation control. Our measurements are available from four radii at a maximum sampling frequency of 5 kHz. Post-discharge analysis of RTV data provides additional information on ion temperature, toroidal velocity and density of carbon impurities. Furthermore, examples of physics studies enabled bymore » RTV measurements from initial operations of NSTX-U are discussed.« less

On the interplay between neoclassical tearing modes and nonlocal transport in toroidal plasmas

NASA Astrophysics Data System (ADS)

Ji, X. Q.; Xu, Y.; Hidalgo, C.; Diamond, P. H.; Liu, Yi; Pan, O.; Shi, Z. B.; Yu, D. L.

2016-09-01

This Letter presents the first observation on the interplay between nonlocal transport and neoclassical tearing modes (NTMs) during transient nonlocal heat transport events in the HL-2A tokamak. The nonlocality is triggered by edge cooling and large-scale, inward propagating avalanches. These lead to a locally enhanced pressure gradient at the q = 3/2 (or 2/1) rational surface and hence the onset of the NTM in relatively low β plasmas (βN < 1). The NTM, in return, regulates the nonlocal transport by truncation of avalanches by local sheared toroidal flows which develop near the magnetic island. These findings have direct implications for understanding the dynamic interaction between turbulence and large-scale mode structures in fusion plasmas.

Ferrotoroidic ground state in a heterometallic {CrIIIDyIII6} complex displaying slow magnetic relaxation.

PubMed

Vignesh, Kuduva R; Soncini, Alessandro; Langley, Stuart K; Wernsdorfer, Wolfgang; Murray, Keith S; Rajaraman, Gopalan

2017-10-18

Toroidal quantum states are most promising for building quantum computing and information storage devices, as they are insensitive to homogeneous magnetic fields, but interact with charge and spin currents, allowing this moment to be manipulated purely by electrical means. Coupling molecular toroids into larger toroidal moments via ferrotoroidic interactions can be pivotal not only to enhance ground state toroidicity, but also to develop materials displaying ferrotoroidic ordered phases, which sustain linear magneto-electric coupling and multiferroic behavior. However, engineering ferrotoroidic coupling is known to be a challenging task. Here we have isolated a {Cr III Dy III 6 } complex that exhibits the much sought-after ferrotoroidic ground state with an enhanced toroidal moment, solely arising from intramolecular dipolar interactions. Moreover, a theoretical analysis of the observed sub-Kelvin zero-field hysteretic spin dynamics of {Cr III Dy III 6 } reveals the pivotal role played by ferrotoroidic states in slowing down the magnetic relaxation, in spite of large calculated single-ion quantum tunneling rates.

Nuclear Fusion Award 2009 speech Nuclear Fusion Award 2009 speech

NASA Astrophysics Data System (ADS)

Sabbagh, Steven Anthony

2011-01-01

This is an exceptional moment in my career, and so I want to thank all of my teachers, colleagues and mentors who have made this possible. From my co-authors and myself, many thanks to the International Atomic Energy Agency, IOP Publishing, the Nuclear Fusion journal team, and the selection committee for the great honor of receiving this award. Also gratitude to Kikuchi-sensei, not only for the inventive and visionary creation of this award, but also for being a key mentor dating back to his efforts in producing high neutron output in JT-60U. It was also a great honor to receive the award directly from IAEA Deputy Director General Burkart during the 23rd IAEA Fusion Energy Conference in Daejeon. Receiving the award at this venue is particularly exciting as Daejeon is home to the new, next-generation KSTAR tokamak device that will lead key magnetic fusion research areas going forward. I would also like to thank the mayor of Daejeon, Dr Yum Hong-Chul, and all of the meeting organizers for giving us all a truly spectacular and singular welcoming event during which the award was presented. The research leading to the award would not have been possible without the support of the US Department of Energy, and I thank the Department for the continued funding of this research. Special mention must be made to a valuable co-author who is no longer with us, Professor A. Bondeson, who was a significant pioneer in resistive wall mode (RWM) research. I would like to thank my wife, Mary, for her infinite patience and encouragement. Finally, I would like to personally thank all of you that have approached and congratulated me directly. There are no units to measure how important your words have been in this regard. When notified that our paper had been shortlisted for the 2009 Nuclear Fusion Award, my co-authors responded echoing how I felt—honored to be included in such a fine collection of research by colleagues. It was unfathomable—would this paper follow the brilliant work of Dr Todd Evans, another significant mentor of mine, as winner of this prestigious award? Then, it happened. The paper covers several key topics related to high beta tokamak physics. For me, the greatest satisfaction in receiving this award is because it was the first Nuclear Fusion Award to recognize research on the National Spherical Torus Experiment (NSTX) located at the Princeton Plasma Physics Laboratory. The achievement of record stability parameters in a mega-Ampere class spherical torus (ST) device reported in the paper represents a multi-year effort, contributed to by the entire research team. Research to maintain such plasmas for an indefinite period continues today. Understanding RWM stabilization physics is crucial for this goal, and leveraging the high beta ST operating space uniquely tests theory for application to future STs and to tokamaks in general, including advanced operational scenarios of ITER. For instance, the RWM was found to have significant amplitude in components with the toroidal mode number greater than unity. This has important implications for general active RWM control. Evidence that the RWM passive stabilization physics and marginal stability criterion are indeed more complex than originally thought was shown in this paper. Present work shows the greater complexity has a direct impact on how we should extrapolate RWM stabilization to future devices. The paper also reported the qualitative observation of neoclassical toroidal viscosity (NTV), followed by a companion paper by our group in 2006 reporting the quantitative observation of this effect and comparison to theory. The physics of this interesting and important phenomenon was introduced to me by Professor J. Callen (who has given an overview talk at this conference including this subject) and Professor Kerchung Shaing of the University of Wisconsin, to whom I am quite indebted. The paper also reported the first measurement of resonant field amplification at high beta in the NSTX, following work of the Columbia University group at DIII-D during that period. My greatest hope in our stability physics research effort is that our insight in this portion of the much larger research effort, of which we all partake, to make fusion reactors a practical reality, will give new and future researchers the input and motivation to amplify our work and create realities that we had thought were just out of reach. Receiving the 2009 IAEA Nuclear Fusion Award is a substantial honor that greatly motivates me to continue to support the international nuclear fusion research effort at the highest level possible. So, please allow me to raise this beautiful trophy high, here today, to best remember this fine honor. Thank you. Steven Anthony Sabbagh 2009 Nuclear Fusion Award winner Columbia University, New York, NY, USA

Realizing "2001: A Space Odyssey": Piloted Spherical Torus Nuclear Fusion Propulsion

NASA Technical Reports Server (NTRS)

Williams, Craig H.; Dudzinski, Leonard A.; Borowski, Stanley K.; Juhasz, Albert J.

2005-01-01

A conceptual vehicle design enabling fast, piloted outer solar system travel was created predicated on a small aspect ratio spherical torus nuclear fusion reactor. The initial requirements were satisfied by the vehicle concept, which could deliver a 172 mt crew payload from Earth to Jupiter rendezvous in 118 days, with an initial mass in low Earth orbit of 1,690 mt. Engineering conceptual design, analysis, and assessment was performed on all major systems including artificial gravity payload, central truss, nuclear fusion reactor, power conversion, magnetic nozzle, fast wave plasma heating, tankage, fuel pellet injector, startup/re-start fission reactor and battery bank, refrigeration, reaction control, communications, mission design, and space operations. Detailed fusion reactor design included analysis of plasma characteristics, power balance/utilization, first wall, toroidal field coils, heat transfer, and neutron/x-ray radiation. Technical comparisons are made between the vehicle concept and the interplanetary spacecraft depicted in the motion picture 2001: A Space Odyssey.

Effects of Equilibrium Toroidal Flow on Locked Mode and Plasma Response in a Tokamak

NASA Astrophysics Data System (ADS)

Zhu, Ping; Huang, Wenlong; Yan, Xingting

2016-10-01

It is widely believed that plasma flow plays significant roles in regulating the processes of mode locking and plasma response in a tokamak in presence of external resonant magnetic perturbations (RMPs). Recently a common analytic relation for both locked mode and plasma response has been developed based on the steady-state solution to the coupled dynamic system of magnetic island evolution and torque balance. The analytic relation predicts the size of the magnetic island of a locked mode or a static nonlinear plasma response for a given RMP amplitude, and rigorously proves a screening effect of the equilibrium toroidal flow. To test the theory, we solve for the locked mode and the nonlinear plasma response in presence of RMP for a circular-shaped limiter tokamak equilibrium with constant toroidal flow, using the initial-value, full MHD simulation code NIMROD. The comparison between the simulation results and the theory prediction, in terms of the quantitative screening effects of equilibrium toroidal flow, will be reported and discussed. Supported by National Magnetic Confinement Fusion Science Program of China Grants 2014GB124002 and 2015GB101004, the 100 Talent Program of the Chinese Academy of Sciences, and U.S. Department of Energy Grants DE-FG02-86ER53218 and DE-FC02-08ER54975.

Toroidal magnet system

DOEpatents

Ohkawa, Tihiro; Baker, Charles C.

1981-01-01

In a plasma device having a toroidal plasma containment vessel, a toroidal field-generating coil system includes fixed linking coils each formed of first and second sections with the first section passing through a central opening through the containment vessel and the second section completing the linking coil to link the containment vessel. A plurality of removable unlinked coils are each formed of first and second C-shaped sections joined to each other at their open ends with their bights spaced apart. The second C-shaped section of each movable coil is removably mounted adjacent the second section of a linking coil, with the containment vessel disposed between the open ends of the first and second C-shaped sections. Electric current is passed through the linking and removable coils in opposite sense in the respective adjacent second sections to produce a net toroidal field.

Stochastic effects on phase-space holes and clumps in kinetic systems near marginal stability

DOE PAGES

Woods, Benjamin J. Q.; Duarte, Vinicius N.; De-Gol, Anthony J.; ...

2018-01-23

The creation and subsequent evolution of marginally-unstable modes have been observed in a wide range of fusion devices. This behaviour has been successfully explained, for a single frequency shifting mode, in terms of phase-space structures known as a 'hole' and 'clump'. Here in this paper, we introduce stochasticity into a 1D kinetic model, affecting the formation and evolution of resonant modes in the system. We find that noise in the fast particle distribution or electric field leads to a shift in the asymptotic behaviour of a chirping resonant mode; this noise heuristically maps onto radial microturbulence via canonical toroidal momentummore » scattering, affecting hole and clump formation. While the mechanism allowing for the formation of the hole and clump is coherent, the lifetime of a hole and clump is shown to be highly sensitive to initial conditions, affecting the temporal profile of a single bursting event in mode amplitude.« less

Effects of internal structure on equilibrium of field-reversed configuration plasma sustained by rotating magnetic field

DOE Office of Scientific and Technical Information (OSTI.GOV)

Yambe, Kiyoyuki; Inomoto, Michiaki; Okada, Shigefumi

The effects of an internal structure on the equilibrium of a field-reversed configuration (FRC) plasma sustained by rotating magnetic field is investigated by using detailed electrostatic probe measurements in the FRC Injection Experiment apparatus [S. Okada, et al., Nucl. Fusion. 45, 1094 (2005)]. An internal structure installed axially on the geometrical axis, which simulates Ohmic transformer or external toroidal field coils on the FRC device, brings about substantial changes in plasma density profile. The internal structure generates steep density-gradients not only on the inner side but on the outer side of the torus. The radial electric field is observed tomore » sustain the ion thermal pressure-gradient in the FRC without the internal structure; however, the radial electric field is not sufficient to sustain the increased ion thermal pressure-gradient in the FRC with the internal structure. Spontaneously driven azimuthal ion flow will be accountable for the imbalance of the radial pressure which is modified by the internal structure.« less

Finding Regions of Interest on Toroidal Meshes

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wu, Kesheng; Sinha, Rishi R; Jones, Chad

2011-02-09

Fusion promises to provide clean and safe energy, and a considerable amount of research effort is underway to turn this aspiration intoreality. This work focuses on a building block for analyzing data produced from the simulation of microturbulence in magnetic confinementfusion devices: the task of efficiently extracting regions of interest. Like many other simulations where a large amount of data are produced,the careful study of ``interesting'' parts of the data is critical to gain understanding. In this paper, we present an efficient approach forfinding these regions of interest. Our approach takes full advantage of the underlying mesh structure in magneticmore » coordinates to produce acompact representation of the mesh points inside the regions and an efficient connected component labeling algorithm for constructingregions from points. This approach scales linearly with the surface area of the regions of interest instead of the volume as shown with bothcomputational complexity analysis and experimental measurements. Furthermore, this new approach is 100s of times faster than a recentlypublished method based on Cartesian coordinates.« less

Fast-ion Dα spectrum diagnostic in the EAST

NASA Astrophysics Data System (ADS)

Hou, Y. M.; Wu, C. R.; Huang, J.; Heidbrink, W. W.; von Hellermann, M. G.; Xu, Z.; Jin, Z.; Chang, J. F.; Zhu, Y. B.; Gao, W.; Chen, Y. J.; Lyu, B.; Hu, R. J.; Zhang, P. F.; Zhang, L.; Gao, W.; Wu, Z. W.; Yu, Y.; Ye, M. Y.

2016-11-01

In toroidal magnetic fusion devices, fast-ion D-alpha diagnostic (FIDA) is a powerful method to study the fast-ion feature. The fast-ion characteristics can be inferred from the Doppler shifted spectrum of Dα light according to charge exchange recombination process between fast ions and probe beam. Since conceptual design presented in the last HTPD conference, significant progress has been made to apply FIDA systems on the Experimental Advanced Superconducting Tokamak (EAST). Both co-current and counter-current neutral beam injectors are available, and each can deliver 2-4 MW beam power with 50-80 keV beam energy. Presently, two sets of high throughput spectrometer systems have been installed on EAST, allowing to capture passing and trapped fast-ion characteristics simultaneously, using Kaiser HoloSpec transmission grating spectrometer and Bunkoukeiki FLP-200 volume phase holographic spectrometer coupled with Princeton Instruments ProEM 1024B eXcelon and Andor DU-888 iXon3 1024 CCD camera, respectively. This paper will present the details of the hardware descriptions and experimental spectrum.

Stochastic effects on phase-space holes and clumps in kinetic systems near marginal stability

DOE Office of Scientific and Technical Information (OSTI.GOV)

Woods, Benjamin J. Q.; Duarte, Vinicius N.; De-Gol, Anthony J.

The creation and subsequent evolution of marginally-unstable modes have been observed in a wide range of fusion devices. This behaviour has been successfully explained, for a single frequency shifting mode, in terms of phase-space structures known as a 'hole' and 'clump'. Here in this paper, we introduce stochasticity into a 1D kinetic model, affecting the formation and evolution of resonant modes in the system. We find that noise in the fast particle distribution or electric field leads to a shift in the asymptotic behaviour of a chirping resonant mode; this noise heuristically maps onto radial microturbulence via canonical toroidal momentummore » scattering, affecting hole and clump formation. While the mechanism allowing for the formation of the hole and clump is coherent, the lifetime of a hole and clump is shown to be highly sensitive to initial conditions, affecting the temporal profile of a single bursting event in mode amplitude.« less

Neoclassical simulation of tokamak plasmas using the continuum gyrokinetic code TEMPEST.

PubMed

Xu, X Q

2008-07-01

We present gyrokinetic neoclassical simulations of tokamak plasmas with a self-consistent electric field using a fully nonlinear (full- f ) continuum code TEMPEST in a circular geometry. A set of gyrokinetic equations are discretized on a five-dimensional computational grid in phase space. The present implementation is a method of lines approach where the phase-space derivatives are discretized with finite differences, and implicit backward differencing formulas are used to advance the system in time. The fully nonlinear Boltzmann model is used for electrons. The neoclassical electric field is obtained by solving the gyrokinetic Poisson equation with self-consistent poloidal variation. With a four-dimensional (psi,theta,micro) version of the TEMPEST code, we compute the radial particle and heat fluxes, the geodesic-acoustic mode, and the development of the neoclassical electric field, which we compare with neoclassical theory using a Lorentz collision model. The present work provides a numerical scheme for self-consistently studying important dynamical aspects of neoclassical transport and electric field in toroidal magnetic fusion devices.

Neoclassical simulation of tokamak plasmas using the continuum gyrokinetic code TEMPEST

NASA Astrophysics Data System (ADS)

Xu, X. Q.

2008-07-01

We present gyrokinetic neoclassical simulations of tokamak plasmas with a self-consistent electric field using a fully nonlinear (full- f ) continuum code TEMPEST in a circular geometry. A set of gyrokinetic equations are discretized on a five-dimensional computational grid in phase space. The present implementation is a method of lines approach where the phase-space derivatives are discretized with finite differences, and implicit backward differencing formulas are used to advance the system in time. The fully nonlinear Boltzmann model is used for electrons. The neoclassical electric field is obtained by solving the gyrokinetic Poisson equation with self-consistent poloidal variation. With a four-dimensional (ψ,θ,γ,μ) version of the TEMPEST code, we compute the radial particle and heat fluxes, the geodesic-acoustic mode, and the development of the neoclassical electric field, which we compare with neoclassical theory using a Lorentz collision model. The present work provides a numerical scheme for self-consistently studying important dynamical aspects of neoclassical transport and electric field in toroidal magnetic fusion devices.

Toroidal rotation induced by 4.6 GHz lower hybrid current drive on EAST tokamak

NASA Astrophysics Data System (ADS)

Yin, Xiang-Hui; Chen, Jun; Hu, Rui-Ji; Li, Ying-Ying; Wang, Fu-Di; Fu, Jia; Ding, Bo-Jiang; Wang, Mao; Liu, Fu-Kun; Zang, Qing; Shi, Yue-Jiang; Lyu, Bo; Wan, Bao-Nian; EAST Team

2017-10-01

Not Available Project supported by the National Magnetic Confinement Fusion Science Program of China (Grant Nos. 2013GB112004 and 2015GB103002), the National Natural Science Foundation of China (Grant Nos. 11405212 and 11261140328), and the Major Program of Development Foundation of Hefei Center for Physical Science and Technology China (Grant No. 2016FXZY008).

Method and system to directly produce electrical power within the lithium blanket region of a magnetically confined, deuterium-tritium (DT) fueled, thermonuclear fusion reactor

DOEpatents

Woolley, Robert D.

1999-01-01

A method for integrating liquid metal magnetohydrodynamic power generation with fusion blanket technology to produce electrical power from a thermonuclear fusion reactor located within a confining magnetic field and within a toroidal structure. A hot liquid metal flows from a liquid metal blanket region into a pump duct of an electromagnetic pump which moves the liquid metal to a mixer where a gas of predetermined pressure is mixed with the pressurized liquid metal to form a Froth mixture. Electrical power is generated by flowing the Froth mixture between electrodes in a generator duct. When the Froth mixture exits the generator the gas is separated from the liquid metal and both are recycled.

21 CFR 886.1880 - Fusion and stereoscopic target.

Code of Federal Regulations, 2010 CFR

2010-04-01

... 21 Food and Drugs 8 2010-04-01 2010-04-01 false Fusion and stereoscopic target. 886.1880 Section... (CONTINUED) MEDICAL DEVICES OPHTHALMIC DEVICES Diagnostic Devices § 886.1880 Fusion and stereoscopic target. (a) Identification. A fusion and stereoscopic target is a device intended for use as a viewing object...

21 CFR 886.1880 - Fusion and stereoscopic target.

Code of Federal Regulations, 2011 CFR

2011-04-01

... 21 Food and Drugs 8 2011-04-01 2011-04-01 false Fusion and stereoscopic target. 886.1880 Section... (CONTINUED) MEDICAL DEVICES OPHTHALMIC DEVICES Diagnostic Devices § 886.1880 Fusion and stereoscopic target. (a) Identification. A fusion and stereoscopic target is a device intended for use as a viewing object...

Plasma-wall interaction studies within the EUROfusion consortium: progress on plasma-facing components development and qualification

NASA Astrophysics Data System (ADS)

Brezinsek, S.; Coenen, J. W.; Schwarz-Selinger, T.; Schmid, K.; Kirschner, A.; Hakola, A.; Tabares, F. L.; van der Meiden, H. J.; Mayoral, M.-L.; Reinhart, M.; Tsitrone, E.; Ahlgren, T.; Aints, M.; Airila, M.; Almaviva, S.; Alves, E.; Angot, T.; Anita, V.; Arredondo Parra, R.; Aumayr, F.; Balden, M.; Bauer, J.; Ben Yaala, M.; Berger, B. M.; Bisson, R.; Björkas, C.; Bogdanovic Radovic, I.; Borodin, D.; Bucalossi, J.; Butikova, J.; Butoi, B.; Čadež, I.; Caniello, R.; Caneve, L.; Cartry, G.; Catarino, N.; Čekada, M.; Ciraolo, G.; Ciupinski, L.; Colao, F.; Corre, Y.; Costin, C.; Craciunescu, T.; Cremona, A.; De Angeli, M.; de Castro, A.; Dejarnac, R.; Dellasega, D.; Dinca, P.; Dittmar, T.; Dobrea, C.; Hansen, P.; Drenik, A.; Eich, T.; Elgeti, S.; Falie, D.; Fedorczak, N.; Ferro, Y.; Fornal, T.; Fortuna-Zalesna, E.; Gao, L.; Gasior, P.; Gherendi, M.; Ghezzi, F.; Gosar, Ž.; Greuner, H.; Grigore, E.; Grisolia, C.; Groth, M.; Gruca, M.; Grzonka, J.; Gunn, J. P.; Hassouni, K.; Heinola, K.; Höschen, T.; Huber, S.; Jacob, W.; Jepu, I.; Jiang, X.; Jogi, I.; Kaiser, A.; Karhunen, J.; Kelemen, M.; Köppen, M.; Koslowski, H. R.; Kreter, A.; Kubkowska, M.; Laan, M.; Laguardia, L.; Lahtinen, A.; Lasa, A.; Lazic, V.; Lemahieu, N.; Likonen, J.; Linke, J.; Litnovsky, A.; Linsmeier, Ch.; Loewenhoff, T.; Lungu, C.; Lungu, M.; Maddaluno, G.; Maier, H.; Makkonen, T.; Manhard, A.; Marandet, Y.; Markelj, S.; Marot, L.; Martin, C.; Martin-Rojo, A. B.; Martynova, Y.; Mateus, R.; Matveev, D.; Mayer, M.; Meisl, G.; Mellet, N.; Michau, A.; Miettunen, J.; Möller, S.; Morgan, T. W.; Mougenot, J.; Mozetič, M.; Nemanič, V.; Neu, R.; Nordlund, K.; Oberkofler, M.; Oyarzabal, E.; Panjan, M.; Pardanaud, C.; Paris, P.; Passoni, M.; Pegourie, B.; Pelicon, P.; Petersson, P.; Piip, K.; Pintsuk, G.; Pompilian, G. O.; Popa, G.; Porosnicu, C.; Primc, G.; Probst, M.; Räisänen, J.; Rasinski, M.; Ratynskaia, S.; Reiser, D.; Ricci, D.; Richou, M.; Riesch, J.; Riva, G.; Rosinski, M.; Roubin, P.; Rubel, M.; Ruset, C.; Safi, E.; Sergienko, G.; Siketic, Z.; Sima, A.; Spilker, B.; Stadlmayr, R.; Steudel, I.; Ström, P.; Tadic, T.; Tafalla, D.; Tale, I.; Terentyev, D.; Terra, A.; Tiron, V.; Tiseanu, I.; Tolias, P.; Tskhakaya, D.; Uccello, A.; Unterberg, B.; Uytdenhoven, I.; Vassallo, E.; Vavpetič, P.; Veis, P.; Velicu, I. L.; Vernimmen, J. W. M.; Voitkans, A.; von Toussaint, U.; Weckmann, A.; Wirtz, M.; Založnik, A.; Zaplotnik, R.; PFC contributors, WP

2017-11-01

The provision of a particle and power exhaust solution which is compatible with first-wall components and edge-plasma conditions is a key area of present-day fusion research and mandatory for a successful operation of ITER and DEMO. The work package plasma-facing components (WP PFC) within the European fusion programme complements with laboratory experiments, i.e. in linear plasma devices, electron and ion beam loading facilities, the studies performed in toroidally confined magnetic devices, such as JET, ASDEX Upgrade, WEST etc. The connection of both groups is done via common physics and engineering studies, including the qualification and specification of plasma-facing components, and by modelling codes that simulate edge-plasma conditions and the plasma-material interaction as well as the study of fundamental processes. WP PFC addresses these critical points in order to ensure reliable and efficient use of conventional, solid PFCs in ITER (Be and W) and DEMO (W and steel) with respect to heat-load capabilities (transient and steady-state heat and particle loads), lifetime estimates (erosion, material mixing and surface morphology), and safety aspects (fuel retention, fuel removal, material migration and dust formation) particularly for quasi-steady-state conditions. Alternative scenarios and concepts (liquid Sn or Li as PFCs) for DEMO are developed and tested in the event that the conventional solution turns out to not be functional. Here, we present an overview of the activities with an emphasis on a few key results: (i) the observed synergistic effects in particle and heat loading of ITER-grade W with the available set of exposition devices on material properties such as roughness, ductility and microstructure; (ii) the progress in understanding of fuel retention, diffusion and outgassing in different W-based materials, including the impact of damage and impurities like N; and (iii), the preferential sputtering of Fe in EUROFER steel providing an in situ W surface and a potential first-wall solution for DEMO.

Plasma–wall interaction studies within the EUROfusion consortium: progress on plasma-facing components development and qualification

DOE PAGES

Brezinsek, S.; Coenen, J. W.; Schwarz-Selinger, T.; ...

2017-06-14

The provision of a particle and power exhaust solution which is compatible with first-wall components and edge-plasma conditions is a key area of present-day fusion research and mandatory for a successful operation of ITER and DEMO. The work package plasma-facing components (WP PFC) within the European fusion programme complements with laboratory experiments, i.e. in linear plasma devices, electron and ion beam loading facilities, the studies performed in toroidally confined magnetic devices, such as JET, ASDEX Upgrade, WEST etc. The connection of both groups is done via common physics and engineering studies, including the qualification and specification of plasma-facing components, andmore » by modelling codes that simulate edge-plasma conditions and the plasma–material interaction as well as the study of fundamental processes. WP PFC addresses these critical points in order to ensure reliable and efficient use of conventional, solid PFCs in ITER (Be and W) and DEMO (W and steel) with respect to heat-load capabilities (transient and steady-state heat and particle loads), lifetime estimates (erosion, material mixing and surface morphology), and safety aspects (fuel retention, fuel removal, material migration and dust formation) particularly for quasi-steady-state conditions. Alternative scenarios and concepts (liquid Sn or Li as PFCs) for DEMO are developed and tested in the event that the conventional solution turns out to not be functional. Here, we present an overview of the activities with an emphasis on a few key results: (i) the observed synergistic effects in particle and heat loading of ITER-grade W with the available set of exposition devices on material properties such as roughness, ductility and microstructure; (ii) the progress in understanding of fuel retention, diffusion and outgassing in different W-based materials, including the impact of damage and impurities like N; and (iii), the preferential sputtering of Fe in EUROFER steel providing an in situ W surface and a potential first-wall solution for DEMO.« less

Plasma–wall interaction studies within the EUROfusion consortium: progress on plasma-facing components development and qualification

DOE Office of Scientific and Technical Information (OSTI.GOV)

Brezinsek, S.; Coenen, J. W.; Schwarz-Selinger, T.

The provision of a particle and power exhaust solution which is compatible with first-wall components and edge-plasma conditions is a key area of present-day fusion research and mandatory for a successful operation of ITER and DEMO. The work package plasma-facing components (WP PFC) within the European fusion programme complements with laboratory experiments, i.e. in linear plasma devices, electron and ion beam loading facilities, the studies performed in toroidally confined magnetic devices, such as JET, ASDEX Upgrade, WEST etc. The connection of both groups is done via common physics and engineering studies, including the qualification and specification of plasma-facing components, andmore » by modelling codes that simulate edge-plasma conditions and the plasma–material interaction as well as the study of fundamental processes. WP PFC addresses these critical points in order to ensure reliable and efficient use of conventional, solid PFCs in ITER (Be and W) and DEMO (W and steel) with respect to heat-load capabilities (transient and steady-state heat and particle loads), lifetime estimates (erosion, material mixing and surface morphology), and safety aspects (fuel retention, fuel removal, material migration and dust formation) particularly for quasi-steady-state conditions. Alternative scenarios and concepts (liquid Sn or Li as PFCs) for DEMO are developed and tested in the event that the conventional solution turns out to not be functional. Here, we present an overview of the activities with an emphasis on a few key results: (i) the observed synergistic effects in particle and heat loading of ITER-grade W with the available set of exposition devices on material properties such as roughness, ductility and microstructure; (ii) the progress in understanding of fuel retention, diffusion and outgassing in different W-based materials, including the impact of damage and impurities like N; and (iii), the preferential sputtering of Fe in EUROFER steel providing an in situ W surface and a potential first-wall solution for DEMO.« less

Design of a toroidal device with a high temperature superconductor coil for non-neutral plasma trap

NASA Astrophysics Data System (ADS)

Ogawa, Yuichi; Morikawa, Junji; Nihei, Hitoshi; Ozawa, Daisaku; Yoshida, Zensho; Mito, Toshiyuki; Yanagi, Nagato; Iwakuma, Masataka

2002-01-01

The non-neutral plasma confinement device with a floating internal coil is under construction, where the high temperature superconductor (HTS) Ag-sheathed BSCCO-2223 is employed as the floating coil. We have two topics with this device: one is a trap of a non-neutral plasma consisting of one species, and another is an exploration of a high beta plasma based on two-fluid MHD relaxation theory. In the latter case the plasma should be non-neutralized in order to drive the plasma flow in the toroidal direction. The expected plasma parameters are discussed. Key elements of engineering issues have already developed. In addition, we have fabricated a small HTS coil and succeeded in levitating it within an accuracy of 25˜30 μm for 4 min or more.

High performance advanced tokamak regimes in DIII-D for next-step experiments

NASA Astrophysics Data System (ADS)

Greenfield, C. M.; Murakami, M.; Ferron, J. R.; Wade, M. R.; Luce, T. C.; Petty, C. C.; Menard, J. E.; Petrie, T. W.; Allen, S. L.; Burrell, K. H.; Casper, T. A.; DeBoo, J. C.; Doyle, E. J.; Garofalo, A. M.; Gorelov, I. A.; Groebner, R. J.; Hobirk, J.; Hyatt, A. W.; Jayakumar, R. J.; Kessel, C. E.; La Haye, R. J.; Jackson, G. L.; Lohr, J.; Makowski, M. A.; Pinsker, R. I.; Politzer, P. A.; Prater, R.; Strait, E. J.; Taylor, T. S.; West, W. P.; DIII-D Team

2004-05-01

Advanced Tokamak (AT) research in DIII-D [K. H. Burrell for the DIII-D Team, in Proceedings of the 19th Fusion Energy Conference, Lyon, France, 2002 (International Atomic Energy Agency, Vienna, 2002) published on CD-ROM] seeks to provide a scientific basis for steady-state high performance operation in future devices. These regimes require high toroidal beta to maximize fusion output and poloidal beta to maximize the self-driven bootstrap current. Achieving these conditions requires integrated, simultaneous control of the current and pressure profiles, and active magnetohydrodynamic stability control. The building blocks for AT operation are in hand. Resistive wall mode stabilization via plasma rotation and active feedback with nonaxisymmetric coils allows routine operation above the no-wall beta limit. Neoclassical tearing modes are stabilized by active feedback control of localized electron cyclotron current drive (ECCD). Plasma shaping and profile control provide further improvements. Under these conditions, bootstrap supplies most of the current. Steady-state operation requires replacing the remaining Ohmic current, mostly located near the half radius, with noninductive external sources. In DIII-D this current is provided by ECCD, and nearly stationary AT discharges have been sustained with little remaining Ohmic current. Fast wave current drive is being developed to control the central magnetic shear. Density control, with divertor cryopumps, of AT discharges with edge localized moding H-mode edges facilitates high current drive efficiency at reactor relevant collisionalities. A sophisticated plasma control system allows integrated control of these elements. Close coupling between modeling and experiment is key to understanding the separate elements, their complex nonlinear interactions, and their integration into self-consistent high performance scenarios. Progress on this development, and its implications for next-step devices, will be illustrated by results of recent experiment and simulation efforts.

High beta plasma operation in a toroidal plasma producing device

DOEpatents

Clarke, John F.

1978-01-01

A high beta plasma is produced in a plasma producing device of toroidal configuration by ohmic heating and auxiliary heating. The plasma pressure is continuously monitored and used in a control system to program the current in the poloidal field windings. Throughout the heating process, magnetic flux is conserved inside the plasma and the distortion of the flux surfaces drives a current in the plasma. As a consequence, the total current increases and the poloidal field windings are driven with an equal and opposing increasing current. The spatial distribution of the current in the poloidal field windings is determined by the plasma pressure. Plasma equilibrium is maintained thereby, and high temperature, high beta operation results.

Fusion reactor pumped laser

DOEpatents

Jassby, D.L.

1987-09-04

A nuclear pumped laser capable of producing long pulses of very high power laser radiation is provided. A toroidal fusion reactor provides energetic neutrons which are slowed down by a moderator. The moderated neutrons are converted to energetic particles capable of pumping a lasing medium. The lasing medium is housed in an annular cell surrounding the reactor. The cell includes an annular reflecting mirror at the bottom and an annular output window at the top. A neutron reflector is disposed around the cell to reflect escaping neutrons back into the cell. The laser radiation from the annular window is focused onto a beam compactor which generates a single coherent output laser beam. 10 figs.

Fusion reactor pumped laser

DOEpatents

Jassby, Daniel L.

1988-01-01

A nuclear pumped laser capable of producing long pulses of very high power laser radiation is provided. A toroidal fusion reactor provides energetic neutrons which are slowed down by a moderator. The moderated neutrons are converted to energetic particles capable of pumping a lasing medium. The lasing medium is housed in an annular cell surrounding the reactor. The cell includes an annular reflecting mirror at the bottom and an annular output window at the top. A neutron reflector is disposed around the cell to reflect escaping neutrons back into the cell. The laser radiation from the annular window is focused onto a beam compactor which generates a single coherent output laser beam.

GTC simulations of ion temperature gradient driven instabilities in W7-X and LHD stellarators

NASA Astrophysics Data System (ADS)

Wang, Hongyu

2017-10-01

We report GTC linear simulations of ion temperature gradient (ITG) instabilities in Wendelstein 7-X (W7-X) and Large Helical Device (LHD) stellarators. GTC has recently been updated to treat 3D equilibria by interfacing with MHD equilibrium code VMEC. GTC simulations of ITG have been carried out in both full torus and partial torus taking into account the toroidal periodicity of the stellarators. The effects of toroidal mode coupling on linear dispersions and mode structures in W7-X and LHD are studied. The mode structure in W7-X is more localized in the toroidal direction, and LHD is more extended in the toroidal direction and tokamak-like. Linear growth rates, real frequencies, and mode structures agree reasonably with results of EUTERPE simulations. In collaboration with I. Holod, J. Riemann, Z. Lin, J. Bao, L. Shi, S. Taimourzadeh, R. Kleiber, and M. Borchardt.

A feasibility study of developing toroidal tanks for a spinning spacecraft

NASA Technical Reports Server (NTRS)

Anderson, J. E.; Fester, D. A.

1973-01-01

A study was made to determine the feasibility of developing toroidal propellant tanks for a bipropellant (N204/MMH) propulsion system to be used in a proposed advanced Pioneer spin-stabilized vehicle intended for a Jupiter-orbiter and possibly a Saturn-orbiter mission. The rationale for considering the use of two toroidal tanks rather than the proposed use of four spherical tanks includes the belief that a more symmetrical distribution of propellant mass and a smaller variation in the position of the vehicle center-of-mass during propellant consumption would result, reducing requirements for attitude-control propellants, for balance weight, and for other weights associated with the dynamics of the spinning spacecraft. Results lead to the conclusion that a toroidal tank containing an effective, passive surface tension propellant acquisition device could be fabricated with available manufacturing methods and could be used interchangeably for either fuel or oxidizer.

Simultaneous feedback control of plasma rotation and stored energy on NSTX-U using neoclassical toroidal viscosity and neutral beam injection

NASA Astrophysics Data System (ADS)

Goumiri, I. R.; Rowley, C. W.; Sabbagh, S. A.; Gates, D. A.; Boyer, M. D.; Gerhardt, S. P.; Kolemen, E.; Menard, J. E.

2017-05-01

A model-based feedback system is presented enabling the simultaneous control of the stored energy through βn and the toroidal rotation profile of the plasma in National Spherical Torus eXperiment Upgrade device. Actuation is obtained using the momentum from six injected neutral beams and the neoclassical toroidal viscosity generated by applying three-dimensional magnetic fields. Based on a model of the momentum diffusion and torque balance, a feedback controller is designed and tested in closed-loop simulations using TRANSP, a time dependent transport analysis code, in predictive mode. Promising results for the ongoing experimental implementation of controllers are obtained.

Simultaneous feedback control of plasma rotation and stored energy on NSTX-U using neoclassical toroidal viscosity and neutral beam injection

PubMed Central

Goumiri, I. R.; Sabbagh, S. A.; Boyer, M. D.; Gerhardt, S. P.; Kolemen, E.; Menard, J. E.

2017-01-01

A model-based feedback system is presented enabling the simultaneous control of the stored energy through βn and the toroidal rotation profile of the plasma in National Spherical Torus eXperiment Upgrade device. Actuation is obtained using the momentum from six injected neutral beams and the neoclassical toroidal viscosity generated by applying three-dimensional magnetic fields. Based on a model of the momentum diffusion and torque balance, a feedback controller is designed and tested in closed-loop simulations using TRANSP, a time dependent transport analysis code, in predictive mode. Promising results for the ongoing experimental implementation of controllers are obtained. PMID:28435207

Simultaneous feedback control of plasma rotation and stored energy on NSTX-U using neoclassical toroidal viscosity and neutral beam injection

DOE PAGES

Goumiri, I. R.; Rowley, C. W.; Sabbagh, S. A.; ...

2017-02-23

In this study, a model-based feedback system is presented enabling the simultaneous control of the stored energy through β n and the toroidal rotation profile of the plasma in National Spherical Torus eXperiment Upgrade device. Actuation is obtained using the momentum from six injected neutral beams and the neoclassical toroidal viscosity generated by applying three-dimensional magnetic fields. Based on a model of the momentum diffusion and torque balance, a feedback controller is designed and tested in closed-loop simulations using TRANSP, a time dependent transport analysis code, in predictive mode. Promising results for the ongoing experimental implementation of controllers are obtained.

Sawtooth mitigation in 3D MHD tokamak modelling with applied magnetic perturbations

NASA Astrophysics Data System (ADS)

Bonfiglio, D.; Veranda, M.; Cappello, S.; Chacón, L.; Escande, D. F.

2017-01-01

The effect of magnetic perturbations (MPs) on the sawtoothing dynamics of the internal kink mode in the tokamak is discussed in the framework of nonlinear 3D MHD modelling. Numerical simulations are performed with the pixie3d code (Chacón 2008 Phys. Plasmas 15 056103) based on a D-shaped configuration in toroidal geometry. MPs are applied as produced by two sets of coils distributed along the toroidal direction, one set located above and the other set below the outboard midplane, like in experimental devices such as DIII-D and ASDEX Upgrade. The capability of n  =  1 MPs to affect quasi-periodic sawteeth is shown to depend on the toroidal phase difference Δ φ between the perturbations produced by the two sets of coils. In particular, sawtooth mitigation is obtained for the Δ φ =π phasing, whereas no significant effect is observed for Δ φ =0 . Numerical findings are explained by the interplay between different poloidal harmonics in the spectrum of applied MPs, and appear to be consistent with experiments performed in the DIII-D device. Sawtooth mitigation and stimulation of self-organized helical states by applied MPs have been previously demonstrated in both circular tokamak and reversed-field pinch (RFP) experiments in the RFX-mod device, and in related 3D MHD modelling.

Thermal island destabilization and the Greenwald limit

DOE PAGES

White, R. B.; Gates, D. A.; Brennan, D. P.

2015-02-24

Magnetic reconnection is ubiquitous in the magnetosphere, the solar corona, and in toroidal fusion research discharges. A magnetic island saturates at a width which produces a minimum in the magnetic energy of the configuration is evident in a fusion device. At saturation, the modified current density profile, a function of the flux in the island, is essentially flat, the growth rate proportional to the difference in the current at the O-point and the X-point. Furthermore, modification of the current density profile in the island interior causes a change in the island stability and additional growth or contraction of the saturatedmore » island. Because field lines in an island are isolated from the outside plasma, an island can heat or cool preferentially depending on the balance of Ohmic heating and radiation loss in the interior, changing the resistivity and hence the current in the island. A simple model of island destabilization due to radiation cooling of the island is constructed, and the effect of modification of the current within an island is calculated. In addition destabilization effect is described, and it is shown that a small imbalance of heating can lead to exponential growth of the island. A destabilized magnetic island near the plasma edge can lead to plasma loss, and because the radiation is proportional to plasma density and charge, this effect can cause an impurity dependent density limit.« less

Thermal island destabilization and the Greenwald limit

DOE Office of Scientific and Technical Information (OSTI.GOV)

White, R. B.; Gates, D. A.; Brennan, D. P.

Magnetic reconnection is ubiquitous in the magnetosphere, the solar corona, and in toroidal fusion research discharges. A magnetic island saturates at a width which produces a minimum in the magnetic energy of the configuration is evident in a fusion device. At saturation, the modified current density profile, a function of the flux in the island, is essentially flat, the growth rate proportional to the difference in the current at the O-point and the X-point. Furthermore, modification of the current density profile in the island interior causes a change in the island stability and additional growth or contraction of the saturatedmore » island. Because field lines in an island are isolated from the outside plasma, an island can heat or cool preferentially depending on the balance of Ohmic heating and radiation loss in the interior, changing the resistivity and hence the current in the island. A simple model of island destabilization due to radiation cooling of the island is constructed, and the effect of modification of the current within an island is calculated. In addition destabilization effect is described, and it is shown that a small imbalance of heating can lead to exponential growth of the island. A destabilized magnetic island near the plasma edge can lead to plasma loss, and because the radiation is proportional to plasma density and charge, this effect can cause an impurity dependent density limit.« less

Thermal island destabilization and the Greenwald limit

DOE Office of Scientific and Technical Information (OSTI.GOV)

White, R. B.; Gates, D. A.; Brennan, D. P.

Magnetic reconnection is ubiquitous in the magnetosphere, the solar corona, and in toroidal fusion research discharges. In a fusion device, a magnetic island saturates at a width which produces a minimum in the magnetic energy of the configuration. At saturation, the modified current density profile, a function of the flux in the island, is essentially flat, the growth rate proportional to the difference in the current at the O-point and the X-point. Further modification of the current density profile in the island interior causes a change in the island stability and additional growth or contraction of the saturated island. Becausemore » field lines in an island are isolated from the outside plasma, an island can heat or cool preferentially depending on the balance of Ohmic heating and radiation loss in the interior, changing the resistivity and hence the current in the island. A simple model of island destabilization due to radiation cooling of the island is constructed, and the effect of modification of the current within an island is calculated. An additional destabilization effect is described, and it is shown that a small imbalance of heating can lead to exponential growth of the island. A destabilized magnetic island near the plasma edge can lead to plasma loss, and because the radiation is proportional to plasma density and charge, this effect can cause an impurity dependent density limit.« less

Tearing mode dynamics and locking in the presence of external magnetic perturbations

NASA Astrophysics Data System (ADS)

Fridström, R.; Munaretto, S.; Frassinetti, L.; Chapman, B. E.; Brunsell, P. R.; Sarff, J. S.

2016-06-01

In normal operation, Madison Symmetric Torus (MST) [R. N. Dexter et al., Fusion Technol. 19, 131 (1991)] reversed-field pinch plasmas exhibit several rotating tearing modes (TMs). Application of a resonant magnetic perturbation (RMP) results in braking of mode rotation and, if the perturbation amplitude is sufficiently high, in a wall-locked state. The coils that produce the magnetic perturbation in MST give rise to RMPs with several toroidal harmonics. As a result, simultaneous deceleration of all modes is observed. The measured TM dynamics is shown to be in qualitative agreement with a magnetohydrodynamical model of the RMP interaction with the TM [R. Fitzpatrick, Nucl. Fusion 33, 1049 (1993)] adapted to MST. To correctly model the TM dynamics, the electromagnetic torque acting on several TMs is included. Quantitative agreement of the TM slowing-down time was obtained for a kinematic viscosity in the order of νki n≈10 -20 m2/s. Analysis of discharges with different plasma densities shows an increase of the locking threshold with increasing density. Modeling results show good agreement with the experimental trend, assuming a density-independent kinematic viscosity. Comparison of the viscosity estimates in this paper to those made previously with other techniques in MST plasmas suggests the possibility that the RMP technique may allow for estimates of the viscosity over a broad range of plasmas in MST and other devices.

Destabilization of counter-propagating TAEs by off-axis, co-current Neutral Beam Injection

NASA Astrophysics Data System (ADS)

Podesta', M.; Fredrickson, E.; Gorelenkova, M.

2017-10-01

Neutral Beam injection (NBI) is a common tool to heat the plasma and drive current non-inductively in fusion devices. Energetic particles (EP) resulting from NBI can drive instabilities that are detrimental for the performance and the predictability of plasma discharges. A broad NBI deposition profile, e.g. by off-axis injection aiming near the plasma mid-radius, is often assumed to limit those undesired effects by reducing the radial gradient of the EP density, thus reducing the ``universal'' drive for instabilities. However, this work presents new evidence that off-axis NBI can also lead to undesired effects such as the destabilization of Alfvénic instabilities, as observed in NSTX-U plasmas. Experimental observations indicate that counter propagating toroidal AEs are destabilized as the radial EP density profile becomes hollow as a result of off-axis NBI. Time-dependent analysis with the TRANSP code, augmented by a reduced fast ion transport model (known as kick model), indicates that instabilities are driven by a combination of radial and energy gradients in the EP distribution. Understanding the mechanisms for wave-particle interaction, revealed by the phase space resolved analysis, is the basis to identify strategies to mitigate or suppress the observed instabilities. Work supported by the U.S. Department of Energy, Office of Science, Office of Fusion Energy Sciences under Contract Number DE-AC02-09CH11466.

Advances in understanding quiescent H-mode plasmas in DIII-Da)

NASA Astrophysics Data System (ADS)

Burrell, K. H.; West, W. P.; Doyle, E. J.; Austin, M. E.; Casper, T. A.; Gohil, P.; Greenfield, C. M.; Groebner, R. J.; Hyatt, A. W.; Jayakumar, R. J.; Kaplan, D. H.; Lao, L. L.; Leonard, A. W.; Makowski, M. A.; McKee, G. R.; Osborne, T. H.; Snyder, P. B.; Solomon, W. M.; Thomas, D. M.; Rhodes, T. L.; Strait, E. J.; Wade, M. R.; Wang, G.; Zeng, L.

2005-05-01

Recent QH-mode research on DIII-D [J. L. Luxon et al., Plasma Physics and Controlled Nuclear Fusion Research 1996 (International Atomic Energy Agency, Vienna, 1987), Vol. I, p. 159] has used the peeling-ballooning modes model of edge magnetohydrodynamic stability as a working hypothesis to organize the data; several predictions of this theory are consistent with the experimental results. Current ramping results indicate that QH modes operate near the edge current limit set by peeling modes. This operating point explains why QH mode is easier to get at lower plasma currents. Power scans have shown a saturation of edge pressure with increasing power input. This allows QH-mode plasmas to remain stable to edge localized modes (ELMs) to the highest powers used in DIII-D. At present, the mechanism for this saturation is unknown; if the edge harmonic oscillation (EHO) is playing a role here, the physics is not a simple amplitude dependence. The increase in edge stability with plasma triangularity predicted by the peeling-ballooning theory is consistent with the substantial improvement in pedestal pressure achieved by changing the plasma shape from a single null divertor to a high triangularity double null. Detailed ELITE calculations for the high triangularity plasmas have demonstrated that the plasma operating point is marginally stable to peeling-ballooning modes. Comparison of ELMing, coinjected and quiescent, counterinjected discharges with the same shape, current, toroidal field, electron density, and electron temperature indicates that the edge radial electric field or the edge toroidal rotation are also playing a role in edge stability. The EHO produces electron, main ion, and impurity particle transport at the plasma edge which is more rapid than that produced by ELMs under similar conditions. The EHO also decreases the edge rotation while producing little change in the edge electron and ion temperatures. Other edge electromagnetic modes also produce particle transport; this includes the incoherent, broadband activity seen at high triangularity. Pedestal values of ν* and βT bracketing, those required for International Experimental Thermonuclear Reactor [Nucl. Fusion 39, 2137 (1999)] have been achieved in DIII-D, demonstrating the QH-mode edge densities are sufficient for future devices.

Advances in understanding quiescent H-mode plasmas in DIII-D

DOE Office of Scientific and Technical Information (OSTI.GOV)

Burrell, K.H.; West, W.P.; Gohil, P.

2005-05-15

Recent QH-mode research on DIII-D [J. L. Luxon et al., Plasma Physics and Controlled Nuclear Fusion Research 1996 (International Atomic Energy Agency, Vienna, 1987), Vol. I, p. 159] has used the peeling-ballooning modes model of edge magnetohydrodynamic stability as a working hypothesis to organize the data; several predictions of this theory are consistent with the experimental results. Current ramping results indicate that QH modes operate near the edge current limit set by peeling modes. This operating point explains why QH mode is easier to get at lower plasma currents. Power scans have shown a saturation of edge pressure with increasingmore » power input. This allows QH-mode plasmas to remain stable to edge localized modes (ELMs) to the highest powers used in DIII-D. At present, the mechanism for this saturation is unknown; if the edge harmonic oscillation (EHO) is playing a role here, the physics is not a simple amplitude dependence. The increase in edge stability with plasma triangularity predicted by the peeling-ballooning theory is consistent with the substantial improvement in pedestal pressure achieved by changing the plasma shape from a single null divertor to a high triangularity double null. Detailed ELITE calculations for the high triangularity plasmas have demonstrated that the plasma operating point is marginally stable to peeling-ballooning modes. Comparison of ELMing, coinjected and quiescent, counterinjected discharges with the same shape, current, toroidal field, electron density, and electron temperature indicates that the edge radial electric field or the edge toroidal rotation are also playing a role in edge stability. The EHO produces electron, main ion, and impurity particle transport at the plasma edge which is more rapid than that produced by ELMs under similar conditions. The EHO also decreases the edge rotation while producing little change in the edge electron and ion temperatures. Other edge electromagnetic modes also produce particle transport; this includes the incoherent, broadband activity seen at high triangularity. Pedestal values of {nu}{sub *} and {beta}{sub T} bracketing, those required for International Experimental Thermonuclear Reactor [Nucl. Fusion 39, 2137 (1999)] have been achieved in DIII-D, demonstrating the QH-mode edge densities are sufficient for future devices.« less

21 CFR 888.3080 - Intervertebral body fusion device.

Code of Federal Regulations, 2011 CFR

2011-04-01

... 21 Food and Drugs 8 2011-04-01 2011-04-01 false Intervertebral body fusion device. 888.3080 Section 888.3080 Food and Drugs FOOD AND DRUG ADMINISTRATION, DEPARTMENT OF HEALTH AND HUMAN SERVICES (CONTINUED) MEDICAL DEVICES ORTHOPEDIC DEVICES Prosthetic Devices § 888.3080 Intervertebral body fusion...

Spectroscopic Measurement of Ion Flow During Merging Start-up of Field-Reversed Configuration

NASA Astrophysics Data System (ADS)

Oka, Hirotaka; Inomoto, Michiaki; Tanabe, Hiroshi; Annoura, Masanobu; Ono, Yasushi; Nemoto, Koshichi

2012-10-01

The counter-helicity merging method [1] of field-reversed configuration (FRC) formation involves generation of bidirectional toroidal flow, known as a ``sling-shot.'' In two fluids regime, reconnection process is strongly affected by the Hall effect [2]. In this study, we have investigated the behavior of toroidal bidirectional flow generated by the counter-helicity merging in two-fluids regime. We use 2D Ion Doppler Spectroscopy to mesure toroidal ion flow during merging start-up of FRC from Ar gas. We defined two cases: one case with a radially pushed-in X line (case I) and the other case with a radially pushed-out X line(case O). The flow during the plasma merging shows radial asymmetry, as expected from the magnetic measurement, but finally relaxes to a unidirectional flow in plasma current direction in both cases. We observed larger toroidal flow in the plasma current direction in case I after FRC is formed, though the FRC in case O has larger magnetic flux. These results suggest that more ions are lost during merging start-up in case I. This selective ion loss might account for stability and confinement of FRCs probably maintained by high energy ions.[4pt] [1] Y. Ono, et al., Nucl. Fusion 39, pp. 2001-2008 (1999).[0pt] [2] M. Inomoto, et al., Phys. Rev. Lett., 97, 135002, (2006)

Developing a compact toroid injector in the ThermoElectric driven Liquid metal plasma facing Structures device

NASA Astrophysics Data System (ADS)

Christenson, Michael; Szott, Matthew; Kalathiparambil, Kishor; Sovinec, Carl; Ruzic, David

2016-10-01

The ThermoElectric-driven Liquid-metal plasma-facing Structures (TELS) device at the University of Illinois is a theta-pinched, plasma-material interaction test stand used to simulate extreme events in the edge and divertor regions of a tokamak plasma. Previous measurements of the electron and ion temperatures have shown that the isotropic heat load on target ranges between 0.1 and 0.2 MJ m-2 over a pulse lasting 0.2 ms. While this compares well to the heat loads from Type 1 ELMs in larger toroidal devices, it is still much less than the energy deposition from Type 1 ELMs expected in ITER, which are in excess of 1 MJ m-2. To this end, a compact toroid (CT) injector has been proposed as a modification to the existing TELS device. By using an externally applied bias field to force reconnection at the muzzle of the coaxial plasma accelerator source that drives ionization, NIMROD MHD simulations have shown a peak magnetic flux of 3.5 mWb is reached 0.025 ms into the pulse - more than sufficient to form a CT. Early calorimetry and magnetic field measurements indicate that a new plasma structure has been formed in the magnetized coaxial plasma source. This work presents the current results of CT generation with respect to the bias field strength as well as the coaxial source geometry. DOE OFES DE-SC0008587, DE-SC0008658, DE-FG02-99ER54515.

Ion flow measurements during the MHD relaxation processes in the HIST spherical torus device

NASA Astrophysics Data System (ADS)

Nishioka, T.; Hashimoto, S.; Ando, K.; Kikuchi, Y.; Fukumoto, N.; Nagata, M.

2008-11-01

Plasma flow is one of the key roles in self-organization and magnetic reconnection processes of helicity-driven spherical torus (ST) and spheromak. The HIST spherical torus can form the standard ST and the flipped ST plasmas by utilizing the variation of the external toroidal field coil current. The flipped ST plasma can be generated by changing the polarity of the toroidal magnetic field during the standard ST discharge [1]. We have developed an ion Doppler spectrometer (IDS) system using a compact 16 channel photomultiplier tube (PMT) in order to measure the spatial profile of ion temperature and rotation velocity in the HIST device. The IDS system consists of a light collection system including optical fibers, 1 m-spectrometer and the PMT detector. As the results, it was observed that ion velocity was about 10 km/s in the same direction as the toroidal current and ExB direction in the standard ST discharge. The observed ion velocity agrees with Mach probe measurements. During the transition from the standard ST to the flipped ST state, the ion temperature was fluctuated and increased. The result implies an ion heating during magnetic reconnections. In addition, the toroidal direction of the ion flow was reversed. The detail physics of the observed phenomenon will be shown. [1] M. Nagata et al., Phys Rev. Lett. 90, pp. 225001-225004 (2003).

Elmo bumpy square plasma confinement device

DOEpatents

Owen, L.W.

1985-01-01

The invention is an Elmo bumpy type plasma confinement device having a polygonal configuration of closed magnet field lines for improved plasma confinement. In the preferred embodiment, the device is of a square configuration which is referred to as an Elmo bumpy square (EBS). The EBS is formed by four linear magnetic mirror sections each comprising a plurality of axisymmetric assemblies connected in series and linked by 90/sup 0/ sections of a high magnetic field toroidal solenoid type field generating coils. These coils provide corner confinement with a minimum of radial dispersion of the confined plasma to minimize the detrimental effects of the toroidal curvature of the magnetic field. Each corner is formed by a plurality of circular or elliptical coils aligned about the corner radius to provide maximum continuity in the closing of the magnetic field lines about the square configuration confining the plasma within a vacuum vessel located within the various coils forming the square configuration confinement geometry.

MHD Effects of a Ferritic Wall on Tokamak Plasmas

NASA Astrophysics Data System (ADS)

Hughes, Paul E.

It has been recognized for some time that the very high fluence of fast (14.1MeV) neutrons produced by deuterium-tritium fusion will represent a major materials challenge for the development of next-generation fusion energy projects such as a fusion component test facility and demonstration fusion power reactor. The best-understood and most promising solutions presently available are a family of low-activation steels originally developed for use in fission reactors, but the ferromagnetic properties of these steels represent a danger to plasma confinement through enhancement of magnetohydrodynamic instabilities and increased susceptibility to error fields. At present, experimental research into the effects of ferromagnetic materials on MHD stability in toroidal geometry has been confined to demonstrating that it is still possible to operate an advanced tokamak in the presence of ferromagnetic components. In order to better quantify the effects of ferromagnetic materials on tokamak plasma stability, a new ferritic wall has been installated in the High Beta Tokamak---Extended Pulse (HBT-EP) device. The development, assembly, installation, and testing of this wall as a modular upgrade is described, and the effect of the wall on machine performance is characterized. Comparative studies of plasma dynamics with the ferritic wall close-fitting against similar plasmas with the ferritic wall retracted demonstrate substantial effects on plasma stability. Resonant magnetic perturbations (RMPs) are applied, demonstrating a 50% increase in n = 1 plasma response amplitude when the ferritic wall is near the plasma. Susceptibility of plasmas to disruption events increases by a factor of 2 or more with the ferritic wall inserted, as disruptions are observed earlier with greater frequency. Growth rates of external kink instabilities are observed to be twice as large in the presence of a close-fitting ferritic wall. Initial studies are made of the influence of mode rotation frequency on the ferritic effect, as well as observations of the effect of the ferritic wall on disruption halo currents.

Ohmic ITBs in Alcator C-Mod

NASA Astrophysics Data System (ADS)

Rowan, William L.; Bespamyatnov, Igor O.; Fiore, C. L.; Dominguez, A.; Hubbard, A. E.; Ince-Cushman, A.; Greenwald, M. J.; Lin, L.; Marmar, E. S.; Reinke, M.; Rice, J. E.; Zhurovich, K.

2007-11-01

Internal transport barrier (ITB) plasmas can arise spontaneously in Ohmic Alcator C-Mod plasmas. The operational prescription for the ITB include formation of an EDA H-mode in a toroidal magnetic field that is ramping down and a subsequent increase in the toroidal magnetic field. Like ITBs generated with off-axis ICRF heating, these have peaked pressure profiles which can be suppressed by on-axis ICRF heating. Recent work on onset conditions for the ICRF generated ITB (K. Zhurovich, et al., To be published in Nuclear Fusion) demonstrates that the broadening of the ion temperature profile due to off-axis ICRF reduces the ion temperature gradient and suppreses the ITG instability driven particle flux as the primary mechanism for ITB formation. The object of this study is to examine the characteristics of Ohmic ITBs to find whether this model for onset is supported.

Turbulent equipartition pinch of toroidal momentum in spherical torus

NASA Astrophysics Data System (ADS)

Hahm, T. S.; Lee, J.; Wang, W. X.; Diamond, P. H.; Choi, G. J.; Na, D. H.; Na, Y. S.; Chung, K. J.; Hwang, Y. S.

2014-12-01

We present a new analytic expression for turbulent equipartition (TEP) pinch of toroidal angular momentum originating from magnetic field inhomogeneity of spherical torus (ST) plasmas. Starting from a conservative modern nonlinear gyrokinetic equation (Hahm et al 1988 Phys. Fluids 31 2670), we derive an expression for pinch to momentum diffusivity ratio without using a usual tokamak approximation of B ∝ 1/R which has been previously employed for TEP momentum pinch derivation in tokamaks (Hahm et al 2007 Phys. Plasmas 14 072302). Our new formula is evaluated for model equilibria of National Spherical Torus eXperiment (NSTX) (Ono et al 2001 Nucl. Fusion 41 1435) and Versatile Experiment Spherical Torus (VEST) (Chung et al 2013 Plasma Sci. Technol. 15 244) plasmas. Our result predicts stronger inward pinch for both cases, as compared to the prediction based on the tokamak formula.

Nonlinear simulations with and computational issues for NIMROD

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sovinec, C R

The NIMROD (Non-Ideal Magnetohydrodynamics with Rotation, Open Discussion) code development project was commissioned by the US Department of Energy in February, 1996 to provide the fusion research community with a computational tool for studying low-frequency behavior in experiments. Specific problems of interest include the neoclassical evolution of magnetic islands and the nonlinear behavior of tearing modes in the presence of rotation and nonideal walls in tokamaks; they also include topics relevant to innovative confinement concepts such as magnetic turbulence. Besides having physics models appropriate for these phenomena, an additional requirement is the ability to perform the computations in realistic geometries.more » The NIMROD Team is using contemporary management and computational methods to develop a computational tool for investigating low-frequency behavior in plasma fusion experiments. The authors intend to make the code freely available, and are taking steps to make it as easy to learn and use as possible. An example application for NIMROD is the nonlinear toroidal RFP simulation--the first in a series to investigate how toroidal geometry affects MHD activity in RFPs. Finally, the most important issue facing the project is execution time, and they are exploring better matrix solvers and a better parallel decomposition to address this.« less

Articulated limiter blade for a tokamak fusion reactor

DOEpatents

Doll, D.W.

1982-10-21

A limiter blade for a large tokomak fusion reactor includes three articulated blade sections for enabling the limiter blade to be adjusted for plasmas of different sizes. Each blade section is formed of a rigid backing plate carrying graphite tiles coated with titanium carbide, and the limiter blade forms a generally elliptic contour in both the poloidal and toroidal directions to uniformly distribute the heat flow to the blade. The limiter blade includes a central blade section movable along the major radius of the vacuum vessel, and upper and lower pivotal blade sections which may be pivoted by linear actuators having rollers held to the back surface of the pivotal blade sections.

Articulated limiter blade for a tokamak fusion reactor

DOEpatents

Doll, David W.

1985-01-01

A limiter blade for a large tokomak fusion reactor includes three articulated blade sections for enabling the limiter blade to be adjusted for plasmas of different sizes. Each blade section is formed of a rigid backing plate carrying graphite tiles coated with titanium carbide, and the limiter blade forms a generally elliptic contour in both the poloidal and toroidal directions to uniformly distribute the heat flow to the blade. The limiter blade includes a central blade section movable along the major radius of the vacuum vessel, and upper and lower pivotal blade sections which may be pivoted by linear actuators having rollers held to the back surface of the pivotal blade sections.

Device for plasma confinement and heating by high currents and nonclassical plasma transport properties

DOEpatents

Coppi, B.; Montgomery, D.B.

1973-12-11

A toroidal plasma containment device having means for inducing high total plasma currents and current densities and at the same time emhanced plasma heating, strong magnetic confinement, high energy density containment, magnetic modulation, microwaveinduced heating, and diagnostic accessibility is described. (Official Gazette)

Integrative Multi-Spectral Sensor Device for Far-Infrared and Visible Light Fusion

NASA Astrophysics Data System (ADS)

Qiao, Tiezhu; Chen, Lulu; Pang, Yusong; Yan, Gaowei

2018-06-01

Infrared and visible light image fusion technology is a hot spot in the research of multi-sensor fusion technology in recent years. Existing infrared and visible light fusion technologies need to register before fusion because of using two cameras. However, the application effect of the registration technology has yet to be improved. Hence, a novel integrative multi-spectral sensor device is proposed for infrared and visible light fusion, and by using the beam splitter prism, the coaxial light incident from the same lens is projected to the infrared charge coupled device (CCD) and visible light CCD, respectively. In this paper, the imaging mechanism of the proposed sensor device is studied with the process of the signals acquisition and fusion. The simulation experiment, which involves the entire process of the optic system, signal acquisition, and signal fusion, is constructed based on imaging effect model. Additionally, the quality evaluation index is adopted to analyze the simulation result. The experimental results demonstrate that the proposed sensor device is effective and feasible.

Pellet injection technology

NASA Astrophysics Data System (ADS)

Combs, S. K.

1993-07-01

During the last 10 to 15 years, significant progress has been made worldwide in the area of pellet injection technology. This specialized field of research originated as a possible solution to the problem of depositing atoms of fuel deep within magnetically confined, hot plasmas for refueling of fusion power reactors. Using pellet injection systems, frozen macroscopic (millimeter-size) pellets composed of the isotopes of hydrogen are formed, accelerated, and transported to the plasma for fueling. The process and benefits of plasma fueling by this approach have been demonstrated conclusively on a number of toroidal magnetic confinement configurations; consequently, pellet injection is the leading technology for deep fueling of magnetically confined plasmas for controlled thermonuclear fusion research. Hydrogen pellet injection devices operate at very low temperatures (≂10 K) at which solid hydrogen ice can be formed and sustained. Most injectors use conventional pneumatic (light gas gun) or centrifuge (mechanical) acceleration concepts to inject hydrogen or deuterium pellets at speeds of ≂1-2 km/s. Pellet injectors that can operate at quasi-steady state (pellet delivery rates of 1-40 Hz) have been developed for long-pulse fueling. The design and operation of injectors with the heaviest hydrogen isotope, tritium, offer some special problems because of tritium's radioactivity. To address these problems, a proof-of-principle experiment was carried out in which tritium pellets were formed and accelerated to speeds of 1.4 km/s. Tritium pellet injection is scheduled on major fusion research devices within the next few years. Several advanced accelerator concepts are under development to increase the pellet velocity. One of these is the two-stage light gas gun, for which speeds of slightly over 4 km/s have already been reported in laboratory experiments with deuterium ice. A few two-stage pneumatic systems (single-shot) have recently been installed on tokamak experiments. This article reviews the equipment and instruments that have been developed for pellet injection with emphasis on recent advances. Prospects for future development are addressed, as are possible applications of this technology to other areas of research.

Local properties of magnetic reconnection in nonlinear resistive- and extended-magnetohydrodynamic toroidal simulations of the sawtooth crash

DOE PAGES

Beidler, M. T.; Cassak, P. A.; Jardin, S. C.; ...

2016-12-15

We diagnose local properties of magnetic reconnection during a sawtooth crash employing the three-dimensional toroidal, extended-magnetohydrodynamic (MHD) code M3D-C 1. To do so, we sample simulation data in the plane in which reconnection occurs, the plane perpendicular to the helical (m, n) = (1, 1) mode at the q = 1 surface, where m and n are the poloidal and toroidal mode numbers and q is the safety factor. We study the nonlinear evolution of a particular test equilibrium in a non-reduced field representation using both resistive-MHD and extended-MHD models. We find growth rates for the extended-MHD reconnection process exhibitmore » a nonlinear acceleration and greatly exceed that of the resistive-MHD model, as is expected from previous experimental, theoretical, and computational work. We compare the properties of reconnection in the two simulations, revealing the reconnecting current sheets are locally different in the two models and we present the first observation of the quadrupole out-of-plane Hall magnetic field that appears during extended-MHD reconnection in a 3D toroidal simulation (but not in resistive-MHD). We also explore the dependence on toroidal angle of the properties of reconnection as viewed in the plane perpendicular to the helical magnetic field, finding qualitative and quantitative effects due to changes in the symmetry of the reconnection process. Furthermore, this study is potentially important for a wide range of magnetically confined fusion applications, from confirming simulations with extended-MHD effects are sufficiently resolved to describe reconnection, to quantifying local reconnection rates for purposes of understanding and predicting transport, not only at the q = 1 rational surface for sawteeth, but also at higher order rational surfaces that play a role in disruptions and edge-confinement degradation.« less

NIMROD Simulations of the HIT-SI and HIT-SI3 Devices

NASA Astrophysics Data System (ADS)

Morgan, Kyle; Jarboe, Tom; Hossack, Aaron; Chandra, Rian; Everson, Chris

2017-10-01

The Helicity Injected Torus with Steady Inductive helicity injection (HIT-SI) experiment uses a set of inductively driven helicity injectors to apply non-axisymmetric current drive on the edge of the plasma, driving an axisymmetric spheromak equilibrium in a central confinement volume. Significant improvements have been made to extended MHD modeling of HIT-SI, with both the resolution of disagreement at high injector frequencies in HIT-SI in addition to successes with the new upgraded HIT-SI3 device. Previous numerical studies of HIT-SI, using a zero-beta eMHD model, focused on operations with a drive frequency of 14.5 kHz, and found reduced agreement with both the magnetic profile and current amplification at higher frequencies (30-70 kHz). HIT-SI3 has three helicity injectors which are able to operate with different mode structures of perturbations through the different relative temporal phasing of the injectors. Simulations that allow for pressure gradients have been performed in the parameter regimes of both devices using the NIMROD code and show improved agreement with experimental results, most notably capturing the observed Shafranov-shift due to increased beta observed at higher finj in HIT-SI and the variety of toroidal perturbation spectra available in HIT-SI3. This material is based upon work supported by the U.S. Department of Energy, Office of Science, Office of Fusion Energy Sciences under Award Number DE-FG02- 96ER54361.

Demountable externally anchored low-stress magnet system and related method

DOEpatents

Powell, James; Hsieh, Shih-Yung; Lehner, John R.

1981-01-01

Toroidal field coils are interlaced with other toroidal structures and are operated under supercooled conditions. To facilitate demounting the toroidal field coils, which are supercooled, they are made in the form of connected segments constituting coils of polygonal form. The segments may be rectilinear in form, but some may also be U-shaped or L-shaped. The segments are detachable from one another and are supported in load relieving manner. Power devices are used to displace the segments to facilitate removal of the coils from the aforesaid toroidal structures and to provide for the accommodation of dimensional changes and stresses due to thermal and magnetic conditions. The segments are formed of spaced parallel conductive slabs with the slabs of one segment being interdigitated with the slabs of the adjacent segment. The interdigitated slabs may be soldered together or slidingly engaged. The slabs are shaped to accommodate superconductors and to provide passages for a cooling medium. The slabs are moreover separated by insulator slabs with which they form a coil structure which is jacketed.

Measurements of the toroidal torque balance of error field penetration locked modes

DOE PAGES

Shiraki, Daisuke; Paz-Soldan, Carlos; Hanson, Jeremy M.; ...

2015-01-05

Here, detailed measurements from the DIII-D tokamak of the toroidal dynamics of error field penetration locked modes under the influence of slowly evolving external fields, enable study of the toroidal torques on the mode, including interaction with the intrinsic error field. The error field in these low density Ohmic discharges is well known based on the mode penetration threshold, allowing resonant and non-resonant torque effects to be distinguished. These m/n = 2/1 locked modes are found to be well described by a toroidal torque balance between the resonant interaction with n = 1 error fields, and a viscous torque inmore » the electron diamagnetic drift direction which is observed to scale as the square of the perturbed field due to the island. Fitting to this empirical torque balance allows a time-resolved measurement of the intrinsic error field of the device, providing evidence for a time-dependent error field in DIII-D due to ramping of the Ohmic coil current.« less

Measuring time of flight of fusion products in an inertial electrostatic confinement fusion device for spatial profiling of fusion reactions

DOE Office of Scientific and Technical Information (OSTI.GOV)

Donovan, D. C.; Boris, D. R.; Kulcinski, G. L.

2013-03-15

A new diagnostic has been developed that uses the time of flight (TOF) of the products from a nuclear fusion reaction to determine the location where the fusion reaction occurred. The TOF diagnostic uses charged particle detectors on opposing sides of the inertial electrostatic confinement (IEC) device that are coupled to high resolution timing electronics to measure the spatial profile of fusion reactions occurring between the two charged particle detectors. This diagnostic was constructed and tested by the University of Wisconsin-Madison Inertial Electrostatic Confinement Fusion Group in the IEC device, HOMER, which accelerates deuterium ions to fusion relevant energies inmore » a high voltage ({approx}100 kV), spherically symmetric, electrostatic potential well [J. F. Santarius, G. L. Kulcinski, R. P. Ashley, D. R. Boris, B. B. Cipiti, S. K. Murali, G. R. Piefer, R. F. Radel, T. E. Radel, and A. L. Wehmeyer, Fusion Sci. Technol. 47, 1238 (2005)]. The TOF diagnostic detects the products of D(d,p)T reactions and determines where along a chord through the device the fusion event occurred. The diagnostic is also capable of using charged particle spectroscopy to determine the Doppler shift imparted to the fusion products by the center of mass energy of the fusion reactants. The TOF diagnostic is thus able to collect spatial profiles of the fusion reaction density along a chord through the device, coupled with the center of mass energy of the reactions occurring at each location. This provides levels of diagnostic detail never before achieved on an IEC device.« less

Extreme sensitive metasensor for targeted biomarkers identification using colloidal nanoparticles-integrated plasmonic unit cells

PubMed Central

Ahmadivand, Arash; Gerislioglu, Burak; Tomitaka, Asahi; Manickam, Pandiaraj; Kaushik, Ajeet; Bhansali, Shekhar; Nair, Madhavan; Pala, Nezih

2018-01-01

Engineered terahertz (THz) plasmonic metamaterials have emerged as promising platforms for quick infection diagnosis, cost-effective and real-time pharmacology applications owing to their non-destructive and harmless interaction with biological tissues in both in vivo and in vitro assays. As a recent member of THz metamaterials family, toroidal metamaterials have been demonstrated to be supporting high-quality sharp resonance modes. Here we introduce a THz metasensor based on a plasmonic surface consisting of metamolecules that support ultra-narrow toroidal resonances excited by the incident radiation and demonstrate detection of an ultralow concertation targeted biomarker. The toroidal plasmonic metasurface was designed and optimized through extensive numerical studies and fabricated by standard microfabrication techniques. The surface then functionalized by immobilizing the antibody for virus-envelope proteins (ZIKV-EPs) for selective sensing. We sensed and quantified the ZIKV-EP in the assays by measuring the spectral shifts of the toroidal resonances while varying the concentration. In an improved protocol, we introduced gold nanoparticles (GNPs) decorated with the same antibodies onto the metamolecules and monitored the resonance shifts for the same concentrations. Our studies verified that the presence of GNPs enhances capturing of biomarker molecules in the surrounding medium of the metamaterial. By measuring the shift of the toroidal dipolar momentum (up to Δω~0.35 cm−1) for different concentrations of the biomarker proteins, we analyzed the sensitivity, repeatability, and limit of detection (LoD) of the proposed toroidal THz metasensor. The results show that up to 100-fold sensitivity enhancement can be obtained by utilizing plasmonic nanoparticles-integrated toroidal metamolecules in comparison to analogous devices. This approach allows for detection of low molecular-weight biomolecules (≈13 kDa) in diluted solutions using toroidal THz plasmonic unit cells. PMID:29552379

Extreme sensitive metasensor for targeted biomarkers identification using colloidal nanoparticles-integrated plasmonic unit cells.

PubMed

Ahmadivand, Arash; Gerislioglu, Burak; Tomitaka, Asahi; Manickam, Pandiaraj; Kaushik, Ajeet; Bhansali, Shekhar; Nair, Madhavan; Pala, Nezih

2018-02-01

Engineered terahertz (THz) plasmonic metamaterials have emerged as promising platforms for quick infection diagnosis, cost-effective and real-time pharmacology applications owing to their non-destructive and harmless interaction with biological tissues in both in vivo and in vitro assays. As a recent member of THz metamaterials family, toroidal metamaterials have been demonstrated to be supporting high-quality sharp resonance modes. Here we introduce a THz metasensor based on a plasmonic surface consisting of metamolecules that support ultra-narrow toroidal resonances excited by the incident radiation and demonstrate detection of an ultralow concertation targeted biomarker. The toroidal plasmonic metasurface was designed and optimized through extensive numerical studies and fabricated by standard microfabrication techniques. The surface then functionalized by immobilizing the antibody for virus-envelope proteins (ZIKV-EPs) for selective sensing. We sensed and quantified the ZIKV-EP in the assays by measuring the spectral shifts of the toroidal resonances while varying the concentration. In an improved protocol, we introduced gold nanoparticles (GNPs) decorated with the same antibodies onto the metamolecules and monitored the resonance shifts for the same concentrations. Our studies verified that the presence of GNPs enhances capturing of biomarker molecules in the surrounding medium of the metamaterial. By measuring the shift of the toroidal dipolar momentum (up to Δ ω ~0.35 cm -1 ) for different concentrations of the biomarker proteins, we analyzed the sensitivity, repeatability, and limit of detection (LoD) of the proposed toroidal THz metasensor. The results show that up to 100-fold sensitivity enhancement can be obtained by utilizing plasmonic nanoparticles-integrated toroidal metamolecules in comparison to analogous devices. This approach allows for detection of low molecular-weight biomolecules (≈13 kDa) in diluted solutions using toroidal THz plasmonic unit cells.

Modeling the Compression of Merged Compact Toroids by Multiple Plasma Jets

NASA Technical Reports Server (NTRS)

Thio, Y. C. Francis; Knapp, Charles E.; Kirkpatrick, Ron; Rodgers, Stephen L. (Technical Monitor)

2000-01-01

A fusion propulsion scheme has been proposed that makes use of the merging of a spherical distribution of plasma jets to dynamically form a gaseous liner. The gaseous liner is used to implode a magnetized target to produce the fusion reaction in a standoff manner. In this paper, the merging of the plasma jets to form the gaseous liner is investigated numerically. The Los Alamos SPHINX code, based on the smoothed particle hydrodynamics method is used to model the interaction of the jets. 2-D and 3-D simulations have been performed to study the characteristics of the resulting flow when these jets collide. The results show that the jets merge to form a plasma liner that converge radially which may be used to compress the central plasma to fusion conditions. Details of the computational model and the SPH numerical methods will be presented together with the numerical results.

Comparison of measured impurity poloidal rotation in DIII-D with neoclassical predictions under low toroidal field conditions

DOE PAGES

Burrell, Keith H.; Grierson, Brian A.; Solomon, Wayne M.; ...

2014-06-26

Here, predictive understanding of plasma transport is a long-term goal of fusion research. This requires testing models of plasma rotation including poloidal rotation. The present experiment was motivated by recent poloidal rotation measurements on spherical tokamaks (NSTX and MAST) which showed that the poloidal rotation of C +6 is much closer to the neoclassical prediction than reported results in larger aspect ratio machines such as TFTR, DIII-D, JT-60U and JET working at significantly higher toroidal field and ion temperature. We investigated whether the difference in aspect ratio (1.44 on NSTX versus 2.7 on DIII-D) could explain this. We measured Cmore » +6 poloidal rotation in DIII-D under conditions which matched, as best possible, those in the NSTX experiment; we matched plasma current (0.65 MA), on-axis toroidal field (0.55T), minor radius (0.6 m), and outer flux surface shape as well as the density and temperature profiles. DIII-D results from this work also show reasonable agreement with neoclassical theory. Accordingly, the different aspect ratio does not explain the previously mentioned difference in poloidal rotation results.« less

Improved confinement region without large magnetohydrodynamic activity in TPE-RX reversed-field pinch plasma

DOE Office of Scientific and Technical Information (OSTI.GOV)

Yambe, Kiyoyuki; Hirano, Yoichi; Sakakita, Hajime

2014-11-15

We found that spontaneous improved confinement was brought about depending on the operating region in the Toroidal Pinch Experiment-Reversed eXperiment (TPE-RX) reversed-field pinch plasma [Y. Yagi et al., Fusion Eng. Des. 45, 421 (1999)]. Gradual decay of the toroidal magnetic field at plasma surface B{sub tw} reversal makes it possible to realize a prolonged discharge, and the poloidal beta value and energy confinement time increase in the latter half of the discharge, where reversal and pinch parameters become shallow and low, respectively. In the latter half of the discharge, the plasma current and volume-averaged toroidal magnetic field 〈B{sub t}〉 increasemore » again, the electron density slowly decays, the electron temperature and soft X-ray radiation intensity increase, and the magnetic fluctuations are markedly reduced. In this period of improved confinement, the value of (〈B{sub t}〉-B{sub tw})/B{sub pw}, where B{sub pw} is the poloidal magnetic field at the plasma surface, stays almost constant, which indicates that the dynamo action occurs without large magnetohydrodynamic activities.« less

Experimental Design of a Magnetic Flux Compression Experiment

NASA Astrophysics Data System (ADS)

Fuelling, Stephan; Awe, Thomas J.; Bauer, Bruno S.; Goodrich, Tasha; Lindemuth, Irvin R.; Makhin, Volodymyr; Siemon, Richard E.; Atchison, Walter L.; Reinovsky, Robert E.; Salazar, Mike A.; Scudder, David W.; Turchi, Peter J.; Degnan, James H.; Ruden, Edward L.

2007-06-01

Generation of ultrahigh magnetic fields is an interesting topic of high-energy-density physics, and an essential aspect of Magnetized Target Fusion (MTF). To examine plasma formation from conductors impinged upon by ultrahigh magnetic fields, in a geometry similar to that of the MAGO experiments, an experiment is under design to compress magnetic flux in a toroidal cavity, using the Shiva Star or Atlas generator. An initial toroidal bias magnetic field is provided by a current on a central conductor. The central current is generated by diverting a fraction of the liner current using an innovative inductive current divider, thus avoiding the need for an auxiliary power supply. A 50-mm-radius cylindrical aluminum liner implodes along glide planes with velocity of about 5 km/s. Inward liner motion causes electrical closure of the toroidal chamber, after which flux in the chamber is conserved and compressed, yielding magnetic fields of 2-3 MG. Plasma is generated on the liner and central rod surfaces by Ohmic heating. Diagnostics include B-dot probes, Faraday rotation, radiography, filtered photodiodes, and VUV spectroscopy. Optical access to the chamber is provided through small holes in the walls.

Free boundary resistive modes in tokamaks

DOE Office of Scientific and Technical Information (OSTI.GOV)

Huysmans, G.T.A.; Goedbloed, J.P.; Kerner, W.

1993-05-01

There exist a number of observations of magnetohydrodynamic (MHD) activity that can be related to resistive MHD modes localized near the plasma boundary. To study the stability of these modes, a free boundary description of the plasma is essential. The resistive plasma--vacuum boundary conditions have been implemented in the fully toroidal resistive spectral code CASTOR (Complex Alfven Spectrum in Toroidal Geometry) [[ital Proceedings] [ital of] [ital the] 18[ital th] [ital Conference] [ital on] [ital Controlled] [ital Fusion] [ital and] [ital Plasma] [ital Physics], Berlin, edited by P. Bachmann and D. C. Robinson (European Physical Society, Petit-Lancy, Switzerland, 1991), p. 89].more » The influence of a free boundary, as compared to a fixed boundary on the stability of low-[ital m] tearing modes, is studied. It is found that the stabilizing (toroidal) effect of a finite pressure due the plasma compression is lost in the free boundary case for modes localized near the boundary. Since the stabilization due to the favorable average curvature in combination with a pressure gradient near the boundary is small, the influence of the pressure on the stability is much less important for free boundary modes than for fixed boundary modes.« less

Neoclassical Toroidal Viscosity Torque Induced by Plasma Response in a Low- β Tokamak with Edge Pedestal

NASA Astrophysics Data System (ADS)

Yan, Xingting; Zhu, Ping; Sun, Youwen

2016-10-01

The characteristic profile and magnitude are predicted in theory for the neoclassical toroidal viscosity (NTV) torque induced by the plasma response to the resonant magnetic perturbation (RMP) in a tokamak with an edge pedestal, using the newly developed module coupling the NIMROD and the NTVTOK codes. For a low β equilibrium, the NTV torque is mainly induced by the dominant toroidal mode of plasma response. The NTV torque profile is radially localized and peaked, which is determined by profiles of both the equilibrium temperature and the plasma response fields. In general, the peak of NTV torque profile is found to trace the pedestal location. The magnitude of NTV torque is extremely sensitive to the β of pedestal top; for a given plasma response, the peak value of NTV torque can increase by three orders of magnitude, when the pedestal β increases by only one order of magnitude. This suggests a more significant role of NTV torque in higher plasma β regimes. Supported by the National Magnetic Confinement Fusion Program of China under Grant Nos. 2014GB124002 and 2015GB101004, and the 100 Talent Program of the Chinese Academy of Sciences.

Two-fluid flowing equilibria of spherical torus sustained by coaxial helicity injection

NASA Astrophysics Data System (ADS)

Kanki, Takashi; Steinhauer, Loren; Nagata, Masayoshi

2007-11-01

Two-dimensional equilibria in helicity-driven systems using two-fluid model were previously computed, showing the existence of an ultra-low-q spherical torus (ST) configuration with diamagnetism and higher beta. However, this computation assumed purely toroidal ion flow and uniform density. The purpose of the present study is to apply the two-fluid model to the two-dimensional equilibria of helicity-driven ST with non-uniform density and both toroidal and poloidal flows for each species by means of the nearby-fluids procedure, and to explore their properties. We focus our attention on the equilibria relevant to the HIST device, which are characterized by either driven or decaying λ profiles. The equilibrium for the driven λ profile has a diamagnetic toroidal field, high-β (βt = 32%), and centrally broad density. By contrast, the decaying equilibrium has a paramagnetic toroidal field, low-β (βt = 10%), and centrally peaked density with a steep gradient in the outer edge region. In the driven case, the toroidal ion and electron flows are in the same direction, and two-fluid effects are less important since the ExB drift is dominant. In the decaying case, the toroidal ion and electron flows are opposite in the outer edge region, and two-fluid effects are significant locally in the edge due to the ion diamagnetic drift.

Electron Temperature and Density in Local Helicity Injection and High betat Plasmas

NASA Astrophysics Data System (ADS)

Schlossberg, David J.

Tokamak startup in a spherical torus (ST) and an ST-based fusion nuclear science facility can greatly benefit from using non-inductive methods. The Pegasus Toroidal Experiment has developed a non-inductive startup technique using local helicity injection (LHI). Electron temperature, T e(r), and density, ne( r), profiles during LHI are unknown. These profiles are critical for understanding both the physics of the injection and relaxation mechanisms, as well as for extrapolating this technique to larger devices. A new Thomson scattering system has been designed, installed, and used to characterize Te(r, t) and ne(r, t) during LHI. The diagnostic leverages new technology in image intensified CCD cameras, high-efficiency diffraction gratings, and reliable Nd:YAG lasers. Custom systems for stray light mitigation, fast shuttering, and precision timing have been developed and implemented. The overall system provides a low-maintenance, economic, and effective means to explore novel physics regimes in Pegasus. Electron temperature and density profiles during LHI have been measured for the first time. Results indicate Te(r) peaked in the core of plasmas, and sustained while plasmas are coupled to injection drive. Electron densities also peak near the core of the tokamak, up to local values of n e ˜ 1.5 x 1019 m -3. A comparison of Te( r, t) has been made between discharges with dominant drive voltage from induction versus helicity injection. In both cases Te ( r, t) profiles remain peaked, with values for Te ,max > 150 eV in dominantly helicity-driven plasmas using high-field side LHI. Sustained values of betat ˜ 100% have been demonstrated in a tokamak for the first time. Plasmas are created and driven entirely non-solenoidally, and exhibit MHD stability. Measured temperature and density profiles are used to constrain magnetic equilibrium reconstructions, which calculate 80% < betat < 100% throughout a toroidal field ramp-down. For a continued decrease in the toroidal field these plasmas disrupt near the ideal MHD no-wall stability limit predicted by the DCON code. Mode analyses of predicted and measured MHD agree, and suggest discharges terminate by an intermediate-m, n=1 external mode. A localized region of minimum |B| has been identified in these discharges, and modeling shows access to it depends on both plasma pressure and magnetic geometry. This magnetic well is shown to persist over several milliseconds, in both constant toroidal field and ramp-down cases.

Low-temperature tensile strength of the ITER-TF model coil insulation system after reactor irradiation

NASA Astrophysics Data System (ADS)

Bittner-Rohrhofer, K.; Humer, K.; Weber, H. W.

The windings of the superconducting magnet coils for the ITER-FEAT fusion device are affected by high mechanical stresses at cryogenic temperatures and by a radiation environment, which impose certain constraints especially on the insulating materials. A glass fiber reinforced plastic (GFRP) laminate, which consists of Kapton/R-glass-fiber reinforcement tapes, vacuum-impregnated in a DGEBA epoxy system, was used for the European toroidal field model coil turn insulation of ITER. In order to assess its mechanical properties under the actual operating conditions of ITER-FEAT, cryogenic (77 K) static tensile tests and tension-tension fatigue measurements were done before and after irradiation to a fast neutron fluence of 1×10 22 m -2 ( E>0.1 MeV), i.e. the ITER-FEAT design fluence level. We find that the mechanical strength and the fracture behavior of this GFRP are strongly influenced by the winding direction of the tape and by the radiation induced delamination process. In addition, the composite swells by 3%, forming bubbles inside the laminate, and loses weight (1.4%) at the design fluence.

The High Field Ultra Low Aspect Ratio Tokamak (HF-ULART)

NASA Astrophysics Data System (ADS)

Ribeiro, Celso

2017-10-01

Recently, a medium-size HF-ULART has been proposed. The major objective is to explore the high beta and pressure under the high toroidal field, using present day technology. This might be one of pathway scenarios for a potential ultra-compact pulsed neutron source (UCP-NS) based on the spherical tokamak (ST) concept, which may lead to more steady-state NS or even to a fusion reactor, via realistic design scaling. The HF-ULART pulsed mode operation is created by quasi-simultaneous adiabatic compression (AC) in both minor and major radius of a very high beta plasma, possibly with further help of passive-wall stabilization, as envisaged in the RULART concept. This may help the revival of the studies of the AC technique in tokamaks, alongside the less compact and more complex ST-40 device, currently under construction. In addition, by similarities, studies in HF-ULART as a UCP-NS may also help to test the feasibility of the compact NS via the spheromak concept, which also uses the AC technique. Simulations of AC in HF-ULART plasmas will be presented.

Neoclassical Simulation of Tokamak Plasmas using Continuum Gyrokinetc Code TEMPEST

DOE Office of Scientific and Technical Information (OSTI.GOV)

Xu, X Q

We present gyrokinetic neoclassical simulations of tokamak plasmas with self-consistent electric field for the first time using a fully nonlinear (full-f) continuum code TEMPEST in a circular geometry. A set of gyrokinetic equations are discretized on a five dimensional computational grid in phase space. The present implementation is a Method of Lines approach where the phase-space derivatives are discretized with finite differences and implicit backwards differencing formulas are used to advance the system in time. The fully nonlinear Boltzmann model is used for electrons. The neoclassical electric field is obtained by solving gyrokinetic Poisson equation with self-consistent poloidal variation. Withmore » our 4D ({psi}, {theta}, {epsilon}, {mu}) version of the TEMPEST code we compute radial particle and heat flux, the Geodesic-Acoustic Mode (GAM), and the development of neoclassical electric field, which we compare with neoclassical theory with a Lorentz collision model. The present work provides a numerical scheme and a new capability for self-consistently studying important aspects of neoclassical transport and rotations in toroidal magnetic fusion devices.« less

NIMROD Simulations of Low-q Disruptions in the Compact Toroidal Hybrid Device (CTH)

NASA Astrophysics Data System (ADS)

Howell, E. C.; Pandya, M. D.; Hanson, J. D.; Mauer, D. A.; Ennis, D. A.; Hartwell, G. J.

2016-10-01

Nonlinear MHD simulations of low-q disruptions in the CTH are presented. CTH is a current carrying stellarator that is used to study the effects of 3D shaping. The application of 3D shaping stabilizes low-q disruptions in CTH. The amount of 3D shaping is controlled by adjusting the external rotational transform, and it is characterized by the ratio of the external rotational transform to the total transform: f =ιvac / ι . Disruptions are routinely observed during operation with weak shaping (f < 0.05). The frequency of disruptions decreases with increasing amounts of 3D shaping, and the disruptions are completely suppressed for f > 0.1 . Nonlinear simulations are performed using the NIMROD code to better understand how the shaping suppresses the disruptions. Comparisons of runs with weak (f = 0.04) and strong (f = 0.10) shaping are shown. This material is based upon work supported by Auburn University and the U.S. Department of Energy, Office of Science, Office of Fusion Energy Sciences under Award Numbers DE-FG02-03ER54692 and DE-FG02-00ER54610.

Modeling and testing of ethernet transformers

NASA Astrophysics Data System (ADS)

Bowen, David

2011-12-01

Twisted-pair Ethernet is now the standard home and office last-mile network technology. For decades, the IEEE standard that defines Ethernet has required electrical isolation between the twisted pair cable and the Ethernet device. So, for decades, every Ethernet interface has used magnetic core Ethernet transformers to isolate Ethernet devices and keep users safe in the event of a potentially dangerous fault on the network media. The current state-of-the-art Ethernet transformers are miniature (<5mm diameter) ferrite-core toroids wrapped with approximately 10 to 30 turns of wire. As small as current Ethernet transformers are, they still limit further Ethernet device miniaturization and require a separate bulky package or jack housing. New coupler designs must be explored which are capable of exceptional miniaturization or on-chip fabrication. This dissertation thoroughly explores the performance of the current commercial Ethernet transformers to both increase understanding of the device's behavior and outline performance parameters for replacement devices. Lumped element and distributed circuit models are derived; testing schemes are developed and used to extract model parameters from commercial Ethernet devices. Transfer relation measurements of the commercial Ethernet transformers are compared against the model's behavior and it is found that the tuned, distributed models produce the best transfer relation match to the measured data. Process descriptions and testing results on fabricated thin-film dielectric-core toroid transformers are presented. The best results were found for a 32-turn transformer loaded with 100Ω, the impedance of twisted pair cable. This transformer gave a flat response from about 10MHz to 40MHz with a height of approximately 0.45. For the fabricated transformer structures, theoretical methods to determine resistance, capacitance and inductance are presented. A special analytical and numerical analysis of the fabricated transformer inductance is presented. Planar cuts of magnetic slope fields around the dielectric-core toroid are shown that describe the effect of core height and winding density on flux uniformity without a magnetic core.

Saturated internal instabilities in advanced-tokamak plasmas

NASA Astrophysics Data System (ADS)

Hua, M.-D.; Chapman, I. T.; Pinches, S. D.; Hastie, R. J.; MAST Team

2010-06-01

"Advanced tokamak" (AT) scenarios were developed with the aim of reaching steady-state operation in future potential tokamak fusion power plants. AT scenarios exhibit non-monotonic to flat safety factor profiles (q, a measure of the magnetic field line pitch), with the minimum q (qmin) slightly above an integer value (qs). However, it has been predicted that these q profiles are unstable to ideal magnetohydrodynamic instabilities as qmin approaches qs. These ideal instabilities, observed and diagnosed as such for the first time in MAST plasmas with AT-like q profiles, have far-reaching consequences like confinement degradation, flattening of the toroidal core rotation or enhanced fast ion losses. These observations motivate the stability analysis of advanced-tokamak plasmas, with a view to provide guidance for stability thresholds in AT scenarios. Additionally, the measured rotation damping is compared to the self-consistently calculated predictions from neoclassical toroidal viscosity theory.

Analysis of recurrent patterns in toroidal magnetic fields.

PubMed

Sanderson, Allen R; Chen, Guoning; Tricoche, Xavier; Pugmire, David; Kruger, Scott; Breslau, Joshua

2010-01-01

In the development of magnetic confinement fusion which will potentially be a future source for low cost power, physicists must be able to analyze the magnetic field that confines the burning plasma. While the magnetic field can be described as a vector field, traditional techniques for analyzing the field's topology cannot be used because of its Hamiltonian nature. In this paper we describe a technique developed as a collaboration between physicists and computer scientists that determines the topology of a toroidal magnetic field using fieldlines with near minimal lengths. More specifically, we analyze the Poincaré map of the sampled fieldlines in a Poincaré section including identifying critical points and other topological features of interest to physicists. The technique has been deployed into an interactive parallel visualization tool which physicists are using to gain new insight into simulations of magnetically confined burning plasmas.

Equilibrium evolution in oscillating-field current-drive experiments

NASA Astrophysics Data System (ADS)

McCollam, K. J.; Anderson, J. K.; Blair, A. P.; Craig, D.; Den Hartog, D. J.; Ebrahimi, F.; O'Connell, R.; Reusch, J. A.; Sarff, J. S.; Stephens, H. D.; Stone, D. R.; Brower, D. L.; Deng, B. H.; Ding, W. X.

2010-08-01

Oscillating-field current drive (OFCD) is a proposed method of steady-state toroidal plasma sustainment in which ac poloidal and toroidal loop voltages are applied to produce a dc plasma current. OFCD is added to standard, inductively sustained reversed-field pinch plasmas in the Madison Symmetric Torus [R. N. Dexter et al., Fusion Technol. 19, 131 (1991)]. Equilibrium profiles and fluctuations during a single cycle are measured and analyzed for different relative phases between the two OFCD voltages and for OFCD off. For OFCD phases leading to the most added plasma current, the measured energy confinement is slightly better than that for OFCD off. By contrast, the phase of the maximum OFCD helicity-injection rate also has the maximum decay rate, which is ascribed to transport losses during discrete magnetic-fluctuation events induced by OFCD. Resistive-magnetohydrodynamic simulations of the experiments reproduce the observed phase dependence of the added current.

Optical design of a dual wave band catadioptric endoscope for the Joint European Torus

NASA Astrophysics Data System (ADS)

Greco, Vincenzo; Maddaluno, Giorgio

2004-02-01

In this paper we describe the optical design of a catadioptric endoscope for the Joint European Torus (JET). The JET is the flagship experiment in the European nuclear fusion research programme. It is a large tokamak (Russian acronym for "toroidal magnetic chamber") system located at Culham (UK). At the centre of this machine there is a toroidal (ring - shaped) vacuum vessel where the plasma is confined by magnetic fields. The endoscope explores in two wave bands (4.2 μm - 4.4 μm and 0.6 μm - 0.7 μm) an entire cross section of the vacuum vessel. It then creates for each wave band an image onto a separate area image sensor, located 5500 mm away from the plasma behind a concrete shield. The endoscope performs two different functions namely: infrared thermography on plasma facing components and in vessel inspection.

Designs of LiMIT as a Limiter in the EAST Tokamak

NASA Astrophysics Data System (ADS)

Szott, Matthew; Christenson, Michael; Kalathiparambil, Kishor; Ruzic, David

2016-10-01

Liquid metal plasma facing components (PFCs) provide a constantly refreshing, self-healing surface that can reduce erosion and thermal stress damage to prolong device lifetime, and additionally decrease edge recycling, reduce impurities, and enhance plasma performance. The Liquid Metal Infused Trench (LiMIT) system, developed at UIUC, has demonstrated thermoelectric magnetohydrodynamic (TEMHD) driven flow of liquid lithium through series of solid trenches. This TEMHD effect drives liquid lithium in fusion systems using the plasma heat flux and the toroidal magnetic field, and the surface tension of the liquid lithium maintains a fresh surface on top of the solid trenches. LiMIT has been successfully tested at UIUC as well as HT-7 and Magnum PSI at heat fluxes up to 3 MW/m2. The next step is demonstrating system viability in full-scale fusion-relevant conditions. In collaboration with a team in Hefei, design and testing has begun for a large scale LiMIT system that will act as a limiter in EAST. The designs improve upon previous versions of LiMIT tested at Illinois and incorporate lessons learned from earlier tests of liquid metal PFCs at EAST. Existing infrastructure is used to load and supply lithium to the system, and the LiMIT trenches will help maintain a smooth, fresh surface as well as aid in propelling the lithium out of direct plasma flux to improve heat transfer. Supported by DOE/ALPS DE-FG02-99ER54515.

APPARATUS FOR TRAPPING ENERGETIC CHARGED PARTICLES AND CONFINING THE RESULTING PLASMA

DOEpatents

Gibson, G.; Jordan, W.C.; Lauer, E.J.

1963-04-01

The present invention relates to a plasma-confining device and a particle injector therefor, the device utilizing a generally toroidal configuration with magnetic fields specifically tailored to the associated injector. The device minimizes the effects of particle end losses and particle drift to the walls with a relatively simple configuration. More particularly, the magnetic field configuration is created by a continuous array of circular, mirror field coils, disposed side-by- side, in particularly spaced relation, to form an endless, toroidal loop. The resulting magnetic field created therein has the appearance of a bumpy'' torus, from which is derived the name Bumpy Torus.'' One of the aforementioned coils is split transverse to its axis, and injection of particles is accomplished along a plane between the halves of such modified coil. The guiding center of the particles follows a constant magnetic field in the plane for a particular distance within the torus, to move therefrom onto a precessional surface which does not intersect the point of injection. (AEC)

Thermomagnetic burn control for magnetic fusion reactor

DOEpatents

Rawls, J.M.; Peuron, A.U.

1980-07-01

Apparatus is provided for controlling the plasma energy production rate of a magnetic-confinement fusion reactor, by controlling the magnetic field ripple. The apparatus includes a group of shield sectors formed of ferromagnetic material which has a temperature-dependent saturation magnetization, with each shield lying between the plasma and a toroidal field coil. A mechanism for controlling the temperature of the magnetic shields, as by controlling the flow of cooling water therethrough, thereby controls the saturation magnetization of the shields and therefore the amount of ripple in the magnetic field that confines the plasma, to thereby control the amount of heat loss from the plasma. This heat loss in turn determines the plasma state and thus the rate of energy production.

Tokamak DEMO-FNS: Concept of magnet system and vacuum chamber

DOE Office of Scientific and Technical Information (OSTI.GOV)

Azizov, E. A., E-mail: Azizov-EA@nrcki.ru; Ananyev, S. S.; Belyakov, V. A.

The level of knowledge accumulated to date in the physics and technologies of controlled thermonuclear fusion (CTF) makes it possible to begin designing fusion—fission hybrid systems that would involve a fusion neutron source (FNS) and which would admit employment for the production of fissile materials and for the transmutation of spent nuclear fuel. Modern Russian strategies for CTF development plan the construction to 2023 of tokamak-based demonstration hybrid FNS for implementing steady-state plasma burning, testing hybrid blankets, and evolving nuclear technologies. Work on designing the DEMO-FNS facility is still in its infancy. The Efremov Institute began designing its magnet systemmore » and vacuum chamber, while the Kurchatov Institute developed plasma-physics design aspects and determined basic parameters of the facility. The major radius of the plasma in the DEMO-FNS facility is R = 2.75 m, while its minor radius is a = 1 m; the plasma elongation is k{sub 95} = 2. The fusion power is P{sub FUS} = 40 MW. The toroidal magnetic field on the plasma-filament axis is B{sub t0} = 5 T. The plasma current is I{sub p} = 5 MA. The application of superconductors in the magnet system permits drastically reducing the power consumed by its magnets but requires arranging a thick radiation shield between the plasma and magnet system. The central solenoid, toroidal-field coils, and poloidal-field coils are manufactured from, respectively, Nb{sub 3}Sn, NbTi and Nb{sub 3}Sn, and NbTi. The vacuum chamber is a double-wall vessel. The space between the walls manufactured from 316L austenitic steel is filled with an iron—water radiation shield (70% of stainless steel and 30% of water).« less

NIMROD: A computational laboratory for studying nonlinear fusion magnetohydrodynamics

NASA Astrophysics Data System (ADS)

Sovinec, C. R.; Gianakon, T. A.; Held, E. D.; Kruger, S. E.; Schnack, D. D.

2003-05-01

Nonlinear numerical studies of macroscopic modes in a variety of magnetic fusion experiments are made possible by the flexible high-order accurate spatial representation and semi-implicit time advance in the NIMROD simulation code [A. H. Glasser et al., Plasma Phys. Controlled Fusion 41, A747 (1999)]. Simulation of a resistive magnetohydrodynamics mode in a shaped toroidal tokamak equilibrium demonstrates computation with disparate time scales, simulations of discharge 87009 in the DIII-D tokamak [J. L. Luxon et al., Plasma Physics and Controlled Nuclear Fusion Research 1986 (International Atomic Energy Agency, Vienna, 1987), Vol. I, p. 159] confirm an analytic scaling for the temporal evolution of an ideal mode subject to plasma-β increasing beyond marginality, and a spherical torus simulation demonstrates nonlinear free-boundary capabilities. A comparison of numerical results on magnetic relaxation finds the n=1 mode and flux amplification in spheromaks to be very closely related to the m=1 dynamo modes and magnetic reversal in reversed-field pinch configurations. Advances in local and nonlocal closure relations developed for modeling kinetic effects in fluid simulation are also described.

Evaluation of fluid behavior in spinning toroidal tanks

NASA Technical Reports Server (NTRS)

Anderson, J. E.; Fester, D. A.; Dugan, D. W.

1976-01-01

An experimental study was conducted to evaluate propellant behavior in spinning toroidal tanks that could be used in a retro-propulsion system of an advanced outer-planet Pioneer orbiter. Information on propellant slosh and settling and on ullage orientation and stability was obtained. The effects of axial acceleration, spin rate, spin rate change, and spacecraft wobble, both singly and in combination, were evaluated using a 1/8-scale transparent tank in one-g and low-g environments. Liquid loadings ranged from 5% to 96% full. The impact of a surface tension acquisition device was assessed. Testing simulated the behavior of F2/N2H4 and N2O4/MMH propellants. Results are presented which indicate no major fluid behavior problems would be encountered with any of the four propellants in the toroidal tanks of a spin-stabilized orbiter spacecraft.

Modelling ion cyclotron emission from KSTAR tokamak and LHD helical device plasmas

NASA Astrophysics Data System (ADS)

Dendy, Richard; Chapman, Ben; Reman, Bernard; Chapman, Sandra; Akiyama, Tsuyoshi; Yun, Gunsu

2017-10-01

New high quality measurements of ion cyclotron emission (ICE) from KSTAR and LHD greatly extend the scope and diversity of plasma conditions under which ICE is observed. Variables include the origin (fusion reactions or neutral beam injection) and energy (sub- or super-Alfvénic) of the minority energetic ions that drive ICE; the composition of the bulk plasma (hydrogen or deuterium) which supports the modes excited; plasma density in the emitting region, and the timescale on which it changes; and toroidal magnetic field geometry (tokamak or helical device). Future exploitation of ICE as a diagnostic for energetic ion populations in JET D-T plasmas and in ITER rests on quantitative understanding of the physics of the emission. This is tested and extended by current KSTAR and LHD measurements of ICE. We report progress on direct numerical simulation using full orbit ion kinetic codes that solve the Maxwell-Lorentz equations for hundreds of millions of particles. In the saturated regime, these simulations yield excited field spectra that correspond directly to the measured ICE spectra under diverse KSTAR and LHD regimes. At early times, comparison of simulation outputs with linear analytical theory confirms the magnetoacoustic cyclotron instability as the basic driver of ICE. Supported by RCUK Energy Programme Grant EP/P012450/1, NRF Korea Grant 2014M1A7A1A03029881, NIFS budget ULHH029 and Euratom.

Production of plasmas by long-wavelength lasers

DOEpatents

Dawson, J.M.

1973-10-01

A long-wavelength laser system for heating low-density plasma to high temperatures is described. In one embodiment, means are provided for repeatedly receiving and transmitting long-wavelength laser light in successive stages to form a laser-light beam path that repeatedly intersects with the equilibrium axis of a magnetically confined toroidal plasma column for interacting the laser light with the plasma for providing controlled thermonuclear fusion. Embodiments for heating specific linear plasmas are also provided. (Official Gazette)

Modeling the excitation of global Alfven modes by an external antenna in the Joint European Torus (JET)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Huysmans, G.T.A.; Kerner, W.; Borba, D.

1995-05-01

The active excitation of global Alfven modes using the saddle coils in the Joint European Torus (JET) [{ital Plasma} {ital Physics} {ital and} {ital Controlled} {ital Nuclear} {ital Fusion} {ital Research} 1984, Proceedings of the 10th International Conference, London (International Atomic Energy Agency, Vienna, 1985), Vol. 1, p. 11] as the external antenna, will provide information on the damping of global modes without the need to drive the modes unstable. For the modeling of the Alfven mode excitation, the toroidal resistive magnetohydrodynamics (MHD) code CASTOR (Complex Alfven Spectrum in TORoidal geometry) [18{ital th} {ital EPS} {ital Conference} {ital On} {italmore » Controlled} {ital Fusion} {ital and} {ital Plasma} {ital Physics}, Berlin, 1991, edited by P. Bachmann and D. C. Robinson (The European Physical Society, Petit-Lancy, 1991), Vol. 15, Part IV, p. 89] has been extended to calculate the response to an external antenna. The excitation of a high-performance, high beta JET discharge is studied numerically. In particular, the influence of a finite pressure is investigated. Weakly damped low-{ital n} global modes do exist in the gaps in the continuous spectrum at high beta. A pressure-driven global mode is found due to the interaction of Alfven and slow modes. Its frequency scales solely with the plasma temperature, not like a pure Alfven mode with a density and magnetic field.« less

Overview of the HIT-SI3 spheromak experiment

NASA Astrophysics Data System (ADS)

Hossack, A. C.; Jarboe, T. R.; Chandra, R. N.; Morgan, K. D.; Sutherland, D. A.; Everson, C. J.; Penna, J. M.; Nelson, B. A.

2017-10-01

The HIT-SI and HIT-SI3 spheromak experiments (a = 23 cm) study efficient, steady-state current drive for magnetic confinement plasmas using a novel method which is ideal for low aspect ratio, toroidal geometries. Sustained spheromaks show coherent, imposed plasma motion and low plasma-generated mode activity, indicating stability. Analysis of surface magnetic fields in HIT-SI indicates large n = 0 and 1 mode amplitudes and little energy in higher modes. Within measurement uncertainties all the n = 1 energy is imposed by the injectors, rather than being plasma-generated. The fluctuating field imposed by the injectors is sufficient to sustain the toroidal current through dynamo action whereas the plasma-generated field is not (Hossack et al., Phys. Plasmas, 2017). Ion Doppler spectroscopy shows coherent, imposed plasma motion inside r 10 cm in HIT-SI and a smaller volume of coherent motion in HIT-SI3. Coherent motion indicates the spheromak is stable and a lack of plasma-generated n = 1 energy indicates the maximum q is maintained below 1 for stability during sustainment. In HIT-SI3, the imposed mode structure is varied to test the plasma response (Hossack et al., Nucl. Fusion, 2017). Imposing n = 2, n = 3, or large, rotating n = 1 perturbations is correlated with transient plasma-generated activity. Work supported by the U.S. Department of Energy, Office of Science, Office of Fusion Energy Sciences, under Award Number DE-FG02-96ER54361.

First AC loss test and analysis of a Bi2212 cable-in-conduit conductor for fusion application

NASA Astrophysics Data System (ADS)

Qin, Jinggang; Shi, Yi; Wu, Yu; Li, Jiangang; Wang, Qiuliang; He, Yuxiang; Dai, Chao; Liu, Fang; Liu, Huajun; Mao, Zhehua; Nijhuis, Arend; Zhou, Chao; Devred, Arnaud

2018-01-01

The main goal of the Chinese fusion engineering test reactor (CFETR) is to build a fusion engineering tokamak reactor with a fusion power of 50-200 MW, and plan to test the breeding tritium during the fusion reaction. This may require a maximum magnetic field of the central solenoid and toroidal field coils up to 15 T. New magnet technologies should be developed for the next generation of fusion reactors with higher requirements. Bi2Sr2CaCu2Ox (Bi2212) is considered as a potential and promising superconductor for the magnets in the CFETR. R&D activities are ongoing at the Institute of Plasma Physics, Chinese Academy of Sciences for demonstration of the feasibility of a CICC based on Bi2212 round wire. One sub-size conductor cabled with 42 wires was designed, manufactured and tested with limited strand indentation during cabling and good transport performance. In this paper, the first test results and analysis on the AC loss of Bi2212 round wires and cabled conductor samples are presented. Furthermore, the impact of mechanical load on the AC loss of the sub-size conductor is investigated to represent the operation conditions with electromagnetic loads. The first tests provide an essential basis for the validation of Bi2212 CICC and its application in fusion magnets.

Temperature Effects on the Magnetic Properties of Silicon-Steel Sheets Using Standardized Toroidal Frame

PubMed Central

Wu, Cheng-Ju; Lin, Shih-Yu; Chou, Shang-Chin; Tsai, Chia-Yun; Yen, Jia-Yush

2014-01-01

This study designed a detachable and standardized toroidal test frame to measure the electromagnetic characteristic of toroidal laminated silicon steel specimens. The purpose of the design was to provide the measurements with standardized and controlled environment. The device also can withstand high temperatures (25–300°C) for short time period to allow high temperature tests. The accompanying driving circuit facilitates testing for high frequency (50–5,000 Hz) and high magnetic flux (0.2–1.8 T) conditions and produces both sinusoidal and nonsinusoidal test waveforms. The thickness of the stacked laminated silicon-steel sheets must be 30~31 mm, with an internal diameter of 72 mm and an outer diameter of 90 mm. With the standardized setup, it is possible to carry out tests for toroidal specimen in high temperature and high flux operation. The test results show that there is a tendency of increased iron loss under high temperature operation. The test results with various driving waveforms also provide references to the required consideration in engineering designs. PMID:25525629

Temperature effects on the magnetic properties of silicon-steel sheets using standardized toroidal frame.

PubMed

Wu, Cheng-Ju; Lin, Shih-Yu; Chou, Shang-Chin; Tsai, Chia-Yun; Yen, Jia-Yush

2014-01-01

This study designed a detachable and standardized toroidal test frame to measure the electromagnetic characteristic of toroidal laminated silicon steel specimens. The purpose of the design was to provide the measurements with standardized and controlled environment. The device also can withstand high temperatures (25-300°C) for short time period to allow high temperature tests. The accompanying driving circuit facilitates testing for high frequency (50-5,000 Hz) and high magnetic flux (0.2-1.8 T) conditions and produces both sinusoidal and nonsinusoidal test waveforms. The thickness of the stacked laminated silicon-steel sheets must be 30~31 mm, with an internal diameter of 72 mm and an outer diameter of 90 mm. With the standardized setup, it is possible to carry out tests for toroidal specimen in high temperature and high flux operation. The test results show that there is a tendency of increased iron loss under high temperature operation. The test results with various driving waveforms also provide references to the required consideration in engineering designs.

Thermal Ablation by High-Intensity-Focused Ultrasound Using a Toroid Transducer Increases the Coagulated Volume and Allows Coagulation Near Portal and Hepatic veins in Pigs

DOE Office of Scientific and Technical Information (OSTI.GOV)

Melodelima, D.; N'Djin, W. A.; Parmentier, H.

2009-04-14

A new geometry of HIFU transducer is described to enlarge the coagulated volume. The geometry of the transducer was not spherical. The surface of the transducer was built based on a toroid geometry. The transducer was generated by the revolution of a circle about an axis lying in its plane. Eight emitters operating at a frequency of 3 MHz were diced out of a single toroid piezocomposite element. Each of the eight emitters was divided into 32 transducers. The focal zone is conical and located at 70 mm from the transducer. A 7.5 MHz ultrasound imaging probe is placed inmore » the centre of the device for guiding the treatment. Our long-term objective is to develop a device that can be used during surgery. In vivo trials have been performed on 13 pigs to demonstrate this new principle and to evaluate the vascular tolerance of the treatment. This new geometry combined with consecutive activation of the eight emitters around the toroid allows achieving a mean thermal ablation of 7.0{+-}2.5 cm3 in 40 seconds. All lesions were visible with high contrast on sonograms. The correlation between the size of lesions observed on sonograms and during gross examination was 92%. This allows the user to easily enlarge the coagulated volume by juxtaposing single lesions. The pigs tolerate the treatment well over the experimental period even when coagulation was produced through portal and/or hepatic veins.« less

Magnetism of toroidal field in two-fluid equilibrium of CHI driven spherical torus

NASA Astrophysics Data System (ADS)

Kanki, T.; Nagata, M.

2016-10-01

Double-pulsing CHI (D-CHI) experiment has been conducted in the HIST device to achieve a quasi-steady sustainment and good confinement of spherical torus (ST) plasmas. The feature of CHI driven ST such as diamagnetic toroidal field in the central open flux column (OFC) region and strong poloidal flow shear around the separatrix in the high field side suggests the two-fluid effect. The relationship between the magnetism of the toroidal field and the poloidal flow velocity is investigated by modelling the D-CHI (mainly driving the poloidal electron flow along the open flux) in the two-fluid equilibrium calculations. The poloidal component of Ampere's law leads that the toroidal field is related to the difference between the stream functions of ion ψi and electron ψe for the poloidal flow, indicating that the toroidal field with ψe >ψi results in a diamagnetic profile, while that with ψe <ψi results in a paramagnetic one. The gradient of the stream function determines the polarity and the strength of the poloidal flow velocity. It is found that the two-fluid equilibrium of CHI driven ST satisfies ψe > 0 and ψi < 0 in the OFC region, and ψe < 0 and ψi < 0 in the closed flux region. The toroidal field is a diamagnetic profile in the OFC region due to ψe >ψi and |uez | > |uiz | , where uez and uiz denote the poloidal electron and ion flow velocities, respectively. It becomes from a diamagnetic to a paramagnetic profile in the closed flux region, because ψe (uez) approaches ψi (uiz) around the magnetic axis. The poloidal ion flow shear is enhanced in the OFC region due to the ion inertial effect through the toroidal ion flow velocity.

Final Report: Levitated Dipole Experiment

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kesner, Jay; Mauel, Michael

2013-03-10

Since the very first experiments with the LDX, research progress was rapid and significant. Initial experiments were conducted with the high-field superconducting coil suspended by three thin rods. These experiments produced long-pulse, quasi-steady-state microwave discharges, lasting more than 10 s, having peak beta values of 20% [Garnier, Phys. Plasmas, v13, p. 056111, 2006]. High-beta, near steady-state discharges have been maintained in LDX for more than 20 seconds, and this capability makes LDX the longest pulse fusion confinement experiment now operating in the U.S. fusion program. In both supported and levitated configurations, detailed measurements are made of discharge evolution, plasma dynamicsmore » and instability, and the roles of gas fueling, microwave power deposition profiles, and plasma boundary shape. High-temperature plasma is created by multifrequency electron cyclotron resonance heating allowing control of heating profiles. Depending upon neutral fueling rates, the LDX discharges contain a fraction of energetic electrons, with mean energies above 50 keV. Depending on whether or not the superconducting dipole is levitated or supported, the peak thermal electron temperature is estimated to exceed 500 eV and peak densities reach 1.0E18 (1/m3). Several significant discoveries resulted from the routine investigation of plasma confinement with a magnetically-levitated dipole. For the first time, toroidal plasma with pressure approaching the pressure of the confining magnetic field was well-confined in steady-state without a toroidal magnetic field. Magnetic levitation proved to be reliable and is now routine. The dipole's cryostat allows up to three hours of "float time" between re-cooling with liquid helium and providing scientists unprecedented access to the physics of magnetizd plasma. Levitation eliminates field-aligned particle sources and sinks and results in a toroidal, magnetically-confined plasma where profiles are determined by cross-field transport. We find levitation causes the central plasma density to increase dramatically and to significantly improve the confinement of thermal plasma [Boxer, Nature-Physics, v8, p. 949, 2010]. Several diagnostic systems have been used to measure plasma fluctuations, and these appear to represent low-frequency convection that may lead to adiabatic heating and strongly peaked pressure profiles. These experiments are remarkable, and the motivate wide-ranging studies of plasma found in space and confined for fusion energy. In the following report, we describe: (i) observations of the centrally-peaked density profile that appears naturally as a consequence of a strong turbulent pinch, (ii) observations of overall density and pressure increases that suggest large improvements to the thermal electron confinement time result occur during levitation, and (iii) the remarkable properties of low-frequency plasma fluctuations that cause magnetized plasma to "self-organize" into well-confined, centrally-peaked profiles that are relative to fusion and to space.« less

Observation of Alpha-Driven TAEs in TFTR

NASA Astrophysics Data System (ADS)

Nazikian, R.; Chang, Z.; Fu, G. Y.; Majeski, R.; Batha, S.; Bell, M.; Budny, R.; Cheng, C. Z.; Darrow, D. S.; Duong, H.; Efthimion, P. C.; Fredrickson, E.; Levinton, F.; Mazzucato, E.; Medley, S.; Taylor, S.; Zweben, G.

1996-11-01

Transient mode activity in the TAE range of frequencies (150-170 kHz) with toroidal mode numbers n=2,3 is observed in reduced magnetic shear DT discharges on TFTR with a fusion power threshold of ~1.5 MW. Mode activity appears 50-100 msec after NBI in discharges with the following machine parameter: B=5.3 T, I=1.6MA, R=260cm, P_NBI=25-28 MW, q(0)>2.0 from MSE and toroidal beta β_T<1%. The elevated q(0) and reduced central shear |s|<0.2 is achieved using a full size plasma startup with delayed NBI. Theoretical calculations using NOVA-K indicate that the combined effect of low shear, low beta and elevated q(0) leads to a very low instability threshold for the alpha-driven TAE with β_α (0) ~ 10-4. This appears to be consistent with experimental observations of mode activity in DT plasmas with β_α ~ 10-4 (determined from TRANSP analysis). Thus far the modes have only been observed on Mirnov coils with fluctuation levels tildeB ~ 1mG. Efforts to determine mode location by perturbing the edge density and inducing strong toroidal velocity shear will be reported, as will efforts to affect mode stability by systematically varying the central safety factor.

Spheromak plasma flow injection into a torus chamber and the HIST plasmas

NASA Astrophysics Data System (ADS)

Hatuzaki, Akinori

2005-10-01

The importance of plasma flow or two-fluid effect is recognized in understanding the relaxed states of high-beta torus plasmas, start-up and current drive by non-coaxial helicity injection, magnetic reconnection and plasma dynamo in fusion, laboratory and space plasmas. As a new approach to create a flowing two-fluid plasma equilibrium, we have tried to inject tangentially the plasma flow with spheromak-type magnetic configurations into a torus vacuum chamber with an external toroidal magnetic field (TF) coil. In the initial experiments, the RFP-like configuration with helical magnetic structures was realized in the torus vessel. The ion flow measurement with Mach probes showed that the ion flow keeps the same direction despite the reversal of the toroidal current and the axial electric field. The ion fluid comes to flow in the opposite direction to the electron fluid by the reversal of TF. This result suggests that not only electron but also ion flow contributes significantly on the reversed toroidal current. In this case, the ratio of ui to the electron flow velocity ue is estimated as ui/ue ˜ 1/2. We also will inject the spheromak flow into the HIST spherical torus plasmas to examine the possibilities to embedding the two-fluid effect in the ST plasmas.

Resonance and Chaotic Trajectories in Magnetic Field Reversed Configuration

DOE Office of Scientific and Technical Information (OSTI.GOV)

A.S. Landsman; S.A. Cohen; M. Edelman

The nonlinear dynamics of a single ion in a field-reversed configuration (FRC) were investigated. FRC is a toroidal fusion device which uses a specific type of magnetic field to confine ions. As a result of angular invariance, the full three-dimensional Hamiltonian system can be expressed as two coupled, highly nonlinear oscillators. Due to the high nonlinearity in the equations of motion, the behavior of the system is extremely complex, showing different regimes, depending on the values of the conserved canonical angular momentum and the geometry of the fusion vessel. Perturbation theory and averaging were used to derive the unperturbed Hamiltonianmore » and frequencies of the two degrees of freedom. The derived equations were then used to find resonances and compare to Poincar{copyright} surface-of-section plots. A regime was found where the nonlinear resonances were clearly separated by KAM [Kolmogorov-Arnold-Mosher] curves. The structure of the observed island chains was explained. The condition for the destruction of KAM curves and the onset of strong chaos was derived, using Chirikov island overlap criterion, and shown qualitatively to depend both on the canonical angular momentum and geometry of the device. After a brief discussion of the adiabatic regime the paper goes on to explore the degenerate regime that sets in at higher values of angular momenta. In this regime, the unperturbed Hamiltonian can be approximated as two uncoupled linear oscillators. In this case, the system is near-integrable, except in cases of a universal resonance, which results in large island structures, due to the smallness of nonlinear terms, which bound the resonance. The linear force constants, dominant in this regime, were derived and the geometry for a large one-to-one resonance identified. The above analysis showed good agreement with numerical simulations and was able to explain characteristic features of the dynamics.« less

The Asfora Bullet Cage System Shows Comparable Fusion Rate Success Versus Control Cage in Posterior Lumbar Interbody Fusion in a Randomized Clinical Trial.

PubMed

Morgan, Jeremy P; Miller, Ashley L; Thompson, Paul A; Asfora, Wilson T

2016-04-01

Low back pain and degeneration of the intervertebral disc are an integrated malady that affects millions of Americans. Cage devices used in association with posterior lumbar interbody fusion (PLIF) have been shown to be an effective approach in the treatment of a number of lower spine disorders attributed to degenerative disc disease (DDD). This study was undertaken as part of a U.S. Food and Drug Administration (FDA) Investigational Device Exemption (IDE) study and compares the effectiveness of the Asfora Bullet Cage System (ABCS) to successfully fuse vertebra at one or two levels between L2 and S1 in patients with DDD to an FDA approved comparison device, the Medtronic-Sofamor Danek Inter Fix Threaded Fusion Device (MSDIFD). A total of 257 randomized participants were implanted with either the ABCS device (n = 132) or the MSDIFD device (n = 125) through an open posterior approach using autogenous local bone graft without the use of pedicle screws. Patients were evaluated prior to surgery and at the 24 month (24-M) visit for fusion status, deep tendon reflex status, sensory function, motor function, straight leg raise status, pain, disability, and device safety. Radiological evaluation and statistical analysis were performed by independent professionals. Evaluation of device success was performed at 24-M visit. From the original group of 257 patients, 59 were lost to follow-up. Primary measures of success at the 24-M visit involved pain and function, fusion, neurological status, and device-related adverse events measures. Pain and function improved in both (MSDIFD: 75.7 percent; ABCS: 82.6 percent). Fusion success with all radiographic points at 24-M visits was 79.4 percent MSDIFD and 88.2 percent ABCS. Neurological improvement was seen in both (MSDIFD: 77.0 percent; ABCS: 87.8 percent). One device-related grade 1 adverse event was reported in the MSDIFD group. Disc height preservation was equivalent for single level fusions (MSDIFD: 16.1 percent; ABCS: 20.0 percent) and second level fusions (MSDIFD: 10.7 percent; ABCS: 14.3 percent). General health and well-being improvement was the same (MSDIFD: 37.0 percent; ABCS: 40.0 percent). Subsequent fusion, up to 10 years, was equivalent (MSDIFD: 83.8 percent; ABCS: 91.2). Results for both devices were considered to be satisfactory, with a slight non-significant superiority for the ABCS. From the ABCS device FDA IDE sanctioned study and the review of the literature, we concluded that the Asfora Bullet Cage System is safe, effective and comparable to other interbody fusion devices which are used stand-alone or in conjunction with pedicle screws, rhBMP-2, or autogenous bone harvested from the iliac crest inserted through anterior, lateral or posterior approaches.

NIMROD simulations and physics assessment of possible designs for a next generation Steady Inductive Helicity Injection HIT device

NASA Astrophysics Data System (ADS)

Penna, James; Morgan, Kyle; Grubb, Isaac; Jarboe, Thomas

2017-10-01

The Helicity Injected Torus - Steady Inductive 3 (HIT-SI3) experiment forms and maintains spheromaks via Steady Inductive Helicity Injection (SIHI) using discrete injectors that inject magnetic helicity via a non-axisymmetric perturbation and drive toroidally symmetric current. Newer designs for larger SIHI-driven spheromaks incorporate a set of injectors connected to a single external manifold to allow more freedom for the toroidal structure of the applied perturbation. Simulations have been carried out using the NIMROD code to assess the effectiveness of various imposed mode structures and injector schema in driving current via Imposed Dynamo Current Drive (IDCD). The results are presented here for varying flux conserver shapes on a device approximately 1.5 times larger than the current HIT-SI3 experiment. The imposed mode structures and spectra of simulated spheromaks are analyzed in order to examine magnetic structure and stability and determine an optimal regime for IDCD sustainment in a large device. The development of scaling laws for manifold operation is also presented, and simulation results are analyzed and assessed as part of the development path for the large scale device.

From Data Acquisition to Data Fusion: A Comprehensive Review and a Roadmap for the Identification of Activities of Daily Living Using Mobile Devices

PubMed Central

Pires, Ivan Miguel; Garcia, Nuno M.; Pombo, Nuno; Flórez-Revuelta, Francisco

2016-01-01

This paper focuses on the research on the state of the art for sensor fusion techniques, applied to the sensors embedded in mobile devices, as a means to help identify the mobile device user’s daily activities. Sensor data fusion techniques are used to consolidate the data collected from several sensors, increasing the reliability of the algorithms for the identification of the different activities. However, mobile devices have several constraints, e.g., low memory, low battery life and low processing power, and some data fusion techniques are not suited to this scenario. The main purpose of this paper is to present an overview of the state of the art to identify examples of sensor data fusion techniques that can be applied to the sensors available in mobile devices aiming to identify activities of daily living (ADLs). PMID:26848664

21 CFR 888.3080 - Intervertebral body fusion device.

Code of Federal Regulations, 2012 CFR

2012-04-01

... devices that contain bone grafting material. The special control is the FDA guidance document entitled... devices that include any therapeutic biologic (e.g., bone morphogenic protein). Intervertebral body fusion...

21 CFR 888.3080 - Intervertebral body fusion device.

Code of Federal Regulations, 2010 CFR

2010-04-01

... devices that contain bone grafting material. The special control is the FDA guidance document entitled... devices that include any therapeutic biologic (e.g., bone morphogenic protein). Intervertebral body fusion...

21 CFR 888.3080 - Intervertebral body fusion device.

Code of Federal Regulations, 2013 CFR

2013-04-01

... devices that contain bone grafting material. The special control is the FDA guidance document entitled... devices that include any therapeutic biologic (e.g., bone morphogenic protein). Intervertebral body fusion...

21 CFR 888.3080 - Intervertebral body fusion device.

Code of Federal Regulations, 2014 CFR

2014-04-01

... devices that contain bone grafting material. The special control is the FDA guidance document entitled... devices that include any therapeutic biologic (e.g., bone morphogenic protein). Intervertebral body fusion...

DOE Office of Scientific and Technical Information (OSTI.GOV)

Goumiri, I. R.; Rowley, C. W.; Sabbagh, S. A.

In this study, a model-based feedback system is presented enabling the simultaneous control of the stored energy through β n and the toroidal rotation profile of the plasma in National Spherical Torus eXperiment Upgrade device. Actuation is obtained using the momentum from six injected neutral beams and the neoclassical toroidal viscosity generated by applying three-dimensional magnetic fields. Based on a model of the momentum diffusion and torque balance, a feedback controller is designed and tested in closed-loop simulations using TRANSP, a time dependent transport analysis code, in predictive mode. Promising results for the ongoing experimental implementation of controllers are obtained.

Magnetless magnetic fusion

NASA Astrophysics Data System (ADS)

Beklemishev, A. D.; Tajima, T.

1994-02-01

The authors propose a concept of thermonuclear fusion reactor in which the plasma pressure is balanced by direct gas-wall interaction in a high-pressure vessel. The energy confinement is achieved by means of the self-contained toroidal magnetic configuration sustained by an external current drive or charged fusion products. This field structure causes the plasma pressure to decrease toward the inside of the discharge and thus it should be magnetohydrodynamically stable. The maximum size, temperature and density profiles of the reactor are estimated. An important feature of confinement physics is the thin layer of cold gas at the wall and the adjacent transitional region of dense arc-like plasma. The burning condition is determined by the balance between these nonmagnetized layers and the current-carrying plasma. They suggest several questions for future investigation, such as the thermal stability of the transition layer and the possibility of an effective heating and current drive behind the dense edge plasma. The main advantage of this scheme is the absence of strong external magnets and, consequently, potentially cheaper design and lower energy consumption.

Thermomagnetic burn control for magnetic fusion reactor

DOEpatents

Rawls, John M.; Peuron, Unto A.

1982-01-01

Apparatus is provided for controlling the plasma energy production rate of a magnetic-confinement fusion reactor, by controlling the magnetic field ripple. The apparatus includes a group of shield sectors (30a, 30b, etc.) formed of ferromagnetic material which has a temperature-dependent saturation magnetization, with each shield lying between the plasma (12) and a toroidal field coil (18). A mechanism (60) for controlling the temperature of the magnetic shields, as by controlling the flow of cooling water therethrough, thereby controls the saturation magnetization of the shields and therefore the amount of ripple in the magnetic field that confines the plasma, to thereby control the amount of heat loss from the plasma. This heat loss in turn determines the plasma state and thus the rate of energy production.

Model for the loop voltage of reversed field pinches

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jarboe, T.R.; Alper, B.

1987-04-01

A simple model is presented that uses the concept of helicity balance to predict the toroidal loop voltage of reversed field pinches (RFP's). Data from the RFP's at Culham (Plasma Phys. Controlled Fusion 27, 1307 (1985)) are used to calibrate and verify the model. The model indicates that most of the helicity dissipation occurs in edge regions that are outside the limiters or in regions where field lines contact the walls. The value of this new interpretation to future RFP and spheromak experiments is discussed.

APPARATUS FOR PRODUCING AND MANIPULATING PLASMAS

DOEpatents

Colgate, S.A.; Ferguson, J.P.; Furth, H.P.; Wright, R.E.

1960-07-26

An electrical pinch discharge apparatus is described for producing and manipulating high-temperature plasmas. The apparatus may be of either the linear or toroidal pinch discharge type. Arrangements are provided whereby stabilizing fields may be trapped in the plasma external to the main pinch discharge path and the boundary condition of the stabilizing field programed so as to stabilize the discharge or to promote instabilities in the discharge as desired. The produced plasmas may be employed for various purposes, and fusion neutrons have been produced with the apparatus.

Control of impurities in toroidal plasma devices

DOEpatents

Ohkawa, Tihiro

1980-01-01

A method and apparatus for plasma impurity control in closed flux plasma systems such as Tokamak reactors is disclosed. Local axisymmetrical injection of hydrogen gas is employed to reverse the normally inward flow of impurities into the plasma.

Inertial-Electrostatic Confinement (IEC) Fusion for Space Propulsion

NASA Technical Reports Server (NTRS)

Nadler, Jon

1999-01-01

An Inertial-Electrostatic Confinement (IEC) device was assembled at the Marshall Space Flight Center (MSFC) Propulsion Research Center (PRC) to study the possibility of using EEC technology for deep space propulsion and power. Inertial-Electrostatic Confinement is capable of containing a nuclear fusion plasma in a series of virtual potential wells. These wells would substantially increase plasma confinement, possibly leading towards a high-gain, breakthrough fusion device. A one-foot in diameter IEC vessel was borrowed from the Fusion Studies Laboratory at the University of Illinois@Urbana-Champaign for the summer. This device was used in initial parameterization studies in order to design a larger, actively cooled device for permanent use at the PRC.

Inertial-Electrostatic Confinement (IEC) Fusion For Space Propulsion

NASA Technical Reports Server (NTRS)

Nadler, Jon

1999-01-01

An Inertial-Electrostatic Confinement (IEC) device was assembled at the Marshall Space Flight Center (MSFC) Propulsion Research Center (PRC) to study the possibility of using IEC technology for deep space propulsion and power. Inertial-Electrostatic Confinement is capable of containing a nuclear fusion plasma in a series of virtual potential wells. These wells would substantially increase plasma confinement, possibly leading towards a high-gain, breakthrough fusion device. A one-foot in diameter IEC vessel was borrowed from the Fusion Studies Laboratory at the University of Illinois @ Urbana-Champaign for the summer. This device was used in initial parameterization studies in order to design a larger, actively cooled device for permanent use at the PRC.

AxiaLIF system: minimally invasive device for presacral lumbar interbody spinal fusion

PubMed Central

Rapp, Steven M; Miller, Larry E; Block, Jon E

2011-01-01

Lumbar fusion is commonly performed to alleviate chronic low back and leg pain secondary to disc degeneration, spondylolisthesis with or without concomitant lumbar spinal stenosis, or chronic lumbar instability. However, the risk of iatrogenic injury during traditional anterior, posterior, and transforaminal open fusion surgery is significant. The axial lumbar interbody fusion (AxiaLIF) system is a minimally invasive fusion device that accesses the lumbar (L4–S1) intervertebral disc spaces via a reproducible presacral approach that avoids critical neurovascular and musculoligamentous structures. Since the AxiaLIF system received marketing clearance from the US Food and Drug Administration in 2004, clinical studies of this device have reported high fusion rates without implant subsidence, significant improvements in pain and function, and low complication rates. This paper describes the design and approach of this lumbar fusion system, details the indications for use, and summarizes the clinical experience with the AxiaLIF system to date. PMID:22915939

AxiaLIF system: minimally invasive device for presacral lumbar interbody spinal fusion.

PubMed

Rapp, Steven M; Miller, Larry E; Block, Jon E

2011-01-01

Lumbar fusion is commonly performed to alleviate chronic low back and leg pain secondary to disc degeneration, spondylolisthesis with or without concomitant lumbar spinal stenosis, or chronic lumbar instability. However, the risk of iatrogenic injury during traditional anterior, posterior, and transforaminal open fusion surgery is significant. The axial lumbar interbody fusion (AxiaLIF) system is a minimally invasive fusion device that accesses the lumbar (L4-S1) intervertebral disc spaces via a reproducible presacral approach that avoids critical neurovascular and musculoligamentous structures. Since the AxiaLIF system received marketing clearance from the US Food and Drug Administration in 2004, clinical studies of this device have reported high fusion rates without implant subsidence, significant improvements in pain and function, and low complication rates. This paper describes the design and approach of this lumbar fusion system, details the indications for use, and summarizes the clinical experience with the AxiaLIF system to date.

Non-inductively driven tokamak plasmas at near-unity βt in the Pegasus toroidal experiment

NASA Astrophysics Data System (ADS)

Reusch, J. A.; Bodner, G. M.; Bongard, M. W.; Burke, M. G.; Fonck, R. J.; Pachicano, J. L.; Perry, J. M.; Pierren, C.; Rhodes, A. T.; Richner, N. J.; Rodriguez Sanchez, C.; Schlossberg, D. J.; Weberski, J. D.

2018-05-01

A major goal of the spherical tokamak (ST) research program is accessing a state of low internal inductance ℓi, high elongation κ, and high toroidal and normalized beta ( βt and βN) without solenoidal current drive. Local helicity injection (LHI) in the Pegasus ST [Garstka et al., Nucl. Fusion 46, S603 (2006)] provides non-solenoidally driven plasmas that exhibit these characteristics. LHI utilizes compact, edge-localized current sources for plasma startup and sustainment. It results in hollow current density profiles with low ℓi. The low aspect ratio ( R0/a ˜1.2 ) of Pegasus allows access to high κ and high normalized plasma currents ( IN=Ip/a BT>14 ). Magnetic reconnection during LHI provides auxiliary ion heating. Together, these features provide access to very high βt plasmas. Equilibrium analyses indicate that βt up to ˜100% is achieved. These high βt discharges disrupt at the ideal no-wall β limit at βN˜7.

A Comparison of Coarse-Grained and Continuum Models for Membrane Bending in Lipid Bilayer Fusion Pores

PubMed Central

Yoo, Jejoong; Jackson, Meyer B.; Cui, Qiang

2013-01-01

To establish the validity of continuum mechanics models quantitatively for the analysis of membrane remodeling processes, we compare the shape and energies of the membrane fusion pore predicted by coarse-grained (MARTINI) and continuum mechanics models. The results at these distinct levels of resolution give surprisingly consistent descriptions for the shape of the fusion pore, and the deviation between the continuum and coarse-grained models becomes notable only when the radius of curvature approaches the thickness of a monolayer. Although slow relaxation beyond microseconds is observed in different perturbative simulations, the key structural features (e.g., dimension and shape of the fusion pore near the pore center) are consistent among independent simulations. These observations provide solid support for the use of coarse-grained and continuum models in the analysis of membrane remodeling. The combined coarse-grained and continuum analysis confirms the recent prediction of continuum models that the fusion pore is a metastable structure and that its optimal shape is neither toroidal nor catenoidal. Moreover, our results help reveal a new, to our knowledge, bowing feature in which the bilayers close to the pore axis separate more from one another than those at greater distances from the pore axis; bowing helps reduce the curvature and therefore stabilizes the fusion pore structure. The spread of the bilayer deformations over distances of hundreds of nanometers and the substantial reduction in energy of fusion pore formation provided by this spread indicate that membrane fusion can be enhanced by allowing a larger area of membrane to participate and be deformed. PMID:23442963

Formation of high-β plasma and stable confinement of toroidal electron plasma in Ring Trap 1a)

NASA Astrophysics Data System (ADS)

Saitoh, H.; Yoshida, Z.; Morikawa, J.; Furukawa, M.; Yano, Y.; Kawai, Y.; Kobayashi, M.; Vogel, G.; Mikami, H.

2011-05-01

Formation of high-β electron cyclotron resonance heating plasma and stable confinement of pure electron plasma have been realized in the Ring Trap 1 device, a magnetospheric configuration generated by a levitated dipole field magnet. The effects of coil levitation resulted in drastic improvements of the confinement properties, and the maximum local β value has exceeded 70%. Hot electrons are major component of electron populations, and its particle confinement time is 0.5 s. Plasma has a peaked density profile in strong field region [H. Saitoh et al., 23rd IAEA Fusion Energy Conference EXC/9-4Rb (2010)]. In pure electron plasma experiment, inward particle diffusion is realized, and electrons are stably trapped for more than 300 s. When the plasma is in turbulent state during beam injection, plasma flow has a shear, which activates the diocotron (Kelvin-Helmholtz) instability. The canonical angular momentum of the particle is not conserved in this phase, realizing the radial diffusion of charged particles across closed magnetic surfaces. [Z. Yoshida et al., Phys Rev. Lett. 104, 235004 (2010); H. Saitoh et al., Phys. Plasmas 17, 112111 (2010).].

Measurements of tungsten migration in the DIII-D divertor

NASA Astrophysics Data System (ADS)

Wampler, W. R.; Rudakov, D. L.; Watkins, J. G.; McLean, A. G.; Unterberg, E. A.; Stangeby, P. C.

2017-12-01

An experimental study of migration of tungsten in the DIII-D divertor is described, in which the outer strike point of L-mode plasmas was positioned on a toroidal ring of tungsten-coated metal inserts. Net deposition of tungsten on the divertor just outside the strike point was measured on graphite samples exposed to various plasma durations using the divertor materials evaluation system. Tungsten coverage, measured by Rutherford backscattering spectroscopy (RBS), was found to be low and nearly independent of both radius and exposure time closer to the strike point, whereas farther from the strike point the W coverage was much larger and increased with exposure time. Depth profiles from RBS show this was due to accumulation of thicker mixed-material deposits farther from the strike point where the plasma temperature is lower. These results are consistent with a low near-surface steady-state coverage on graphite undergoing net erosion, and continuing accumulation in regions of net deposition. This experiment provides data needed to validate, and further improve computational simulations of erosion and deposition of material on plasma-facing components and transport of impurities in magnetic fusion devices. Such simulations are underway and will be reported later.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kaita, Robert; Boyle, Dennis; Gray, Timothy

Liquid metal walls have been proposed to address the first wall challenge for fusion reactors. The Lithium Tokamak Experiment (LTX) at the Princeton Plasma Physics Laboratory (PPPL) is the first magnetic confinement device to have liquid metal plasma-facing components (PFC's) that encloses virtually the entire plasma. In the Current Drive Experiment-Upgrade (CDX-U), a predecessor to LTX at PPPL, the highest improvement in energy confinement ever observed in Ohmically-heated tokamak plasmas was achieved with a toroidal liquid lithium limiter. The LTX extends this liquid lithium PFC by using a conducting conformal shell that almost completely surrounds the plasma. By heating themore » shell, a lithium coating on the plasma-facing side can be kept liquefied. A consequence of the low-recycling conditions from liquid lithium walls is the need for efficient plasma fueling. For this purpose, a molecular cluster injector is being developed. Future plans include the installation of a neutral beam for core plasma fueling, and also ion temperature measurements using charge-exchange recombination spectroscopy. Low edge recycling is also predicted to reduce temperature gradients that drive drift wave turbulence. Gyrokinetic simulations are in progress to calculate fluctuation levels and transport for LTX plasmas, and new fluctuation diagnostics are under development to test these predictions. __________________________________________________« less

Rotation profile flattening and toroidal flow shear reversal due to the coupling of magnetic islands in tokamaks

DOE PAGES

Tobias, B.; Chen, M.; Classen, I. G. J.; ...

2016-04-15

The electromagnetic coupling of helical modes, including those having different toroidal mode numbers, modifies the distribution of toroidal angular momentum in tokamak discharges. This can have deleterious effects on other transport channels as well as on magnetohydrodynamic (MHD) stability and disruptivity. At low levels of externally injected momentum, the coupling of core-localized modes initiates a chain of events, whereby flattening of the core rotation profile inside successive rational surfaces leads to the onset of a large m/n = 2/1 tearing mode and locked-mode disruption. Furthermore, with increased torque from neutral beam injection, neoclassical tearing modes in the core may phase-lockmore » to each other without locking to external fields or structures that are stationary in the laboratory frame. The dynamic processes observed in these cases are in general agreement with theory, and detailed diagnosis allows for momentum transport analysis to be performed, revealing a significant torque density that peaks near the 2/1 rational surface. However, as the coupled rational surfaces are brought closer together by reducing q95, additional momentum transport in excess of that required to attain a phase-locked state is sometimes observed. Rather than maintaining zero differential rotation (as is predicted to be dynamically stable by single-fluid, resistive MHD theory), these discharges develop hollow toroidal plasma fluid rotation profiles with reversed plasma flow shear in the region between the m/n = 3/2 and 2/1 islands. Additional forces expressed in this state are not readily accounted for, and therefore, analysis of these data highlights the impact of mode coupling on torque balance and the challenges associated with predicting the rotation dynamics of a fusion reactor-a key issue for ITER. Published by AIP Publishing.« less

Rotation profile flattening and toroidal flow shear reversal due to the coupling of magnetic islands in tokamaks

DOE Office of Scientific and Technical Information (OSTI.GOV)

Tobias, B.; Grierson, B. A.; Okabayashi, M.

2016-05-15

The electromagnetic coupling of helical modes, even those having different toroidal mode numbers, modifies the distribution of toroidal angular momentum in tokamak discharges. This can have deleterious effects on other transport channels as well as on magnetohydrodynamic (MHD) stability and disruptivity. At low levels of externally injected momentum, the coupling of core-localized modes initiates a chain of events, whereby flattening of the core rotation profile inside successive rational surfaces leads to the onset of a large m/n = 2/1 tearing mode and locked-mode disruption. With increased torque from neutral beam injection, neoclassical tearing modes in the core may phase-lock to each othermore » without locking to external fields or structures that are stationary in the laboratory frame. The dynamic processes observed in these cases are in general agreement with theory, and detailed diagnosis allows for momentum transport analysis to be performed, revealing a significant torque density that peaks near the 2/1 rational surface. However, as the coupled rational surfaces are brought closer together by reducing q{sub 95}, additional momentum transport in excess of that required to attain a phase-locked state is sometimes observed. Rather than maintaining zero differential rotation (as is predicted to be dynamically stable by single-fluid, resistive MHD theory), these discharges develop hollow toroidal plasma fluid rotation profiles with reversed plasma flow shear in the region between the m/n = 3/2 and 2/1 islands. The additional forces expressed in this state are not readily accounted for, and therefore, analysis of these data highlights the impact of mode coupling on torque balance and the challenges associated with predicting the rotation dynamics of a fusion reactor—a key issue for ITER.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Beidler, M. T.; Cassak, P. A.; Jardin, S. C.

We diagnose local properties of magnetic reconnection during a sawtooth crash employing the three-dimensional toroidal, extended-magnetohydrodynamic (MHD) code M3D-C 1. To do so, we sample simulation data in the plane in which reconnection occurs, the plane perpendicular to the helical (m, n) = (1, 1) mode at the q = 1 surface, where m and n are the poloidal and toroidal mode numbers and q is the safety factor. We study the nonlinear evolution of a particular test equilibrium in a non-reduced field representation using both resistive-MHD and extended-MHD models. We find growth rates for the extended-MHD reconnection process exhibitmore » a nonlinear acceleration and greatly exceed that of the resistive-MHD model, as is expected from previous experimental, theoretical, and computational work. We compare the properties of reconnection in the two simulations, revealing the reconnecting current sheets are locally different in the two models and we present the first observation of the quadrupole out-of-plane Hall magnetic field that appears during extended-MHD reconnection in a 3D toroidal simulation (but not in resistive-MHD). We also explore the dependence on toroidal angle of the properties of reconnection as viewed in the plane perpendicular to the helical magnetic field, finding qualitative and quantitative effects due to changes in the symmetry of the reconnection process. Furthermore, this study is potentially important for a wide range of magnetically confined fusion applications, from confirming simulations with extended-MHD effects are sufficiently resolved to describe reconnection, to quantifying local reconnection rates for purposes of understanding and predicting transport, not only at the q = 1 rational surface for sawteeth, but also at higher order rational surfaces that play a role in disruptions and edge-confinement degradation.« less

Toroid Joining Gun For Fittings And Couplings

NASA Technical Reports Server (NTRS)

Fox, Robert L.; Swaim, Robert J.; Johnson, Samuel D.; Buckley, John D.; Copeland, Carl E.; Coultrip, Robert H.; Johnston, David F.; Phillips, William M.

1992-01-01

Hand-held gun used to join metal heat-to-shrink couplings. Uses magnetic induction (eddy currents) to produce heat in metal coupling, and thermocouple to measure temperature and signals end of process. Gun, called "toroid joining gun" concentrates high levels of heat in localized areas. Reconfigured for use on metal heat-to-shrink fitting and coupling applications. Provides rapid heating, operates on low power, lightweight and portable. Safe for use around aircraft fuel and has no detrimental effects on surrounding surfaces or objects. Reliable in any environment and under all weather conditions. Gun logical device for taking full advantage of capabilities of new metal heat-to-shrink couplings and fittings.

Resonance transparency with low-loss in toroidal planar metamaterial

NASA Astrophysics Data System (ADS)

Xiang, Tianyu; Lei, Tao; Hu, Sen; Chen, Jiao; Huang, Xiaojun; Yang, Helin

2018-03-01

A compact planar construction composed of asymmetric split ring resonators was designed with a low-loss, high Q-factor resonance transparency at microwave frequency. The singularity property of the proposed metamaterial owing to the enhanced toroidal dipole T is demonstrated via numerical and experimental methods. The transmission peak can reach up to 0.91 and the loss is perfectly repressed, which can be testified by radiated power, H-field distributions, and the imaginary parts of effective permittivity and permeability. The designed planar metamaterial may have numerous potential applications at microwave, terahertz, and optical frequency, e.g., for ultrasensitive sensing, slow-light devices, lasing spacers, even invisible information transfer.

Toroidal midplane neutral beam armor and plasma limiter

DOEpatents

Kugel, Henry W.; Hand Jr, Samuel W.; Ksayian, Haig

1986-02-04

For use in a tokamak fusion reactor having a midplane magnetic coil on the inner wall of an evacuated toriodal chamber within which a neutral beam heated, fusing plasma is magnetically confined, a neutral beam armor shield and plasma limiter is provided on the inner wall of the toroidal chamber to shield the midplane coil from neutral beam shine-thru and plasma deposition. The armor shield/plasma limiter forms a semicircular enclosure around the midplane coil with the outer surface of the armor shield/plasma limiter shaped to match, as closely as practical, the inner limiting magnetic flux surface of the toroidally confined, indented, bean-shaped plasma. The armor shield/plasma limiter includes a plurality of semicircular graphite plates each having a pair of coupled upper and lower sections with each plate positioned in intimate contact with an adjacent plate on each side thereof so as to form a closed, planar structure around the entire outer periphery of the circular midplane coil. The upper and lower plate sections are adapted for coupling to heat sensing thermocouples and to a circulating water conduit system for cooling the armor shield/plasma limiter.The inner center portion of each graphite plate is adapted to receive and enclose a section of a circular diagnostic magnetic flux loop so as to minimize the power from the plasma confinement chamber incident upon the flux loop.

Toroidal midplane neutral beam armor and plasma limiter

DOEpatents

Kugel, Henry W.; Hand, Jr, Samuel W.; Ksayian, Haig

1986-01-01

For use in a tokamak fusion reactor having a midplane magnetic coil on the inner wall of an evacuated toriodal chamber within which a neutral beam heated, fusing plasma is magnetically confined, a neutral beam armor shield and plasma limiter is provided on the inner wall of the toroidal chamber to shield the midplane coil from neutral beam shine-thru and plasma deposition. The armor shield/plasma limiter forms a semicircular enclosure around the midplane coil with the outer surface of the armor shield/plasma limiter shaped to match, as closely as practical, the inner limiting magnetic flux surface of the toroidally confined, indented, bean-shaped plasma. The armor shield/plasma limiter includes a plurality of semicircular graphite plates each having a pair of coupled upper and lower sections with each plate positioned in intimate contact with an adjacent plate on each side thereof so as to form a closed, planar structure around the entire outer periphery of the circular midplane coil. The upper and lower plate sections are adapted for coupling to heat sensing thermocouples and to a circulating water conduit system for cooling the armor shield/plasma limiter.The inner center portion of each graphite plate is adapted to receive and enclose a section of a circular diagnostic magnetic flux loop so as to minimize the power from the plasma confinement chamber incident upon the flux loop.

The Grad-Shafranov Reconstruction of Toroidal Magnetic Flux Ropes: Method Development and Benchmark Studies

NASA Astrophysics Data System (ADS)

Hu, Qiang

2017-09-01

We develop an approach of the Grad-Shafranov (GS) reconstruction for toroidal structures in space plasmas, based on in situ spacecraft measurements. The underlying theory is the GS equation that describes two-dimensional magnetohydrostatic equilibrium, as widely applied in fusion plasmas. The geometry is such that the arbitrary cross-section of the torus has rotational symmetry about the rotation axis, Z, with a major radius, r0. The magnetic field configuration is thus determined by a scalar flux function, Ψ, and a functional F that is a single-variable function of Ψ. The algorithm is implemented through a two-step approach: i) a trial-and-error process by minimizing the residue of the functional F(Ψ) to determine an optimal Z-axis orientation, and ii) for the chosen Z, a χ2 minimization process resulting in a range of r0. Benchmark studies of known analytic solutions to the toroidal GS equation with noise additions are presented to illustrate the two-step procedure and to demonstrate the performance of the numerical GS solver, separately. For the cases presented, the errors in Z and r0 are 9° and 22%, respectively, and the relative percent error in the numerical GS solutions is smaller than 10%. We also make public the computer codes for these implementations and benchmark studies.

Nonlinear Plasma Response to Resonant Magnetic Perturbation in Rutherford Regime

NASA Astrophysics Data System (ADS)

Zhu, Ping; Yan, Xingting; Huang, Wenlong

2017-10-01

Recently a common analytic relation for both the locked mode and the nonlinear plasma response in the Rutherford regime has been developed based on the steady-state solution to the coupled dynamic system of magnetic island evolution and torque balance equations. The analytic relation predicts the threshold and the island size for the full penetration of resonant magnetic perturbation (RMP). It also rigorously proves a screening effect of the equilibrium toroidal flow. In this work, we test the theory by solving for the nonlinear plasma response to a single-helicity RMP of a circular-shaped limiter tokamak equilibrium with a constant toroidal flow, using the initial-value, full MHD simulation code NIMROD. Time evolution of the parallel flow or ``slip frequency'' profile and its asymptotic approach to steady state obtained from the NIMROD simulations qualitatively agree with the theory predictions. Further comparisons are carried out for the saturated island size, the threshold for full mode penetration, as well as the screening effects of equilibrium toroidal flow in order to understand the physics of nonlinear plasma response in the Rutherford regime. Supported by National Magnetic Confinement Fusion Science Program of China Grants 2014GB124002 and 2015GB101004, the 100 Talent Program of the Chinese Academy of Sciences, and U.S. Department of Energy Grants DE-FG02-86ER53218 and DE-FC02-08ER54975.

Flow profile measurement with multi-Mach probes on the HIST spherical torus device

NASA Astrophysics Data System (ADS)

Hashimoto, S.; Nishioka, T.; Ando, K.; Kikuchi, Y.; Fukumoto, N.; Nagata, M.

2008-11-01

Role of plasma flow during MHD relaxation and magnetic reconnection processes is still underlying physics. The HIST spherical torus can generate various spherical torus (ST) configurations by changing the external toroidal magnetic field. Especially, the flipped ST (F-ST) configuration has been for the first time found in the HIST device [1]. In the present study, plasma flow measurements were performed by multi-Mach probes in the ST and the F-ST configurations. In addition, the measured plasma flow was compared with that evaluated by an ion Doppler spectrometer (IDS) system and plasma images measured by a high-speed camera. As the result, it was shown that the toroidal plasma flow (˜ 20 km/s) at the location far from the plasma gun was clearly reversed after the transition from the ST to the F-ST. However, the direction of the toroidal flow was not changed near the plasma gun. Therefore, it can be considered that there are flipped and non-reversal regions in the plasma. The result agrees well with a magnetic configuration predicted by magnetic field measurements. The plasma images measured by the high-speed camera also indicated that a helically twisted structure appeared from the gun region, and it localized at the edge region. [1] M. Nagata et al., Phys. Rev. Lett. 90, pp. 225001-225004 (2003).

Ion flow measurements during the rotating kink behavior of the central column in the HIST device

NASA Astrophysics Data System (ADS)

Yamada, S.; Yoshikawa, T.; Hashimoto, S.; Nishioka, T.; Kikuchi, Y.; Fukumoto, N.; Nagata, M.

2007-11-01

Plasma flow is essentially driven in self-organization and magnetic reconnection process of compact spherical torus (ST) and spheromak in the helicity-driven systems. For example, when reversing the external toroidal field of ST, the direction not only of the plasma current but also of the toroidal ion flow is self-reversed during the formation of the flipped ST relaxed states. Mach probe measurement shows that the velocity of the ion flow reversed after the flip increases to about 20 km/s. We have been newly developing an ion Doppler spectrometer (IDS) system using a compact 16 or 64 channel photomultiplier tube (PMT) in order to measure the spatial profile of ion temperature and rotation velocity in the HIST device. The IDS system consists of a light collection system including optical fibers, 1 m-spectrometer and the PMT detector. The optical fibers covered with glass tubes are inserted into the plasma. The glass tubes can be rotated in the poloidal and the toroidal directions. The new IDS system will be applied to observations of ion temperature and plasma rotation in the flipped ST formation and in the MHD control of kinking behaviors of the central column by using the rotating magnetic field (RMF). Preliminary IDS results will be compared to those from Mach probe measurements in space.

Nonlinear MHD simulation of magnetic relaxation during DC helicity injection in spherical torus plasmas

NASA Astrophysics Data System (ADS)

Kanki, Takashi; Nagata, Masayoshi; Kagei, Yasuhiro

2009-11-01

Recently, the intermittent plasma flow has been observed to be correlated with the fluctuations of the toroidal current It and n=1 mode in the HIST spherical torus device. During the partially driven phase mixed with a resistive decay, the toroidal ion flow velocity (˜ 40 km/s) in the opposite direction of It is driven in the central open flux region, and the oscillations in n=1 mode occur there, while during the resistive decay phase, this flow velocity reverses and results in the same as that of It, and the oscillations in n=1 mode disappear there. The purpose of the present study is to investigate the plasma flow reversal process and the relevant MHD relaxation by using the 3-D nonlinear MHD simulations. The numerical results exhibit that during the driven phase, the toroidal flow velocity (˜ 37 km/s) is in the opposite direction to It, but in the same direction as the ExB rotation induced by an applied voltage. This flow is driven by the magnetic reconnection occurring at the X-point during the repetitive process of the non-axisymmetric magnetized plasmoid ejection from the helicity injector. The oscillations of poloidal flux ψp are out of phase with those of toroidal flux ψt and magnetic energy for the dominant n=1 mode, indicating the flux conversion from ψt to ψp. The effect of the vacuum toroidal field strength on the plasma dynamics is discussed.

Study of edge turbulence in dimensionally similar laboratory plasmas

NASA Astrophysics Data System (ADS)

Stroth, Ulrich

2003-10-01

In recent years, the numerical simulation of turbulence has made considerable progress. Predictions are made for large plasma volumes taking into account realistic magnetic geometries. Because of diagnostic limitations, in fusion plasmas the means of experimental testing of the models are rather limited. Toroidal low-temperature plasmas offer the possibility for detailed comparisons between experiment and simulation. Due to the reduced plasma parameters, the relevant quantities can be measured in the entire plasma. At the same time, the relevant non-dimensional parameters can be comparable to those in the edge of fusion plasmas. This presentation reports on results from the torsatron TJ-K [1,2] operated with a low-temperature plasma. The data are compared with simulations using the drift-Alfven-wave code DALF3 [3]. Langmuir probe arrays with 64 tips are used to measure the spatial structure of the turbulence. The same analyses techniques are applied to experimental and numerical data. The measured properties of spectra and probability density functions are reproduced by the code. Although the plasma in experiment and simulation does not exhibit critical pressure gradients, the radial transport fluctuations are strongly intermittent in both cases. Using Hydrogen, Helium and Argon as working gases, the scale parameter ρs could be varied by more than a factor of ten. As predicted by theory, the size of the turbulent eddies increases with ρ_s. The measured cross-phase between density and potential fluctuations are small, indicating the importance of the drift-wave dynamics for the turbulence in toroidal plasmas. The wave number spectra decay with an exponent of -3 as one would expect for the enstrophy cascade in 2D turbulence. [1] N. Krause et al., Rev. Sci. Instr. 73, 3474 (2002) [2] C. Lechte et al., New J. of Physics 4, 34 (2002) [3] B. Scott, Plasma Phys. Contr. Fusion 39, 1635 (1997)

Determination of plasma displacement based on eddy current diagnostics for the Keda Torus eXperiment

NASA Astrophysics Data System (ADS)

Tu, Cui; Li, Hong; Liu, Adi; Li, Zichao; Zhang, Yuan; You, Wei; Tan, Mingsheng; Luo, Bing; Adil, Yolbarsop; Hu, Jintong; Wu, Yanqi; Yan, Wentan; Xie, Jinlin; Lan, Tao; Mao, Wenzhe; Ding, Weixing; Xiao, Chijin; Zhuang, Ge; Liu, Wandong

2017-10-01

The measurement of plasma displacement is one of the most basic diagnostic tools in the study of plasma equilibrium and control in a toroidal magnetic confinement configuration. During pulse discharge, the eddy current induced in the vacuum vessel and shell will produce an additional magnetic field at the plasma boundary, which will have a significant impact on the measurement of plasma displacement using magnetic probes. In the newly built Keda Torus eXperiment (KTX) reversed field pinch device, the eddy current in the composite shell can be obtained at a high spatial resolution. This device offers a new way to determine the plasma displacement for KTX through the multipole moment expansion of the eddy current, which can be obtained by unique probe arrays installed on the inner and outer surfaces of the composite shell. In an ideal conductor shell approximation, the method of multipole moment expansion of the poloidal eddy current for measuring the plasma displacement in toroidal coordinates, is more accurate than the previous method based on symmetrical magnetic probes, which yielded results in cylindrical coordinates. Through an analytical analysis of many current filaments and numerical simulations of the current distribution in toroidal coordinates, the scaling relation between the first moment of the eddy current and the center of gravity of the plasma current is obtained. In addition, the origin of the multipole moment expansion of the eddy current in KTX is retrieved simultaneously. Preliminary data on the plasma displacement have been collected using these two methods during short pulse discharges in the KTX device, and the results of the two methods are in reasonable agreement.

MHD Equilibrium with Reversed Current Density and Magnetic Islands Revisited: the Vacuum Vector Potential Calculus

NASA Astrophysics Data System (ADS)

L. Braga, F.

2013-10-01

The solution of Grad-Shafranov equation determines the stationary behavior of fusion plasma inside a tokamak. To solve the equation it is necessary to know the toroidal current density profile. Recent works show that it is possible to determine a magnetohydrodynamic (MHD) equilibrium with reversed current density (RCD) profiles that presents magnetic islands. In this work we show analytical MHD equilibrium with a RCD profile and analyze the structure of the vacuum vector potential associated with these equilibria using the virtual casing principle.

[Hybrid stabilization technique with spinal fusion and interlaminar device to reduce the length of fusion and to protect symptomatic adjacent segments : Clinical long-term follow-up].

PubMed

Fleege, C; Rickert, M; Werner, I; Rauschmann, M; Arabmotlagh, M

2016-09-01

Determination of the extent of spinal fusion for lumbar degenerative diseases is often difficult due to minor pathologies in the adjacent segment. Although surgical intervention is required, fusion seems to be an overtreatment. Decompression alone may be not enough as this segment is affected by multiple factors such as destabilization, low grade degeneration and an unfavorable biomechanical transition next to a rigid construct. An alternative surgical treatment is a hybrid construct, consisting of fusion and implantation of an interlaminar stabilization device at the adjacent level. The aim of this study was to compare long-term clinical outcome after lumbar fusion with a hybrid construct including an interlaminar stabilization device as "topping-off". A retrospective analysis of 25 lumbar spinal fusions from 2003 to 2010 with additional interlaminar stabilization device was performed. Through a matched case controlled procedure 25 congruent patients who received lumbar spinal fusion in one or two levels were included as a control group. At an average follow-up of 43 months pre- and postoperative pain, ODI, SF-36 as well as clinical parameters, such as leg and back pain, walking distance and patient satisfaction were recorded. Pain relief, ODI improvement and patient satisfaction was significantly higher in the hybrid group compared to the control group. SF-36 scores improved in both groups but was higher in the hybrid group, although without significance. Evaluation of walking distance showed no significant differences. Many outcome parameters present significantly better long-term results in the hybrid group compared to sole spinal fusion. Therefore, in cases with a clear indication for lumbar spinal fusion with the need for decompression at the adjacent level due to spinal stenosis or moderate spondylarthrosis, support of this segment with an interlaminar stabilization device demonstrates a reasonable treatment option with good clinical outcome. Also, the length of the fusion construct can be reduced allowing for a softer and more harmonic transition.

Resonant Transparency and Non-Trivial Non-Radiating Excitations in Toroidal Metamaterials

PubMed Central

Fedotov, V. A.; Rogacheva, A. V.; Savinov, V.; Tsai, D. P.; Zheludev, N. I.

2013-01-01

Engaging strongly resonant interactions allows dramatic enhancement of functionalities of many electromagnetic devices. However, resonances can be dampened by Joule and radiation losses. While in many cases Joule losses may be minimized by the choice of constituting materials, controlling radiation losses is often a bigger problem. Recent solutions include the use of coupled radiant and sub-radiant modes yielding narrow asymmetric Fano resonances in a wide range of systems, from defect states in photonic crystals and optical waveguides with mesoscopic ring resonators to nanoscale plasmonic and metamaterial systems exhibiting interference effects akin to electromagnetically-induced transparency. Here we demonstrate theoretically and confirm experimentally a new mechanism of resonant electromagnetic transparency, which yields very narrow isolated symmetric Lorentzian transmission lines in toroidal metamaterials. It exploits the long sought non-trivial non-radiating charge-current excitation based on interfering electric and toroidal dipoles that was first proposed by Afanasiev and Stepanovsky in [J. Phys. A Math. Gen. 28, 4565 (1995)]. PMID:24132231

Non-axisymmetric equilibrium reconstruction on the Compact Toroidal Hybrid Experiment using external magnetic and soft x-ray inversion radius measurements

NASA Astrophysics Data System (ADS)

Ma, X.; Cianciosa, M.; Hanson, J. D.; Hartwell, G. J.; Knowlton, S. F.; Maurer, D. A.; Ennis, D. A.; Herfindal, J. L.

2015-11-01

Non-axisymmetric free-boundary equilibrium reconstructions of stellarator plasmas are performed for discharges in which the magnetic configuration is strongly modified by the driven plasma current. Studies were performed on the Compact Toroidal Hybrid device using the V3FIT reconstruction code incorporating a set of 50 magnetic diagnostics external to the plasma, combined with information from soft X-ray (SXR) arrays. With the assumption of closed magnetic flux surfaces, the reconstructions using external magnetic measurements allow accurate estimates of the net toroidal flux within the last closed flux surface, the edge safety factor, and the outer boundary of these highly non-axisymmetric plasmas. The inversion radius for sawtoothing plasmas is used to identify the location of the q = 1 surface, and thus infer the current profile near the magnetic axis. With external magnetic diagnostics alone, we find the reconstruction to be insufficiently constrained. This work is supported by US Department of Energy Grant No. DE-FG02-00ER54610.

Chaos of energetic positron orbits in a dipole magnetic field and its potential application to a new injection scheme

NASA Astrophysics Data System (ADS)

Saitoh, H.; Yoshida, Z.; Yano, Y.; Nishiura, M.; Kawazura, Y.; Horn-Stanja, J.; Pedersen, T. Sunn

2016-10-01

We study the behavior of high-energy positrons emitted from a radioactive source in a magnetospheric dipole field configuration. Because the conservation of the first and second adiabatic invariants is easily destroyed in a strongly inhomogeneous dipole field for high-energy charged particles, the positron orbits are nonintegrable, resulting in chaotic motions. In the geometry of a typical magnetospheric levitated dipole experiment, it is shown that a considerable ratio of positrons from a 22Na source, located at the edge of the confinement region, has chaotic long orbit lengths before annihilation. These particles make multiple toroidal circulations and form a hollow toroidal positron cloud. Experiments with a small 22Na source in the Ring Trap 1 (RT-1) device demonstrated the existence of such long-lived positrons in a dipole field. Such a chaotic behavior of high-energy particles is potentially applicable to the formation of a dense toroidal positron cloud in the strong-field region of the dipole field in future studies.

Alfvén eigenmode evolution computed with the VENUS and KINX codes for the ITER baseline scenario

DOE Office of Scientific and Technical Information (OSTI.GOV)

Isaev, M. Yu., E-mail: isaev-my@nrcki.ru; Medvedev, S. Yu.; Cooper, W. A.

A new application of the VENUS code is described, which computes alpha particle orbits in the perturbed electromagnetic fields and its resonant interaction with the toroidal Alfvén eigenmodes (TAEs) for the ITER device. The ITER baseline scenario with Q = 10 and the plasma toroidal current of 15 MA is considered as the most important and relevant for the International Tokamak Physics Activity group on energetic particles (ITPA-EP). For this scenario, typical unstable TAE-modes with the toroidal index n = 20 have been predicted that are localized in the plasma core near the surface with safety factor q = 1.more » The spatial structure of ballooning and antiballooning modes has been computed with the ideal MHD code KINX. The linear growth rates and the saturation levels taking into account the damping effects and the different mode frequencies have been calculated with the VENUS code for both ballooning and antiballooning TAE-modes.« less

Beam current sensor

DOEpatents

Kuchnir, M.; Mills, F.E.

1984-09-28

A current sensor for measuring the dc component of a beam of charged particles employs a superconducting pick-up loop probe, with twisted superconducting leads in combination with a Superconducting Quantum Interference Device (SQUID) detector. The pick-up probe is in the form of a single-turn loop, or a cylindrical toroid, through which the beam is directed and within which a first magnetic flux is excluded by the Meisner effect. The SQUID detector acts as a flux-to-voltage converter in providing a current to the pick-up loop so as to establish a second magnetic flux within the electrode which nulls out the first magnetic flux. A feedback voltage within the SQUID detector represents the beam current of the particles which transit the pick-up loop. Meisner effect currents prevent changes in the magnetic field within the toroidal pick-up loop and produce a current signal independent of the beam's cross-section and its position within the toroid, while the combination of superconducting elements provides current measurement sensitivities in the nano-ampere range.

Magnetic Effects in a Moderate-Temperature, High-Beta, Toroidal Plasma Device

NASA Astrophysics Data System (ADS)

Edwards, W. F.; Singh, A. K.; Held, E. D.

2011-10-01

A small toroidal machine (STOR-1M; minor radius 4.5 cm), on loan from the University of Saskatchewan, has been modified to operate at hydrogen ionization levels ~0.1%, beta values between 0.1 and 1, electron number density ~5x1016/m3, temperature ~5 eV, and applied toroidal magnetic field ~20 gauss. Plasma is generated using magnetron-produced microwaves. Langmuir and Hall probes determine radial profiles of electron number density, temperature, and magnetic field. For most values of the externally-applied magnetic field, the internal field is the same with or without plasma, however, in a narrow window of B, diamagnetism and other effects are present. The effect is observed with no externally induced current; plasma currents are self generated through some sort of relaxation process. Beta and radius conditions correlate well with similar magnetic structures in the laboratory (eg., plasma focus, Z pinch) and in space (eg., Venus flux ropes, solar coronal loops).

Beam current sensor

DOEpatents

Kuchnir, Moyses; Mills, Frederick E.

1987-01-01

A current sensor for measuring the DC component of a beam of charged particles employs a superconducting pick-up loop probe, with twisted superconducting leads in combination with a Superconducting Quantum Interference Device (SQUID) detector. The pick-up probe is in the form of a single-turn loop, or a cylindrical toroid, through which the beam is directed and within which a first magnetic flux is excluded by the Meisner effect. The SQUID detector acts as a flux-to-voltage converter in providing a current to the pick-up loop so as to establish a second magnetic flux within the electrode which nulls out the first magnetic flux. A feedback voltage within the SQUID detector represents the beam current of the particles which transit the pick-up loop. Meisner effect currents prevent changes in the magnetic field within the toroidal pick-up loop and produce a current signal independent of the beam's cross-section and its position within the toroid, while the combination of superconducting elements provides current measurement sensitivites in the nano-ampere range.

Effects of Density and Impurity on Edge Localized Modes in Tokamaks

NASA Astrophysics Data System (ADS)

Zhu, Ping

2017-10-01

Plasma density and impurity concentration are believed to be two of the key elements governing the edge tokamak plasma conditions. Optimal levels of plasma density and impurity concentration in the edge region have been searched for in order to achieve the desired fusion gain and divertor heat/particle load mitigation. However, how plasma density or impurity would affect the edge pedestal stability may have not been well known. Our recent MHD theory modeling and simulations using the NIMROD code have found novel effects of density and impurity on the dynamics of edge-localized modes (ELMs) in tokamaks. First, previous MHD analyses often predict merely a weak stabilizing effect of toroidal flow on ELMs in experimentally relevant regimes. We find that the stabilizing effects on the high- n ELMs from toroidal flow can be significantly enhanced with the increased edge plasma density. Here n denotes the toroidal mode number. Second, the stabilizing effects of the enhanced edge resistivity due to lithium-conditioning on the low- n ELMs in the high confinement (H-mode) discharges in NSTX have been identified. Linear stability analysis of the experimentally constrained equilibrium suggests that the change in the equilibrium plasma density and pressure profiles alone due to lithium-conditioning may not be sufficient for a complete suppression of the low- n ELMs. The enhanced resistivity due to the increased effective electric charge number Zeff after lithium-conditioning provides additional stabilization of the low- n ELMs. These new effects revealed in our theory analyses may help further understand recent ELM experiments and suggest new control schemes for ELM suppression and mitigation in future experiments. They may also pose additional constraints on the optimal levels of plasma density and impurity concentration in the edge region for H-mode tokamak operation. Supported by National Magnetic Confinement Fusion Science Program of China Grants 2014GB124002 and 2015GB101004, the 100 Talent Program of the Chinese Academy of Sciences, and U.S. Department of Energy Grants DE-FG02-86ER53218 and DE-FC02-08ER54975.

Interaction of rotating helical magnetic field with the HIST spherical torus plasmas

NASA Astrophysics Data System (ADS)

Kikuchi, Yusuke; Sugahara, Masato; Yamada, Satoshi; Yoshikawa, Tatsuya; Fukumoto, Naoyuki; Nagata, Masayoshi

2006-10-01

The physical mechanism of current drive by co-axial helicity injection (CHI) has been experimentally investigated on both spheromak and spherical torus (ST) configurations on the HIST device [1]. It has been observed that the n = 1 kink mode rotates toroidally with a frequency of 10-20 kHz in the ExB direction. It seems that the induced toroidal current by CHI strongly relates with the observed rotating kink mode. On the other hand, it is well known that MHD instabilities can be controlled or even suppressed by an externally applied helical magnetic field in tokamak devices. Therefore, we have started to install two sets of external helical coils in order to produce a rotating helical magnetic field on HIST. Mode structures of the generated rotating helical magnetic field and preliminary experimental results of the interaction of the rotating helical magnetic field with the HIST plasmas will be shown in the conference. [1] M. Nagata, et al., Physics of Plasmas 10, 2932 (2003)

Gaseous detonation initiation via wave implosion

NASA Astrophysics Data System (ADS)

Jackson, Scott Irving

Efficient detonation initiation is a topic of intense interest to designers of pulse detonation engines. This experimental work is the first to detonate propane-air mixtures with an imploding detonation wave and to detonate a gas mixture with a non-reflected, imploding shock. In order to do this, a unique device has been developed that is capable of generating an imploding toroidal detonation wave inside of a tube from a single ignition point without any obstruction to the tube flow path. As part of this study, an initiator that creates a large-aspect-ratio planar detonation wave in gas-phase explosive from a single ignition point has also been developed. The effectiveness of our initiation devices has been evaluated. The minimum energy required by the imploding shock for initiation was determined to scale linearly with the induction zone length, indicating the presence of a planar initiation mode. The imploding toroidal detonation initiator was found to be more effective at detonation initiation than the imploding shock initiator, using a comparable energy input to that of current initiator tubes.

Frontier of Fusion Research: Path to the Steady State Fusion Reactor by Large Helical Device

NASA Astrophysics Data System (ADS)

Motojima, Osamu

2006-12-01

The ITER, the International Thermonuclear Experimental Reactor, which will be built in Cadarache in France, has finally started this year, 2006. Since the thermal energy produced by fusion reactions divided by the external heating power, i.e., the Q value, will be larger than 10, this is a big step of the fusion research for half a century trying to tame the nuclear fusion for the 6.5 Billion people on the Earth. The source of the Sun's power is lasting steadily and safely for 8 Billion years. As a potentially safe environmentally friendly and economically competitive energy source, fusion should provide a sustainable future energy supply for all mankind for ten thousands of years. At the frontier of fusion research important milestones are recently marked on a long road toward a true prototype fusion reactor. In its own merits, research into harnessing turbulent burning plasmas and thereby controlling fusion reaction, is one of the grand challenges of complex systems science. After a brief overview of a status of world fusion projects, a focus is given on fusion research at the National Institute for Fusion Science (NIFS) in Japan, which is playing a role of the Inter University Institute, the coordinating Center of Excellence for academic fusion research and by the Large Helical Device (LHD), the world's largest superconducting heliotron device, as a National Users' facility. The current status of LHD project is presented focusing on the experimental program and the recent achievements in basic parameters and in steady state operations. Since, its start in a year 1998, a remarkable progress has presently resulted in the temperature of 140 Million degree, the highest density of 500 Thousand Billion/cc with the internal density barrier (IDB) and the highest steady average beta of 4.5% in helical plasma devices and the largest total input energy of 1.6 GJ, in all magnetic confinement fusion devices. Finally, a perspective is given of the ITER Broad Approach program as an integrated part of ITER and Development of Fusion Energy project Agreement. Moreover, the relationship with the NIFS' new parent organization the National Institutes of Natural Sciences and with foreign research institutions is briefly explained.

On fully three-dimensional resistive wall mode and feedback stabilization computationsa)

NASA Astrophysics Data System (ADS)

Strumberger, E.; Merkel, P.; Sempf, M.; Günter, S.

2008-05-01

Resistive walls, located close to the plasma boundary, reduce the growth rates of external kink modes to resistive time scales. For such slowly growing resistive wall modes, the stabilization by an active feedback system becomes feasible. The fully three-dimensional stability code STARWALL, and the feedback optimization code OPTIM have been developed [P. Merkel and M. Sempf, 21st IAEA Fusion Energy Conference 2006, Chengdu, China (International Atomic Energy Agency, Vienna, 2006, paper TH/P3-8] to compute the growth rates of resistive wall modes in the presence of nonaxisymmetric, multiply connected wall structures and to model the active feedback stabilization of these modes. In order to demonstrate the capabilities of the codes and to study the effect of the toroidal mode coupling caused by multiply connected wall structures, the codes are applied to test equilibria using the resistive wall structures currently under debate for ITER [M. Shimada et al., Nucl. Fusion 47, S1 (2007)] and ASDEX Upgrade [W. Köppendörfer et al., Proceedings of the 16th Symposium on Fusion Technology, London, 1990 (Elsevier, Amsterdam, 1991), Vol. 1, p. 208].

On fully three-dimensional resistive wall mode and feedback stabilization computations

DOE Office of Scientific and Technical Information (OSTI.GOV)

Strumberger, E.; Merkel, P.; Sempf, M.

2008-05-15

Resistive walls, located close to the plasma boundary, reduce the growth rates of external kink modes to resistive time scales. For such slowly growing resistive wall modes, the stabilization by an active feedback system becomes feasible. The fully three-dimensional stability code STARWALL, and the feedback optimization code OPTIM have been developed [P. Merkel and M. Sempf, 21st IAEA Fusion Energy Conference 2006, Chengdu, China (International Atomic Energy Agency, Vienna, 2006, paper TH/P3-8] to compute the growth rates of resistive wall modes in the presence of nonaxisymmetric, multiply connected wall structures and to model the active feedback stabilization of these modes.more » In order to demonstrate the capabilities of the codes and to study the effect of the toroidal mode coupling caused by multiply connected wall structures, the codes are applied to test equilibria using the resistive wall structures currently under debate for ITER [M. Shimada et al., Nucl. Fusion 47, S1 (2007)] and ASDEX Upgrade [W. Koeppendoerfer et al., Proceedings of the 16th Symposium on Fusion Technology, London, 1990 (Elsevier, Amsterdam, 1991), Vol. 1, p. 208].« less

Investigation of diocotron modes in toroidally trapped electron plasmas using non-destructive method

NASA Astrophysics Data System (ADS)

Lachhvani, Lavkesh; Pahari, Sambaran; Sengupta, Sudip; Yeole, Yogesh G.; Bajpai, Manu; Chattopadhyay, P. K.

2017-10-01

Experiments with trapped electron plasmas in a SMall Aspect Ratio Toroidal device (SMARTEX-C) have demonstrated a flute-like mode represented by oscillations on capacitive (wall) probes. Although analogous to diocotron mode observed in linear electron traps, the mode evolution in toroids can have interesting consequences due to the presence of in-homogeneous magnetic field. In SMARTEX-C, the probe signals are observed to undergo transition from small, near-sinusoidal oscillations to large amplitude, non-linear "double-peaked" oscillations. To interpret the wall probe signal and bring forth the dynamics, an expression for the induced current on the probe for an oscillating charge is derived, utilizing Green's Reciprocation Theorem. Equilibrium position, poloidal velocity of the charge cloud, and charge content of the cloud, required to compute the induced current, are estimated from the experiments. Signal through capacitive probes is thereby computed numerically for possible charge cloud trajectories. In order to correlate with experiments, starting with an intuitive guess of the trajectory, the model is evolved and tweaked to arrive at a signal consistent with experimentally observed probe signals. A possible vortex like dynamics is predicted, hitherto unexplored in toroidal geometries, for a limited set of experimental observations from SMARTEX-C. Though heuristic, a useful interpretation of capacitive probe data in terms of charge cloud dynamics is obtained.

Fusion plasma theory project summaries

NASA Astrophysics Data System (ADS)

1993-10-01

This Project Summary book is a published compilation consisting of short descriptions of each project supported by the Fusion Plasma Theory and Computing Group of the Advanced Physics and Technology Division of the Department of Energy, Office of Fusion Energy. The summaries contained in this volume were written by the individual contractors with minimal editing by the Office of Fusion Energy. Previous summaries were published in February of 1982 and December of 1987. The Plasma Theory program is responsible for the development of concepts and models that describe and predict the behavior of a magnetically confined plasma. Emphasis is given to the modelling and understanding of the processes controlling transport of energy and particles in a toroidal plasma and supporting the design of the International Thermonuclear Experimental Reactor (ITER). A tokamak transport initiative was begun in 1989 to improve understanding of how energy and particles are lost from the plasma by mechanisms that transport them across field lines. The Plasma Theory program has actively participated in this initiative. Recently, increased attention has been given to issues of importance to the proposed Tokamak Physics Experiment (TPX). Particular attention has been paid to containment and thermalization of fast alpha particles produced in a burning fusion plasma as well as control of sawteeth, current drive, impurity control, and design of improved auxiliary heating. In addition, general models of plasma behavior are developed from physics features common to different confinement geometries. This work uses both analytical and numerical techniques. The Fusion Theory program supports research projects at U.S. government laboratories, universities and industrial contractors. Its support of theoretical work at universities contributes to the office of Fusion Energy mission of training scientific manpower for the U.S. Fusion Energy Program.

Non-solenoidal Plasma Startup in the Pegasus Toroidal Experiment

NASA Astrophysics Data System (ADS)

Sontag, Aaron

2008-11-01

Non-solenoidal (NS) startup will simplify the design of future tokamaks by eliminating need for a central solenoid and is required for an ST based CTF. In Pegasus, washer-stack current sources (plasma guns) are used to initiate NS discharges via point-source DC helicity injection. Current injected parallel to the helical vacuum field can relax into a tokamak-like configuration with toroidally-averaged closed flux and tokamak-like confinement. This requires no modification of the vacuum vessel and is scalable to fusion grade systems with proper geometry. Guns in the divertor region create discharges with Ip up to 50 kA, 3 times the vacuum windup. Nonlinear 3D simulation with NIMROD shows excitation of a line-tied kink, producing poloidal flux amplification. Evidence of flux amplification includes: reversal of edge poloidal magnetic flux; Ip increase over vacuum geometric windup; plasma position subject to radial force balance; and persistence of Ip after gun shut-off. Equilibria show high edge current (li = 0.2) and elevated q (qmin> 6), allowing access to high IN (IN> 12). Guns at the outboard midplane produce Ip up to 7 times the vacuum windup with large n=1 activity when edge q passes through rational surfaces. Line averaged density up to 2x10^19 m-3 after relaxation shows an increase in particle confinement over non-relaxed cases. Maximum Ip is determined by helicity and radial force balance, tokamak stability, and Taylor relaxation. Coupling midplane gun discharges to other CD is straightforward due to Ip decay times >3 ms. Poloidal field induction has been used to create NS discharges up to 80 kA and gun plasmas with Ip of 60 kA have been ramped to over 100 kA by including OH drive. Present research is aimed at understanding the physics of this technique in order to form NS targets in excess of 200 kA and design NS startup systems for larger devices.

Asymptotic and spectral analysis of the gyrokinetic-waterbag integro-differential operator in toroidal geometry

NASA Astrophysics Data System (ADS)

Besse, Nicolas; Coulette, David

2016-08-01

Achieving plasmas with good stability and confinement properties is a key research goal for magnetic fusion devices. The underlying equations are the Vlasov-Poisson and Vlasov-Maxwell (VPM) equations in three space variables, three velocity variables, and one time variable. Even in those somewhat academic cases where global equilibrium solutions are known, studying their stability requires the analysis of the spectral properties of the linearized operator, a daunting task. We have identified a model, for which not only equilibrium solutions can be constructed, but many of their stability properties are amenable to rigorous analysis. It uses a class of solution to the VPM equations (or to their gyrokinetic approximations) known as waterbag solutions which, in particular, are piecewise constant in phase-space. It also uses, not only the gyrokinetic approximation of fast cyclotronic motion around magnetic field lines, but also an asymptotic approximation regarding the magnetic-field-induced anisotropy: the spatial variation along the field lines is taken much slower than across them. Together, these assumptions result in a drastic reduction in the dimensionality of the linearized problem, which becomes a set of two nested one-dimensional problems: an integral equation in the poloidal variable, followed by a one-dimensional complex Schrödinger equation in the radial variable. We show here that the operator associated to the poloidal variable is meromorphic in the eigenparameter, the pulsation frequency. We also prove that, for all but a countable set of real pulsation frequencies, the operator is compact and thus behaves mostly as a finite-dimensional one. The numerical algorithms based on such ideas have been implemented in a companion paper [D. Coulette and N. Besse, "Numerical resolution of the global eigenvalue problem for gyrokinetic-waterbag model in toroidal geometry" (submitted)] and were found to be surprisingly close to those for the original gyrokinetic-Vlasov equations. The purpose of the present paper is to make these new ideas accessible to two readerships: applied mathematicians and plasma physicists.

Change of Paradigm for the Reversed Field Pinch

NASA Astrophysics Data System (ADS)

Escande, D. F.

2010-11-01

The reversed field pinch (RFP) is a magnetic configuration germane to the tokamak, but it produces most of its magnetic field by the currents flowing inside the plasma; external coils provide only a small edge toroidal field whose sign is reversed with respect to the central one, whence the name of the configuration. Because of the presence of magnetic turbulence and chaos, the RFP had been considered for a long period as a terrible confinement configuration. However, recently a change of paradigm occurred for this device. Indeed, when the toroidal current is increased in the RFX-mod RFP in Padua (Italy), a self-organized helical state with an internal transport barrier (ITB) develops, and a broad zone of the plasma becomes hot (above 1 keV for a magnetic field above 0.8 T). The present theoretical picture of the RFP mainly comes from three-dimensional nonlinear visco-resistive MHD simulations whose dynamics has strong similarities with the experimental one, and triggered the experimental search for RFP states with improved confinement. The RFP ohmic state involves a helical electrostatic potential generating, as an electric drift, the so-called dynamo velocity field. The magnetic topology can bifurcate from a magnetic island to kink-like magnetic surfaces with higher resilience to magnetic chaos. This theoretical scenario was found to be relevant when ITB's enclosing a broad hot domain were discovered. The ITBs occur in the vicinity of the maximum of the safety factor. The new paradigm for the RFP supports its reappraisal as a low-external field, non-disruptive, ohmically heated approach to magnetic fusion, exploiting both self-organization and technological simplicity. Furthermore the RFP has the same Greenwald density limit as the tokamak, and it is an excellent test bed for the efficient control of multiple resistive wall modes. Its helical magnetic structure makes it germane to the stellarator too. As a result the RFP is also useful to bring support to the present two main lines of magnetic confinement.

Neutronics Analysis of Water-Cooled Ceramic Breeder Blanket for CFETR

NASA Astrophysics Data System (ADS)

Zhu, Qingjun; Li, Jia; Liu, Songlin

2016-07-01

In order to investigate the nuclear response to the water-cooled ceramic breeder blanket models for CFETR, a detailed 3D neutronics model with 22.5° torus sector was developed based on the integrated geometry of CFETR, including heterogeneous WCCB blanket models, shield, divertor, vacuum vessel, toroidal and poloidal magnets, and ports. Using the Monte Carlo N-Particle Transport Code MCNP5 and IAEA Fusion Evaluated Nuclear Data Library FENDL2.1, the neutronics analyses were performed. The neutron wall loading, tritium breeding ratio, the nuclear heating, neutron-induced atomic displacement damage, and gas production were determined. The results indicate that the global TBR of no less than 1.2 will be a big challenge for the water-cooled ceramic breeder blanket for CFETR. supported by the National Magnetic Confinement Fusion Science Program of China (Nos. 2013GB108004, 2014GB122000, and 2014GB119000), and National Natural Science Foundation of China (No. 11175207)

Advanced tokamak research with integrated modeling in JT-60 Upgrade

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hayashi, N.

2010-05-15

Researches on advanced tokamak (AT) have progressed with integrated modeling in JT-60 Upgrade [N. Oyama et al., Nucl. Fusion 49, 104007 (2009)]. Based on JT-60U experimental analyses and first principle simulations, new models were developed and integrated into core, rotation, edge/pedestal, and scrape-off-layer (SOL)/divertor codes. The integrated models clarified complex and autonomous features in AT. An integrated core model was implemented to take account of an anomalous radial transport of alpha particles caused by Alfven eigenmodes. It showed the reduction in the fusion gain by the anomalous radial transport and further escape of alpha particles. Integrated rotation model showed mechanismsmore » of rotation driven by the magnetic-field-ripple loss of fast ions and the charge separation due to fast-ion drift. An inward pinch model of high-Z impurity due to the atomic process was developed and indicated that the pinch velocity increases with the toroidal rotation. Integrated edge/pedestal model clarified causes of collisionality dependence of energy loss due to the edge localized mode and the enhancement of energy loss by steepening a core pressure gradient just inside the pedestal top. An ideal magnetohydrodynamics stability code was developed to take account of toroidal rotation and clarified a destabilizing effect of rotation on the pedestal. Integrated SOL/divertor model clarified a mechanism of X-point multifaceted asymmetric radiation from edge. A model of the SOL flow driven by core particle orbits which partially enter the SOL was developed by introducing the ion-orbit-induced flow to fluid equations.« less

Novel Reactor Relevant RF Actuator Schemes for the Lower Hybrid and the Ion Cyclotron Range of Frequencies

NASA Astrophysics Data System (ADS)

Bonoli, Paul

2014-10-01

This paper presents a fresh physics perspective on the onerous problem of coupling and successfully utilizing ion cyclotron range of frequencies (ICRF) and lower hybrid range of frequencies (LHRF) actuators in the harsh environment of a nuclear fusion reactor. The ICRF and LH launchers are essentially first wall components in a fusion reactor and as such will be subjected to high heat fluxes. The high field side (HFS) of the plasma offers a region of reduced heat flux together with a quiescent scrape off layer (SOL). Placement of the ICRF and LHRF launchers on the tokamak HFS also offers distinct physics advantages: The higher toroidal magnetic field makes it possible to couple faster phase velocity LH waves that can penetrate farther into the plasma core and be absorbed by higher energy electrons, thereby increasing the current drive efficiency. In addition, re-location of the LH launcher off the mid-plane (i.e., poloidal ``steering'') allows further control of the deposition location. Also ICRF waves coupled from the HFS couple strongly to mode converted ion Bernstein waves and ion cyclotron waves waves as the minority density is increased, thus opening the possibility of using this scheme for flow drive and pressure control. Finally the quiescent nature of the HFS scrape off layer should minimize the effects of RF wave scattering from density fluctuations. Ray tracing / Fokker Planck simulations will be presented for LHRF applications in devices such as the proposed Advanced Divertor Experiment (ADX) and extending to ITER and beyond. Full-wave simulations will also be presented which demonstrate the possible combinations of electron and ion heating via ICRF mode conversion. Work supported by the US DoE under Contract Numbers DE-FC02-01ER54648 and DE-FC02-99ER54512.

Fusion proton diagnostic for the C-2 field reversed configurationa)

NASA Astrophysics Data System (ADS)

Magee, R. M.; Clary, R.; Korepanov, S.; Smirnov, A.; Garate, E.; Knapp, K.; Tkachev, A.

2014-11-01

Measurements of the flux of fusion products from high temperature plasmas provide valuable insights into the ion energy distribution, as the fusion reaction rate is a very sensitive function of ion energy. In C-2, where field reversed configuration plasmas are formed by the collision of two compact toroids and partially sustained by high power neutral beam injection [M. Binderbauer et al., Phys. Rev. Lett. 105, 045003 (2010); M. Tuszewski et al., Phys. Rev. Lett. 108, 255008 (2012)], measurements of DD fusion neutron flux are used to diagnose ion temperature and study fast ion confinement and dynamics. In this paper, we will describe the development of a new 3 MeV proton detector that will complement existing neutron detectors. The detector is a large area (50 cm2), partially depleted, ion implanted silicon diode operated in a pulse counting regime. While the scintillator-based neutron detectors allow for high time resolution measurements (˜100 kHz), they have no spatial or energy resolution. The proton detector will provide 10 cm spatial resolution, allowing us to determine if the axial distribution of fast ions is consistent with classical fast ion theory or whether anomalous scattering mechanisms are active. We will describe in detail the diagnostic design and present initial data from a neutral beam test chamber.

Conceptual design study for heat exhaust management in the ARC fusion pilot plant

NASA Astrophysics Data System (ADS)

Dennett, C. A.; Cao, N. M.; Creely, A. J.; Hecla, J.; Hoffman, H.; Kuang, A. Q.; Major, M.; Ruiz Ruiz, J.; Tinguely, R. A.; Tolman, E. A.; Brunner, D.; Labombard, B.; Sorbom, B. N.; Whyte, D. G.; Grover, P.; Laughman, C.

2017-10-01

The ARC pilot plant conceptual design study has been extended to explore solutions for managing heat exhaust resulting from 525 MW of fusion power in a compact (R 3.3 m) tokamak. Superconducting poloidal field coils are configured to produce double-null equilibria that support X-point target divertors while maintaining the original core plasma shape and toroidal field coil size. Long outer divertor legs are appended to the original vacuum vessel, providing both large surface areas for surface dissipation of radiative heat and significantly reduced neutron damage for divertor components. A molten salt FLiBe blanket adequately shields all superconductors and functions as a tritium breeder, with advanced neutronics calculations indicating a tritium breeding ratio of 1.08. In addition, FLiBe is used as the active coolant for the entire vessel. A tungsten swirl-tube cooling channel is implemented in the divertor, capable of exhausting 12 MW/m2, heat flux while keeping total FliBe pumping power below 1% of fusion power. Finally, three novel diagnostics are explored: Cherenkov radiation emitted in FLiBe to measure fusion reaction rate, microwave interferometry to measure divertor detachment front location, and IR imaging through the FLiBe blanket to monitor selected divertor ``hotspots.''

Measurement of scrape-off-layer current dynamics during MHD activity and disruptions in HBT-EP

NASA Astrophysics Data System (ADS)

Levesque, J. P.; Brooks, J. W.; Abler, M. C.; Bialek, J.; Byrne, P. J.; Hansen, C. J.; Hughes, P. E.; Mauel, M. E.; Navratil, G. A.; Rhodes, D. J.

2017-08-01

We report scrape-off layer (SOL) current measurements during magnetohydrodynamic (MHD) mode activity, resonant magnetic perturbations (RMPs), and disruptions in the High Beta Tokamak—Extended Pulse (HBT-EP) device. Currents are measured via segmented plasma current Rogowski coils, jumpers running toroidally between otherwise-isolated vessel sections, and a grounded electrode in the scrape-off layer. These currents strongly depend on the plasma’s major radius, and amplitude and phase of non-axisymmetric field components. SOL currents connecting through the vessel are seen to reach  ∼0.2{--}0.5 % of the plasma current during typical kink activity and RMPs. Plasma current asymmetries and scrape-off-layer currents generated during disruptions, which are commonly called halo currents, reach  ∼4 % of I p. Asymmetric toroidal currents between vessel sections rotate at tens of kHz through most of the current quench, then symmetrize once I p reaches  ∼30 % of its pre-disruptive value. Toroidal jumper currents oscillate between co- and counter-I p, with co-I p being dominant on average during disruptions. A relative increase in local plasma current measured by a segmented I p Rogowski coil correlates with counter-I p current in the nearest toroidal jumper. Measurements are interpreted in the context of two models that produce contrary predictions for the toroidal vessel current polarity during disruptions. Plasma current asymmetry measurements are consistent with both models, and SOL currents scale with plasma displacement toward the vessel wall. The design of an upcoming SOL current diagnostic and control upgrade is also briefly presented.

Investigation of electrostatic waves in the ion cyclotron range of frequencies in L-4 and ACT-1

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ono, Masayuki

Electrostatic waves in the ion cyclotron range of frequencies (ICRF) were studied in the Princeton L-4 and ACT-1 devices for approximately ten years, from 1975 to 1985. The investigation began in the L-4 linear device, looking for the parametric excitation of electrostatic ion cyclotron waves in multi-ion-species plasmas. In addition, this investigation verified multi-ion-species effects on the electrostatic ion cyclotron wave dispersion religion including the ion-ion hybrid resonance. Finite-Larmor-radius modification of the wave dispersion relation was also observed, even for ion temperatures of T{sub i} {approx} 1/40 eV. Taking advantage of the relatively high field and long device length ofmore » L-4, the existence of the cold electrostatic ion cyclotron wave (CES ICW) was verified. With the arrival of the ACT-1 toroidal device, finite-Larmor-radius (FLR) waves were studied in a relatively collisionless warm-ion hydrogen plasma. Detailed investigations of ion Bernstein waves (IBW) included the verification of mode-transformation in their launching, their wave propagation characteristics, their absorption, and the resulting ion heating. This basic physics activity played a crucial role in developing a new reactor heating concept termed ion Bernstein wave heating. Experimental research in the lower hybrid frequency range confirmed the existence of FLR effects near the lower hybrid resonance, predicted by Stix in 1965. In a neon plasma with a carefully placed phased wave exciter, the neutralized ion Bernstein wave was observed for the first time. Using a fastwave ICRF antenna, two parasitic excitation processes for IBW -- parametric instability and density-gradient-driven excitation -- were also discovered. In the concluding section of this paper, a possible application of externally launched electrostatic waves is suggested for helium ash removal from fusion reactor plasmas.« less

Investigation of electrostatic waves in the ion cyclotron range of frequencies in L-4 and ACT-1

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ono, Masayuki.

Electrostatic waves in the ion cyclotron range of frequencies (ICRF) were studied in the Princeton L-4 and ACT-1 devices for approximately ten years, from 1975 to 1985. The investigation began in the L-4 linear device, looking for the parametric excitation of electrostatic ion cyclotron waves in multi-ion-species plasmas. In addition, this investigation verified multi-ion-species effects on the electrostatic ion cyclotron wave dispersion religion including the ion-ion hybrid resonance. Finite-Larmor-radius modification of the wave dispersion relation was also observed, even for ion temperatures of T[sub i] [approx] 1/40 eV. Taking advantage of the relatively high field and long device length ofmore » L-4, the existence of the cold electrostatic ion cyclotron wave (CES ICW) was verified. With the arrival of the ACT-1 toroidal device, finite-Larmor-radius (FLR) waves were studied in a relatively collisionless warm-ion hydrogen plasma. Detailed investigations of ion Bernstein waves (IBW) included the verification of mode-transformation in their launching, their wave propagation characteristics, their absorption, and the resulting ion heating. This basic physics activity played a crucial role in developing a new reactor heating concept termed ion Bernstein wave heating. Experimental research in the lower hybrid frequency range confirmed the existence of FLR effects near the lower hybrid resonance, predicted by Stix in 1965. In a neon plasma with a carefully placed phased wave exciter, the neutralized ion Bernstein wave was observed for the first time. Using a fastwave ICRF antenna, two parasitic excitation processes for IBW -- parametric instability and density-gradient-driven excitation -- were also discovered. In the concluding section of this paper, a possible application of externally launched electrostatic waves is suggested for helium ash removal from fusion reactor plasmas.« less

Effects of Lumbar Fusion Surgery with ISOBAR Devices Versus Posterior Lumbar Interbody Fusion Surgery on Pain and Disability in Patients with Lumbar Degenerative Diseases: A Meta-Analysis.

PubMed

Su, Shu-Fen; Wu, Meng-Shan; Yeh, Wen-Ting; Liao, Ying-Chin

2018-06-01

Purpose/Aim: Lumbar degenerative diseases (LDDs) cause pain and disability and are treated with lumbar fusion surgery. The aim of this study was to evaluate the efficacy of lumbar fusion surgery with ISOBAR devices versus posterior lumbar interbody fusion (PLIF) surgery for alleviating LDD-associated pain and disability. We performed a literature review and meta-analysis conducted in accordance with Cochrane methodology. The analysis included Group Reading Assessment and Diagnostic Evaluation assessments, Jadad Quality Score evaluations, and Risk of Bias in Non-randomized Studies of Interventions assessments. We searched PubMed, MEDLINE, the Cumulative Index to Nursing and Allied Health Literature, the Cochrane Library, ProQuest, the Airiti Library, and the China Academic Journals Full-text Database for relevant randomized controlled trials and cohort studies published in English or Chinese between 1997 and 2017. Outcome measures of interest included general pain, lower back pain, and disability. Of the 18 studies that met the inclusion criteria, 16 examined general pain (802 patients), 5 examined lower back pain (274 patients), and 15 examined disability (734 patients). General pain, lower back pain, and disability scores were significantly lower after lumbar fusion surgery with ISOBAR devices compared to presurgery. Moreover, lumbar fusion surgery with ISOBAR devices was more effective than PLIF for decreasing postoperative disability, although it did not provide any benefit in terms of general pain or lower back pain. Lumbar fusion surgery with ISOBAR devices alleviates general pain, lower back pain, and disability in LDD patients and is superior to PLIF for reducing postoperative disability. Given possible publication bias, we recommend further large-scale studies.

Resonant optical transducers for in-situ gas detection

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bond, Tiziana C.; Cole, Garrett; Goddard, Lynford

Configurations for in-situ gas detection are provided, and include miniaturized photonic devices, low-optical-loss, guided-wave structures and state-selective adsorption coatings. High quality factor semiconductor resonators have been demonstrated in different configurations, such as micro-disks, micro-rings, micro-toroids, and photonic crystals with the properties of very narrow NIR transmission bands and sensitivity up to 10.sup.-9 (change in complex refractive index). The devices are therefore highly sensitive to changes in optical properties to the device parameters and can be tunable to the absorption of the chemical species of interest. Appropriate coatings applied to the device enhance state-specific molecular detection.

Soft x-ray pinhole imaging diagnostics for compact toroid plasmas

NASA Astrophysics Data System (ADS)

Crawford, E. A.; Taggart, D. P.; Bailey, A. D., III

1990-10-01

Soft x-ray pinhole imaging has recently become established as a valuable diagnostic for visualization of field reversed configuration (FRC) plasmas in the TRX-2, FRX-C/LSM devices. Gated MCP image converter devices with CsI cathodes and Be filters with a peak response around 11 nm wavelength are used for exposure durations ranging from a few tenths up to several microseconds. Results of experiments with single and Chevron channel plates are discussed along with estimates of linear exposure limitations with both film and CCD cameras as recording media. Plans for multiframe devices on the FRX-C/LSM and the LSX devices are also discussed.

Resonant optical transducers for in-situ gas detection

DOEpatents

Bond, Tiziana C; Cole, Garrett; Goddard, Lynford

2016-06-28

Configurations for in-situ gas detection are provided, and include miniaturized photonic devices, low-optical-loss, guided-wave structures and state-selective adsorption coatings. High quality factor semiconductor resonators have been demonstrated in different configurations, such as micro-disks, micro-rings, micro-toroids, and photonic crystals with the properties of very narrow NIR transmission bands and sensitivity up to 10.sup.-9 (change in complex refractive index). The devices are therefore highly sensitive to changes in optical properties to the device parameters and can be tunable to the absorption of the chemical species of interest. Appropriate coatings applied to the device enhance state-specific molecular detection.

Active control of multiple resistive wall modes

NASA Astrophysics Data System (ADS)

Brunsell, P. R.; Yadikin, D.; Gregoratto, D.; Paccagnella, R.; Liu, Y. Q.; Bolzonella, T.; Cecconello, M.; Drake, J. R.; Kuldkepp, M.; Manduchi, G.; Marchiori, G.; Marrelli, L.; Martin, P.; Menmuir, S.; Ortolani, S.; Rachlew, E.; Spizzo, G.; Zanca, P.

2005-12-01

A two-dimensional array of saddle coils at Mc poloidal and Nc toroidal positions is used on the EXTRAP T2R reversed-field pinch (Brunsell P R et al 2001 Plasma Phys. Control. Fusion 43 1457) to study active control of resistive wall modes (RWMs). Spontaneous growth of several RWMs with poloidal mode number m = 1 and different toroidal mode number n is observed experimentally, in agreement with linear MHD modelling. The measured plasma response to a controlled coil field and the plasma response computed using the linear circular cylinder MHD model are in quantitive agreement. Feedback control introduces a linear coupling of modes with toroidal mode numbers n, n' that fulfil the condition |n - n'| = Nc. Pairs of coupled unstable RWMs are present in feedback experiments with an array of Mc × Nc = 4 × 16 coils. Using intelligent shell feedback, the coupled modes are generally not controlled even though the field is suppressed at the active coils. A better suppression of coupled modes may be achieved in the case of rotating modes by using the mode control feedback scheme with individually set complex gains. In feedback with a larger array of Mc × Nc = 4 × 32 coils, the coupling effect largely disappears, and with this array, the main internal RWMs n = -11, -10, +5, +6 are all simultaneously suppressed throughout the discharge (7 8 wall times). With feedback there is a two-fold extension of the pulse length, compared to discharges without feedback.

Toroidal modeling of the n = 1 intrinsic error field correction experiments in EAST

NASA Astrophysics Data System (ADS)

Yang, Xu; Liu, Yueqiang; Sun, Youwen; Wang, Huihui; Gu, Shuai; Jia, Manni; Li, Li; Liu, Yue; Wang, Zhirui; Zhou, Lina

2018-05-01

The m/n = 2/1 resonant vacuum error field (EF) in the EAST tokamak experiments, inferred from the compass coil current amplitude and phase scan for mode locking, was found to depend on the parity between the upper and lower rows of the EF correction (EFC) coils (Wang et al 2016 Nucl. Fusion 56 066011). Here m and n are the poloidal and toroidal harmonic numbers in a torus, respectively. This experimental observation implies that the compass scan results cannot be simply interpreted as reflecting the true intrinsic EF. This work aims at understanding this puzzle, based on toroidal modeling of the EFC plasma discharge in EAST using the MARS-F code (Liu et al 2000 Phys. Plasmas 7 3681). By varying the amplitude and phase of the assumed n = 1 intrinsic vacuum EF with different poloidal spectra, and by computing the plasma response to the assumed EF, the compass scan predicted 2/1 EF, based on minimizing the computed resonant electromagnetic torque, can be made to match well with that of the EFC experiments using both even and odd parity coils. Moreover, the compass scan predicted vacuum EFs are found to be significantly differing from the true intrinsic EF used as input to the MARS-F code. While the puzzling result remains to be fully resolved, the results from this study offer an improved understanding of the EFC experiments and the compass scan technique for determining the intrinsic resonant EF.

Non-inductively driven tokamak plasmas at near-unity β t in the Pegasus toroidal experiment

DOE PAGES

Reusch, Joshua A.; Bodner, Grant M.; Bongard, Michael W.; ...

2018-03-14

Amore » major goal of the spherical tokamak (ST) research program is accessing a state of low internal inductance ℓ i , high elongation κ , and high toroidal and normalized beta ( β t and β N ) without solenoidal current drive. Local helicity injection (LHI) in the Pegasus ST [Garstka et al., Nucl. Fusion 46, S603 (2006)] provides non-solenoidally driven plasmas that exhibit these characteristics. LHI utilizes compact, edge-localized current sources for plasma startup and sustainment. It results in hollow current density profiles with low ℓ i . The low aspect ratio ( R 0 / a ~ 1.2 ) of Pegasus allows access to high κ and high normalized plasma currents I N = I p / a B T > 14 ). Magnetic reconnection during LHI provides auxiliary ion heating. Together, these features provide access to very high β t plasmas. Equilibrium analyses indicate that β t up to ~100% is achieved. Finally, these high β t discharges disrupt at the ideal no-wall β limit at β N ~ 7. « less

Fully Parallel MHD Stability Analysis Tool

NASA Astrophysics Data System (ADS)

Svidzinski, Vladimir; Galkin, Sergei; Kim, Jin-Soo; Liu, Yueqiang

2014-10-01

Progress on full parallelization of the plasma stability code MARS will be reported. MARS calculates eigenmodes in 2D axisymmetric toroidal equilibria in MHD-kinetic plasma models. It is a powerful tool for studying MHD and MHD-kinetic instabilities and it is widely used by fusion community. Parallel version of MARS is intended for simulations on local parallel clusters. It will be an efficient tool for simulation of MHD instabilities with low, intermediate and high toroidal mode numbers within both fluid and kinetic plasma models, already implemented in MARS. Parallelization of the code includes parallelization of the construction of the matrix for the eigenvalue problem and parallelization of the inverse iterations algorithm, implemented in MARS for the solution of the formulated eigenvalue problem. Construction of the matrix is parallelized by distributing the load among processors assigned to different magnetic surfaces. Parallelization of the solution of the eigenvalue problem is made by repeating steps of the present MARS algorithm using parallel libraries and procedures. Initial results of the code parallelization will be reported. Work is supported by the U.S. DOE SBIR program.

A Pre-ionization System to Limit Neutral Gas in a Compact Toroid Injector

NASA Astrophysics Data System (ADS)

Allfrey, Ian; Roche, Thomas; Matsumoto, Tadafumi; Garate, Eusebio; Gota, Hiroshi; Asai, Tomohiko; the TAE Team

2016-10-01

Fusion plasmas require long lifetimes and high temperatures, both of which are limited by particle loss, among other factors. Therefore, refueling a long-lived advanced beam-driven field-reversed configuration (FRC) plasma in C-2U is necessary, and injecting a supersonic compact toroid (CT) is an effective means of introducing particles into the FRC core. However, neutral gas that trails the CT into the target chamber cools the FRC. Pre-ionization (PI) system assists the break down between electrodes of the CT injector (CTI), so the amount of introduced gas can be lowered by up to a factor of two, effectively increasing the ionization fraction; thus, reducing the amount of neutral gas in the system. Additionally, the PI decreases the delay in CTI breakdown so a highly reproducible operation is achievable. The PI system consists of a fast, high voltage, pulse discharge circuit coupled to a Teflon insulated semi-rigid coaxial cable inserted into the CTI. System details and experimental data will be presented, in addition to issues such as the introduction of impurities and pre-ionizer lifetime.

Modeling of Feedback Stabilization of External MHD Modes in Toroidal Geometry

NASA Astrophysics Data System (ADS)

Chu, M. S.; Chance, M. S.; Okabayashi, M.

2000-10-01

The intelligent shell feedback scheme(C.M. Bishop, Plasma Phys. Contr. Nucl. Fusion 31), 1179 (1989). seeks to utilize external coils to suppress the unstable MHD modes slowed down by the resistive shell. We present a new formulation and numerical results of the interaction between the plasma and its outside vacuum region, with complete plasma response and the inclusion of a resistive vessel in general toroidal geometry. This is achieved by using the Green's function technique, which is a generalization of that previously used for the VACUUM(M.S. Chance, Phys. Plasmas 4), 2161 (1997). code and coupled with the ideal MHD code GATO. The effectiveness of different realizations of the intelligent shell concept is gauged by their ability to minimize the available free energy to drive the MHD mode. Computations indicate poloidal coverage of 30% of the total resistive wall surface area and 6 or 7 segments of ``intelligent coil'' arrays superimposed on the resistive wall will allow recovery of up to 90% the effectiveness of the ideal shell in stabilizing the ideal external kink.

Tridimensional Thermonuclear Instability in Subignited Plasmas and on the Surface of the Pulsars

NASA Astrophysics Data System (ADS)

Cardinali, A.; Coppi, B.

2017-10-01

Tridimensional modes involving an increase of the electron temperature can be excited as a result of alpha-particle heating in subignited D-T fusion burning plasmas when a nearly time- independent external source of heating is applied. The analyzed modes are shown to emerge from an axisymmetric toroidal configurations and are radially localized around rational magnetic surfaces corresponding to q(r =r0) =m0 /n0 where m0 and n0 are the relevant poloidal and toroidal mode numbers. The radial width of the mode is of the order of the thermal scale distance. The mode has a rather severe damping rate, that has to be overcome by the relevant heating rate. Thus the temperature range to be considered is that where the D-T plasma reactivity undergoes a relatively large increase as a function of temperature. This kind of theory has been applied to the plasmas that are envisioned to be associated with surface of pulsar and be subjects to (spatially) inhomogenous thermonuclear burning. Sponsored in part by the U.S. DoE.

The effect of a radial electric field on ripple-trapped ions observed by neutral particle fluxes

DOE Office of Scientific and Technical Information (OSTI.GOV)

Heikkinen, J.A.; Herrmann, W.; Kurki-Suonio, T.

1997-10-01

The effect of a radial electric field on nonthermal ripple-trapped ions is investigated using toroidal Monte Carlo simulations for edge tokamak plasmas. The increase in the neutral particle flux from the ions trapped in local magnetic wells observed by the charge exchange (CX) detector at a low confinement to high confinement transition at ASDEX (Axially Symmetric Divertor Experiment). Upgrade tokamak [{ital Proceedings of the 20th European Conference on Controlled Fusion and Plasma Physics}, Lisbon (European Physical Society, Petit-Lancy, Switzerland, 1993), Vol. 17C, Part I, p. 267] is reproduced in the simulations by turning on a radial electric field near themore » plasma periphery. The poloidal and toroidal angles at which the CX detector signal is most sensitive to the radial electric field are determined. A fast response time of the signal in the range of 50{endash}100 {mu}s to the appearance of the electric field can be found in the simulations with a relatively large half-width of the negative electric field region. {copyright} {ital 1997 American Institute of Physics.}« less

Non-inductively driven tokamak plasmas at near-unity β t in the Pegasus toroidal experiment

DOE Office of Scientific and Technical Information (OSTI.GOV)

Reusch, Joshua A.; Bodner, Grant M.; Bongard, Michael W.

Amore » major goal of the spherical tokamak (ST) research program is accessing a state of low internal inductance ℓ i , high elongation κ , and high toroidal and normalized beta ( β t and β N ) without solenoidal current drive. Local helicity injection (LHI) in the Pegasus ST [Garstka et al., Nucl. Fusion 46, S603 (2006)] provides non-solenoidally driven plasmas that exhibit these characteristics. LHI utilizes compact, edge-localized current sources for plasma startup and sustainment. It results in hollow current density profiles with low ℓ i . The low aspect ratio ( R 0 / a ~ 1.2 ) of Pegasus allows access to high κ and high normalized plasma currents I N = I p / a B T > 14 ). Magnetic reconnection during LHI provides auxiliary ion heating. Together, these features provide access to very high β t plasmas. Equilibrium analyses indicate that β t up to ~100% is achieved. Finally, these high β t discharges disrupt at the ideal no-wall β limit at β N ~ 7. « less

Toroidal Alfvénic Eigenmodes Driven by Energetic Particles with Maxwell and Slowing-down Distributions

NASA Astrophysics Data System (ADS)

Hou, Yawei; Zhu, Ping; Zou, Zhihui; Kim, Charlson C.; Hu, Zhaoqing; Wang, Zhengxiong

2016-10-01

The energetic-particle (EP) driven toroidal Alfvén eigenmodes (TAEs) in a circular-shaped large aspect ratio tokamak are studied using the hybrid kinetic-MHD model in the NIMROD code, where the EPs are advanced using the δf particle-in-cell (PIC) method and their kinetic effects are coupled to the bulk plasma through moment closures. Two initial distributions of EPs, Maxwell and slowing-down, are considered. The influence of EP parameters, including density, temperature and density gradient, on the frequency and the growth rate of TAEs are obtained and benchmarked with theory and gyrokinetic simulations for the Maxwell distribution with good agreement. When the density and temperature of EPs are above certain thresholds, the transition from TAE to energetic particle modes (EPM) occurs and the mode structure also changes. Comparisons between Maxwell and slowing-down distributions in terms of EP-driven TAEs and EPMs will also be presented and discussed. Supported by the National Magnetic Confinement Fusion Science Program of China Grant Nos. 2014GB124002 and 2015GB101004, and the Natural Science Foundation of China Grant No. 11205194.

Biomechanical Characteristics of an Integrated Lumbar Interbody Fusion Device

PubMed Central

Voronov, Leonard I.; Vastardis, Georgios; Zelenakova, Julia; Carandang, Gerard; Havey, Robert M.; Waldorff, Erik I.; Zindrick, Michael R.

2014-01-01

Introduction We hypothesized that an Integrated Lumbar Interbody Fusion Device (PILLAR SA, Orthofix, Lewisville, TX) will function biomechanically similar to a traditional anterior interbody spacer (PILLAR AL, Orthofix, Lewisville, TX) plus posterior instrumentation (FIREBIRD, Orthofix, Lewisville, TX). Purpose of this study was to determine if an Integrated Interbody Fusion Device (PILLAR SA) can stabilize single motion segments as well as an anterior interbody spacer (PILLAR AL) + pedicle screw construct (FIREBIRD). Methods Eight cadaveric lumbar spines (age: 43.9±4.3 years) were used. Each specimen's range of motion was tested in flexion-extension (FE), lateral bending (LB), and axial rotation (AR) under intact condition, after L4-L5 PILLAR SA with intervertebral screws and after L4-L5 360° fusion (PILLAR AL + Pedicle Screws and rods (FIREBIRD). Each specimen was tested in flexion (8Nm) and extension (6Nm) without preload (0 N) and under 400N of preload, in lateral bending (±6 Nm) and axial rotation (±5 Nm) without preload. Results Integrated fusion using the PILLAR SA device demonstrated statistically significant reductions in range of motion of the L4-L5 motion segment as compared to the intact condition for each test direction. PILLAR SA reduced ROM from 8.9±1.9 to 2.9±1.1° in FE with 400N follower preload (67.4%), 8.0±1.7 to 2.5±1.1° in LB, and 2.2±1.2 to 0.7±0.3° in AR. A comparison between the PILLAR SA integrated fusion device versus 360° fusion construct with spacer and bilateral pedicle screws was statistically significant in FE and LB. The 360° fusion yielded motion of 1.0±0.5° in FE, 1.0±0.8° in LB (p0.05). Conclusions The PILLAR SA resulted in motions of less than 3° in all modes of motion and was not as motion restricting as the traditional 360° using bilateral pedicle screws. The residual segmental motions compare very favorably with published biomechanical studies of other interbody integrated fusion devices. PMID:25694931

A condition for small bootstrap current in three-dimensional toroidal configurations

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mikhailov, M. I., E-mail: mikhaylov-mi@nrcki.ru; Nührenberg, J.; Zille, R.

2016-11-15

It is shown that, if the maximum of the magnetic field strength on a magnetic surface in a threedimensional magnetic confinement configuration with stellarator symmetry constitutes a line that is orthogonal to the field lines and crosses the symmetry line, then the bootstrap current density is smaller compared to that in quasi-axisymmetric (qa) [J. Nührenberg et al., in Proc. of Joint Varenna−Lausanne Int. Workshop on Theory of Fusion Plasmas, Varenna, 1994, p. 3] and quasi-helically (qh) symmetric [J. Nührenberg and R. Zille, Phys. Lett. A 129, 113 (1988)] configurations.

Neutron flux measurements around PLT

DOE Office of Scientific and Technical Information (OSTI.GOV)

Zankl, G.; Strachan, J.D.; Lewis, R.

1980-09-01

Using Indium activation foils, the toroidal and poloidal neutron emission patterns were determined for PLT plasmas which include ICRF and neutral beam heating. The activities produced the /sup 115/In (n,n') /sup 115m/In reaction were determined by counting the 336 keV ..gamma.. line of the /sup 115m/In decay. This activation cross section falls just below 2.5 MeV so that the influence of scattered neutrons of degraded energies is reduced. From the magnitude of the activity, the absolute calibration of the PLT fusion neutron emission is obtained with less than or equal to 40% accuracy.

Suppression of large edge-localized modes in high-confinement DIII-D plasmas with a stochastic magnetic boundary.

PubMed

Evans, T E; Moyer, R A; Thomas, P R; Watkins, J G; Osborne, T H; Boedo, J A; Doyle, E J; Fenstermacher, M E; Finken, K H; Groebner, R J; Groth, M; Harris, J H; La Haye, R J; Lasnier, C J; Masuzaki, S; Ohyabu, N; Pretty, D G; Rhodes, T L; Reimerdes, H; Rudakov, D L; Schaffer, M J; Wang, G; Zeng, L

2004-06-11

A stochastic magnetic boundary, produced by an applied edge resonant magnetic perturbation, is used to suppress most large edge-localized modes (ELMs) in high confinement (H-mode) plasmas. The resulting H mode displays rapid, small oscillations with a bursty character modulated by a coherent 130 Hz envelope. The H mode transport barrier and core confinement are unaffected by the stochastic boundary, despite a threefold drop in the toroidal rotation. These results demonstrate that stochastic boundaries are compatible with H modes and may be attractive for ELM control in next-step fusion tokamaks.

Toroid Joining Gun. [thermoplastic welding system using induction heating

NASA Technical Reports Server (NTRS)

Buckley, J. D.; Fox, R. L.; Swaim, R J.

1985-01-01

The Toroid Joining Gun is a low cost, self-contained, portable low powered (100-400 watts) thermoplastic welding system developed at Langley Research Center for joining plastic and composite parts using an induction heating technique. The device developed for use in the fabrication of large space sructures (LSST Program) can be used in any atmosphere or in a vacuum. Components can be joined in situ, whether on earth or on a space platform. The expanded application of this welding gun is in the joining of thermoplastic composites, thermosetting composites, metals, and combinations of these materials. Its low-power requirements, light weight, rapid response, low cost, portability, and effective joining make it a candidate for solving many varied and unique bonding tasks.

Demonstration of Tokamak Ohmic Flux Saving by Transient Coaxial Helicity Injection in the National Spherical Torus Experiment

NASA Astrophysics Data System (ADS)

Raman, R.; Mueller, D.; Nelson, B. A.; Jarboe, T. R.; Gerhardt, S.; Kugel, H. W.; Leblanc, B.; Maingi, R.; Menard, J.; Ono, M.; Paul, S.; Roquemore, L.; Sabbagh, S.; Soukhanovskii, V.

2010-03-01

Transient coaxial helicity injection (CHI) started discharges in the National Spherical Torus Experiment (NSTX) have attained peak currents up to 300 kA and when coupled to induction, it has produced up to 200 kA additional current over inductive-only operation. CHI in NSTX has shown to be energetically quite efficient, producing a plasma current of about 10 A/J of capacitor bank energy. In addition, for the first time, the CHI-produced toroidal current that couples to induction continues to increase with the energy supplied by the CHI power supply at otherwise similar values of the injector flux, indicating the potential for substantial current generation capability by CHI in NSTX and in future toroidal devices.

Characteristic time for halo current growth and rotation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Boozer, Allen H., E-mail: ahb17@columbia.edu

2015-10-15

A halo current flows for part of its path through the plasma edge and for part through the chamber walls and during tokamak disruptions can be as large as tenths of the plasma current. The primary interest in halo currents is the large force that they can exert on machine components particularly if the toriodal rotation of the halo current resonates with a natural oscillation frequency of the tokamak device. Halo currents arise when required to slow down the growth of a kink that is too unstable to be stabilized by the chamber walls. The width of the current channelmore » in the halo plasma is comparable to the amplitude of the kink, and the halo current grows linearly, not exponentially, in time. The current density in the halo is comparable to that of the main plasma body. The rocket force due to plasma flowing out of the halo and recombining on the chamber walls can cause the non-axisymmetric magnetic structure produced by the kink to rotate toroidally at a speed comparable to the halo speed of sound. Gerhardt's observations of the halo current in NSTX shot 141 687 [Nucl. Fusion 53, 023005 (2013)] illustrate many features of the theory of halo currents and are discussed as a summary of the theory.« less

Theory and observation of the onset of nonlinear structures due to eigenmode destabilization by fast ions in tokamaks

DOE PAGES

Duarte, V. N.; Berk, H. L.; Gorelenkov, N. N.; ...

2017-12-12

Alfvén waves can induce the ejection of fast ions in different forms in tokamaks. In order to develop predictive capabilities to anticipate the nature of fast ion transport, a methodology is proposed to differentiate the likelihood of energetic-particle-driven instabilities to produce frequency chirping or fixed-frequency oscillations. Furthermore, the proposed method employs numerically calculated eigenstructures and multiple resonance surfaces of a given mode in the presence of energetic ion drag and stochasticity (due to collisions and micro-turbulence). Toroidicity-induced, reversed-shear and beta-induced Alfvén-acoustic eigenmodes are used as examples. Waves measured in experiments are characterized, and compatibility is found between the proposed criterionmore » predictions and the experimental observation or lack of observation of chirping behavior of Alfvénic modes in different tokamaks. It is found that the stochastic diffusion due to micro-turbulence can be the dominant energetic particle detuning mechanism near the resonances in many plasma experiments, and its strength is the key as to whether chirping solutions are likely to arise. We proposed a criterion that constitutes a useful predictive tool in assessing whether the nature of the transport for fast ion losses in fusion devices will be dominated by convective or diffusive processes.« less

Inductive current startup in large tokamaks with expanding minor radius and RF assist

DOE Office of Scientific and Technical Information (OSTI.GOV)

Borowski, S.K.

1983-01-01

Auxiliary RF heating of electrons before and during the current rise phase of a large tokamak, such as the Fusion Engineering Device, is examined as a means of reducing both the initiation loop voltage and resistive flux expenditure during startup. Prior to current initiation, 1 to 2 MW of electron cyclotron resonance heating power at approx.90 GHz is used to create a small volume of high conductivity plasma (T/sub e/ approx. = 100 eV, n/sub e/ approx. = 10/sup 19/m/sup -3/) near the upper hybrid resonance (UHR) region. This plasma conditioning permits a small radius (a/sup 0/ approx.< 0.4 m)more » current channel to be established with a relatively low initial loop voltage (approx.< 25 V as opposed to approx.100 V without RF assist). During the subsequent plasma expansion and current ramp phase, additional RF power is introduced to reduce volt-second consumption due to plasma resistance. To study the preheating phase, a near classical particle and energy transport model is developed to estimate the electron heating efficiency in a currentless toroidal plasma. The model assumes that preferential electron heating at the UHR leads to the formation of an ambipolar sheath potential between the neutral plasma and the conducting vacuum vessel and limiter.« less

Theory and observation of the onset of nonlinear structures due to eigenmode destabilization by fast ions in tokamaks

NASA Astrophysics Data System (ADS)

Duarte, V. N.; Berk, H. L.; Gorelenkov, N. N.; Heidbrink, W. W.; Kramer, G. J.; Nazikian, R.; Pace, D. C.; Podestà, M.; Van Zeeland, M. A.

2017-12-01

Alfvén waves can induce the ejection of fast ions in different forms in tokamaks. In order to develop predictive capabilities to anticipate the nature of fast ion transport, a methodology is proposed to differentiate the likelihood of energetic-particle-driven instabilities to produce frequency chirping or fixed-frequency oscillations. The proposed method employs numerically calculated eigenstructures and multiple resonance surfaces of a given mode in the presence of energetic ion drag and stochasticity (due to collisions and micro-turbulence). Toroidicity-induced, reversed-shear and beta-induced Alfvén-acoustic eigenmodes are used as examples. Waves measured in experiments are characterized, and compatibility is found between the proposed criterion predictions and the experimental observation or lack of observation of chirping behavior of Alfvénic modes in different tokamaks. It is found that the stochastic diffusion due to micro-turbulence can be the dominant energetic particle detuning mechanism near the resonances in many plasma experiments, and its strength is the key as to whether chirping solutions are likely to arise. The proposed criterion constitutes a useful predictive tool in assessing whether the nature of the transport for fast ion losses in fusion devices will be dominated by convective or diffusive processes.

Development of a Grid-Based Gyro-Kinetic Simulation Code

NASA Astrophysics Data System (ADS)

Lapillonne, Xavier; Brunetti, Maura; Tran, Trach-Minh; Brunner, Stephan

2006-10-01

A grid-based semi-Lagrangian code using cubic spline interpolation is being developed at CRPP, for solving the electrostatic drift-kinetic equations [M. Brunetti et. al, Comp. Phys. Comm. 163, 1 (2004)] in a cylindrical system. This 4-dim code, CYGNE, is part of a project with long term aim of studying microturbulence in toroidal fusion devices, in the more general frame of gyro-kinetic equations. Towards their non-linear phase, the simulations from this code are subject to significant overshoot problems, reflected by the development of negative value regions of the distribution function, which leads to bad energy conservation. This has motivated the study of alternative schemes. On the one hand, new time integration algorithms are considered in the semi-Lagrangian frame. On the other hand, fully Eulerian schemes, which separate time and space discretisation (method of lines), are investigated. In particular, the Essentially Non Oscillatory (ENO) approach, constructed so as to minimize the overshoot problem, has been considered. All these methods have first been tested in the simpler case of the 2-dim guiding-center model for the Kelvin-Helmholtz instability, which enables to address the specific issue of the E xB drift also met in the more complex gyrokinetic-type equations. Based on these preliminary studies, the most promising methods are being implemented and tested in CYGNE.

Theory and observation of the onset of nonlinear structures due to eigenmode destabilization by fast ions in tokamaks

DOE Office of Scientific and Technical Information (OSTI.GOV)

Duarte, V. N.; Berk, H. L.; Gorelenkov, N. N.

Alfvén waves can induce the ejection of fast ions in different forms in tokamaks. In order to develop predictive capabilities to anticipate the nature of fast ion transport, a methodology is proposed to differentiate the likelihood of energetic-particle-driven instabilities to produce frequency chirping or fixed-frequency oscillations. Furthermore, the proposed method employs numerically calculated eigenstructures and multiple resonance surfaces of a given mode in the presence of energetic ion drag and stochasticity (due to collisions and micro-turbulence). Toroidicity-induced, reversed-shear and beta-induced Alfvén-acoustic eigenmodes are used as examples. Waves measured in experiments are characterized, and compatibility is found between the proposed criterionmore » predictions and the experimental observation or lack of observation of chirping behavior of Alfvénic modes in different tokamaks. It is found that the stochastic diffusion due to micro-turbulence can be the dominant energetic particle detuning mechanism near the resonances in many plasma experiments, and its strength is the key as to whether chirping solutions are likely to arise. We proposed a criterion that constitutes a useful predictive tool in assessing whether the nature of the transport for fast ion losses in fusion devices will be dominated by convective or diffusive processes.« less

Measured vs. Predicted Pedestal Pressure During RMP ELM Control in DIII-D

NASA Astrophysics Data System (ADS)

Zywicki, Bailey; Fenstermacher, Max; Groebner, Richard; Meneghini, Orso

2017-10-01

From database analysis of DIII-D plasmas with Resonant Magnetic Perturbations (RMPs) for ELM control, we will compare the experimental pedestal pressure (p_ped) to EPED code predictions and present the dependence of any p_ped differences from EPED on RMP parameters not included in the EPED model e.g. RMP field strength, toroidal and poloidal spectrum etc. The EPED code, based on Peeling-Ballooning and Kinetic Ballooning instability constraints, will also be used by ITER to predict the H-mode p_ped without RMPs. ITER plans to use RMPs as an effective ELM control method. The need to control ELMs in ITER is of the utmost priority, as it directly correlates to the lifetime of the plasma facing components. An accurate means of determining the impact of RMP ELM control on the p_ped is needed, because the device fusion power is strongly dependent on p_ped. With this new collection of data, we aim to provide guidance to predictions of the ITER pedestal during RMP ELM control that can be incorporated in a future predictive code. Work supported in part by US DoE under the Science Undergraduate Laboratory Internship (SULI) program and under DE-FC02-04ER54698, and DE-AC52-07NA27344.

Magnetic Fluctuation-Driven Intrinsic Flow in a Toroidal Plasma

NASA Astrophysics Data System (ADS)

Brower, D. L.; Ding, W. X.; Lin, L.; Almagri, A. F.; den Hartog, D. J.; Sarff, J. S.

2012-10-01

Magnetic fluctuations have been long observed in various magnetic confinement configurations. These perturbations may arise naturally from plasma instabilities such as tearing modes and energetic particle driven modes, but they can also be externally imposed by error fields or external magnetic coils. It is commonly observed that large MHD modes lead to plasma locking (no rotation) due to torque produced by eddy currents on the wall, and it is predicted that stochastic field induces flow damping where the radial electric field is reduced. Flow generation is of great importance to fusion plasma research, especially low-torque devices like ITER, as it can act to improve performance. Here we describe new measurements in the MST reversed field pinch (RFP) showing that the coherent interaction of magnetic and particle density fluctuations can produce a turbulent fluctuation-induced kinetic force, which acts to drive intrinsic plasma rotation. Key observations include; (1) the average kinetic force resulting from density fluctuations, ˜ 0.5 N/m^3, is comparable to the intrinsic flow acceleration, and (2) between sawtooth crashes, the spatial distribution of the kinetic force is directed to create a sheared parallel flow profile that is consistent with the measured flow profile in direction and amplitude, suggesting the kinetic force is responsible for intrinsic plasma rotation.

Poloidal structure of the plasma response to n = 1 Resonant Magnetic Perturbations in ASDEX Upgrade

NASA Astrophysics Data System (ADS)

Marrelli, L.; Bettini, P.; Piovesan, P.; Terranova, D.; Giannone, L.; Igochine, V.; Maraschek, M.; Suttrop, W.; Teschke, M.; Liu, Y. Q.; Ryan, D.; Eurofusion Mst1 Team; ASDEX Upgrade Team

2017-10-01

The hybrid scenario, a candidate for high-beta steady-state tokamak operations, becomes highly sensitive to 3D magnetic field near the no-wall limit. A predictive understanding of the plasma response to 3D fields near ideal MHD limits in terms of validated MHD stability codes is therefore important in order to safely operate future devices. Slowly rotating (5 - 10 Hz) n = 1 external magnetic fields have been applied in hybrid discharges in ASDEX Upgrade for an experimental characterization: the global n = 1 kink response has been measured by means of SXR and complete poloidal arrays of bθ probes located at different toroidal angles and compared to predictions of MHD codes such as MARS-F and V3FIT-VMEC. A Least-Squares Spectral Analysis approach has been developed together with a Monte Carlo technique to extract the small plasma response and its confidence interval from the noisy magnetic signals. MARS-F correctly reproduces the poloidal structure of the n = 1 measurements: for example, the dependence of the dominant poloidal mode number at the plasma edge from q95 is the same as in the experiment. Similar comparisons with V3FIT-VMEC and will be presented. See author list of ``H. Meyer et al. 2017 Nucl. Fusion 57 102014''.

Overview of HIT-SI Results and Plans

NASA Astrophysics Data System (ADS)

Ennis, D. A.; Akcay, C.; Hansen, C. J.; Hicks, N. K.; Hossack, A. C.; Jarboe, T. R.; Marklin, G. J.; Nelson, B. A.; Victor, B. S.

2011-10-01

Experiments in the Helicity Injected Torus-Steady Inductive (HIT-SI) device have achieved record spheromak current amplification during operations in deuterium plasmas. HIT-SI investigates steady inductive helicity injection with the aim of forming and sustaining a high-beta equilibrium in a spheromak geometry using two semi-toroidal injectors. Recent operations in deuterium plasmas have produced toroidal plasma currents greater than 50 kA, with current amplifications (Itor / Iinj) > 3 , and poloidal flux amplifications (ψpol /ψinj) > 10 . High performance deuterium discharges are achieved by initially conditioning the plasma-facing alumina surface of the HIT-SI confinement volume with helium plasmas. During subsequent deuterium operation the alumina surface strongly pumps deuterium, thereby limiting the density in the confinement volume. Additional measurements during high current deuterium discharges demonstrate reduced current and electron density fluctuations, impurity O III ion temperatures up to 50 eV and a toroidal current persistence for 0.6 ms after the injectors are shut off. Progress and plans for the HIT-SI3 configuration, with three injectors mounted on the same side of the confinement volume, will also be presented. Work supported by USDoE and ARRA.

The role of the density gradient on intermittent cross-field transport events in a simple magnetized toroidal plasma

NASA Astrophysics Data System (ADS)

Theiler, C.; Diallo, A.; Fasoli, A.; Furno, I.; Labit, B.; Podestà, M.; Poli, F. M.; Ricci, P.

2008-04-01

Intermittent cross-field particle transport events (ITEs) are studied in the basic toroidal device TORPEX [TORoidal Plasma EXperiment, A. Fasoli et al., Phys. Plasmas 13, 055902 (2006)], with focus on the role of the density gradient. ITEs are due to the intermittent radial elongation of an interchange mode. The elongating positive wave crests can break apart and form blobs. This is not necessary, however, for plasma particles to be convected a considerable distance across the magnetic field lines. Conditionally sampled data reveal two different scenarios leading to ITEs. In the first case, the interchange mode grows radially from a slab-like density profile and leads to the ITE. A novel analysis technique reveals a monotonic dependence between the vertically averaged inverse radial density scale length and the probability for a subsequent ITE. In the second case, the mode is already observed before the start of the ITE. It does not elongate radially in a first stage, but at a later time. It is shown that this elongation is preceded by a steepening of the density profile as well.

Dual-function magnetic structure for toroidal plasma devices

DOEpatents

Brown, Robert L.

1978-01-01

This invention relates to a support system wherein the iron core and yoke of the plasma current system of a tokamak plasma containment device is redesigned to support the forces of the magnet coils. The containment rings, which occupy very valuable space around the magnet coils, are utilized to serve as yokes for the core such that the conventional yoke is eliminated. The overall result is an improved aspect ratio, reduction in structure, smaller overall size, and improved access to the plasma ring.

Monte Carlo simulation of ion-material interactions in nuclear fusion devices

NASA Astrophysics Data System (ADS)

Nieto Perez, M.; Avalos-Zuñiga, R.; Ramos, G.

2017-06-01

One of the key aspects regarding the technological development of nuclear fusion reactors is the understanding of the interaction between high-energy ions coming from the confined plasma and the materials that the plasma-facing components are made of. Among the multiple issues important to plasma-wall interactions in fusion devices, physical erosion and composition changes induced by energetic particle bombardment are considered critical due to possible material flaking, changes to surface roughness, impurity transport and the alteration of physicochemical properties of the near surface region due to phenomena such as redeposition or implantation. A Monte Carlo code named MATILDA (Modeling of Atomic Transport in Layered Dynamic Arrays) has been developed over the years to study phenomena related to ion beam bombardment such as erosion rate, composition changes, interphase mixing and material redeposition, which are relevant issues to plasma-aided manufacturing of microelectronics, components on object exposed to intense solar wind, fusion reactor technology and other important industrial fields. In the present work, the code is applied to study three cases of plasma material interactions relevant to fusion devices in order to highlight the code's capabilities: (1) the Be redeposition process on the ITER divertor, (2) physical erosion enhancement in castellated surfaces and (3) damage to multilayer mirrors used on EUV diagnostics in fusion devices due to particle bombardment.

Ohmic ignition with high engineering beta based on the RFP

NASA Astrophysics Data System (ADS)

Sarff, J. S.; Anderson, J. K.; Chapman, B. E.; McCollam, K. J.

2017-10-01

The RFP configuration allows the possibility of ohmic ignition for fusion energy, eliminating the need for auxiliary heating by rf or neutral beam injection. Complex plasma-facing antennas and NBI sources are therefore not required, simplifying the difficult fusion materials challenge. While all toroidal configurations require a volume-average 〈 B 〉 >= 5 T, the field strength at the magnet in the RFP is only Bcoil 3T since plasma current generates almost all of the field. Engineering beta is therefore maximized. We summarize access to ohmic ignition by examining a Lawson-like power balance for an RFP fusion plasma comparable to the ARIES-AT advanced tokamak, which generates neutron wall loading Pn / A 5 MW/m2. The required energy confinement for ohmic ignition in an RFP is similar to that for a tokamak. Confinement in MST is comparable to a same-size, same-field tokamak plasma, but 〈 B 〉 in MST is only 1/20th that required for fusion. While transport could ultimately be dominated by micro turbulence, extrapolation of stochastic transport using Lundquist number scaling for MHD tearing indicates standard RFP confinement (not enhanced by current profile control) could be sufficient to access ohmic ignition. This bolsters the possibility for steady-state inductive sustainment using oscillating field current drive. The high beta and classical energetic ion confinement measured in MST also bolster the RFP's fusion potential. Work supported by U.S. DoE.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dahlburg, Jill; Corones, James; Batchelor, Donald

Fusion is potentially an inexhaustible energy source whose exploitation requires a basic understanding of high-temperature plasmas. The development of a science-based predictive capability for fusion-relevant plasmas is a challenge central to fusion energy science, in which numerical modeling has played a vital role for more than four decades. A combination of the very wide range in temporal and spatial scales, extreme anisotropy, the importance of geometric detail, and the requirement of causality which makes it impossible to parallelize over time, makes this problem one of the most challenging in computational physics. Sophisticated computational models are under development for many individualmore » features of magnetically confined plasmas and increases in the scope and reliability of feasible simulations have been enabled by increased scientific understanding and improvements in computer technology. However, full predictive modeling of fusion plasmas will require qualitative improvements and innovations to enable cross coupling of a wider variety of physical processes and to allow solution over a larger range of space and time scales. The exponential growth of computer speed, coupled with the high cost of large-scale experimental facilities, makes an integrated fusion simulation initiative a timely and cost-effective opportunity. Worldwide progress in laboratory fusion experiments provides the basis for a recent FESAC recommendation to proceed with a burning plasma experiment (see FESAC Review of Burning Plasma Physics Report, September 2001). Such an experiment, at the frontier of the physics of complex systems, would be a huge step in establishing the potential of magnetic fusion energy to contribute to the world’s energy security. An integrated simulation capability would dramatically enhance the utilization of such a facility and lead to optimization of toroidal fusion plasmas in general. This science-based predictive capability, which was cited in the FESAC integrated planning document (IPPA, 2000), represents a significant opportunity for the DOE Office of Science to further the understanding of fusion plasmas to a level unparalleled worldwide.« less

Proposal for a novel type of small scale aneutronic fusion reactor

NASA Astrophysics Data System (ADS)

Gruenwald, J.

2017-02-01

The aim of this work is to propose a novel scheme for a small scale aneutronic fusion reactor. This new reactor type makes use of the advantages of combining laser driven plasma acceleration and electrostatic confinement fusion. An intense laser beam is used to create a lithium-proton plasma with high density, which is then collimated and focused into the centre of the fusion reaction chamber. The basic concept presented here is based on the 7Li-proton fusion reaction. However, the physical and technological fundamentals may generally as well be applied to 11B-proton fusion. The former fusion reaction path offers higher energy yields while the latter has larger fusion cross sections. Within this paper a technological realisation of such a fusion device, which allows a steady state operation with highly energetic, well collimated ion beam, is presented. It will be demonstrated that the energetic break even can be reached with this device by using a combination of already existing technologies.

Development of a magnetized coaxial plasma gun for compact toroid injection into the C-2 field-reversed configuration device.

PubMed

Matsumoto, T; Sekiguchi, J; Asai, T; Gota, H; Garate, E; Allfrey, I; Valentine, T; Morehouse, M; Roche, T; Kinley, J; Aefsky, S; Cordero, M; Waggoner, W; Binderbauer, M; Tajima, T

2016-05-01

A compact toroid (CT) injector was developed for the C-2 device, primarily for refueling of field-reversed configurations. The CTs are formed by a magnetized coaxial plasma gun (MCPG), which consists of coaxial cylindrical electrodes and a bias coil for creating a magnetic field. First, a plasma ring is generated by a discharge between the electrodes and is accelerated by Lorenz self-force. Then, the plasma ring is captured by an interlinkage flux (poloidal flux). Finally, the fully formed CT is ejected from the MCPG. The MCPG described herein has two gas injection ports that are arranged tangentially on the outer electrode. A tungsten-coated inner electrode has a head which can be replaced with a longer one to extend the length of the acceleration region for the CT. The developed MCPG has achieved supersonic CT velocities of ∼100 km/s. Plasma parameters for electron density, electron temperature, and the number of particles are ∼5 × 10(21) m(-3), ∼40 eV, and 0.5-1.0 × 10(19), respectively.

Electron Temperature Fluctuation Measurements and Transport Model Validation at Alcator C-Mod

DOE Office of Scientific and Technical Information (OSTI.GOV)

White, Anne

The tokamak is a type of toroidal device used to confine a fusion plasma using large magnetic fields. Tokamaks and stellarators the leading devices for confining plasmas for fusion, and the capability to predict performance in these magnetically confined plasmas is essential for developing a sustainable fusion energy source. The magnetic configuration of tokamaks and stellarators does not exist in Nature, yet, the fundamental processes governing transport in fusion plasmas are universal – turbulence and instabilities, driven by inhomogeneity and asymmetry in the plasma, conspire to transport heat and particles across magnetic field lines and can play critical roles inmore » impurity confinement and generation of intrinsic rotation. Turbulence exists in all plasmas, and in neutral fluids as well. The study of turbulence is essential to developing a fundamental understanding of the nature of the fourth state of matter, plasmas. Experimental studies of turbulence in tokamaks date back to early scattering observations from the late 1970s. Since that time, great advances in turbulence diagnostics have been made, all of which have significantly enhanced our knowledge and understanding of turbulence in tokamaks. Through comparisons with advanced gyrokinetic theory and turbulent-transport models a great deal of evidence exists to implicate turbulent-driven transport as an important mechanism determining transport in all channels: heat, particle and momentum However, prediction and control of turbulent-driven transport remains elusive. Key to development of predictive transport models for magnetically confined fusion plasmas is validation of the nonlinear gyrokinetic transport model, which describes transport due to turbulence. Validation of gyrokinetic codes must include detailed and quantitative comparisons with measured turbulence characteristics, in addition to comparisons with inferred transport levels and equilibrium profiles. For this reason, advanced plasma diagnostics for studying core turbulence are needed in order to assess the accuracy of gyrokinetic models for turbulent-driven particle, heat and momentum transport. New core turbulence diagnostics at the world-class tokamaks Alcator C-Mod at MIT and ASDEX Upgrade at the Max Planck Institute for Plasma Physics have been designed, developed, and operated over the course of this project. These new instruments are capable of measuring electron temperature fluctuations and the phase angle between density and temperature fluctuations locally and quantitatively. These new data sets from Alcator C-Mod and ASDEX Upgrade are being used to fill key gaps in our understanding of turbulent transport in tokamaks. In particular, this project has results in new results on the topics of the Transport Shortfall, the role of ETG turbulence in tokamak plasmas, profile stiffness, the LOC/SOC transition, and intrinsic rotation reversals. These data are used in a rigorous process of “Transport model validation”, and this group is a world-leader on using turbulence models to design new hardware and new experiments at tokamaks. A correlation electron cyclotron emission (CECE) diagnostic is an instrument used to measure micro-scale fluctuations (mm-scale, compared to the machine size of meters) of electron temperature in magnetically confined fusion plasmas, such as those in tokamaks and stellarators. These micro-scale fluctuations are associated with drift-wave type turbulence, which leads to enhanced cooling and mixing of particles in fusion plasmas and limits achieving the required temperatures and densities for self-sustained fusion reactions. A CECE system can also be coupled with a reflectometer system that measured micro-scale density fluctuations, and from these simultaneous measurements, one can extract the phase between the density (n) and temperature (T) fluctuations, creating an nT phase diagnostic. Measurements of the fluctuations and the phase angle between them are extremely useful for testing and validating predictive models for the transport of heat and particles in fusion plasmas due to turbulence. Once validated, the models are used to predict performance in ITER and other burning plasmas, such as the MIT ARC design. Most recently, data from the newly developed, so-called “CECE diagnostic” [Cima 1995, White 2008] and “nT phase angle measurements” [Haese 1999, White 2010] ]will be combined with data from density fluctuation diagnostics at ASDEX Upgrade to support a long-term program of physics research in turbulence and transport that will allow for more stringent testing and validation of gyrokinetic turbulent-transport codes. This work directly impacts the development of predictive transport models in the U.S. FES program, such as TGLF, developed by General Atomics, which are used to predict performance in ITER and other burning plasma devices as part of advancing the development of fusion energy sciences.« less

Impaction durability of porous polyether-ether-ketone (PEEK) and titanium-coated PEEK interbody fusion devices.

PubMed

Torstrick, F Brennan; Klosterhoff, Brett S; Westerlund, L Erik; Foley, Kevin T; Gochuico, Joanna; Lee, Christopher S D; Gall, Ken; Safranski, David L

2018-05-01

Various surface modifications, often incorporating roughened or porous surfaces, have recently been introduced to enhance osseointegration of interbody fusion devices. However, these topographical features can be vulnerable to damage during clinical impaction. Despite the potential negative impact of surface damage on clinical outcomes, current testing standards do not replicate clinically relevant impaction loading conditions. The purpose of this study was to compare the impaction durability of conventional smooth polyether-ether-ketone (PEEK) cervical interbody fusion devices with two surface-modified PEEK devices that feature either a porous structure or plasma-sprayed titanium coating. A recently developed biomechanical test method was adapted to simulate clinically relevant impaction loading conditions during cervical interbody fusion procedures. Three cervical interbody fusion devices were used in this study: smooth PEEK, plasma-sprayed titanium-coated PEEK, and porous PEEK (n=6). Following Kienle et al., devices were impacted between two polyurethane blocks mimicking vertebral bodies under a constant 200 N preload. The posterior tip of the device was placed at the entrance between the polyurethane blocks, and a guided 1-lb weight was impacted upon the anterior face with a maximum speed of 2.6 m/s to represent the strike force of a surgical mallet. Impacts were repeated until the device was fully impacted. Porous PEEK durability was assessed using micro-computed tomography (µCT) pre- and postimpaction. Titanium-coating coverage pre- and postimpaction was assessed using scanning electron microscopy (SEM) and energy dispersive X-ray spectroscopy. Changes to the surface roughness of smooth and titanium-coated devices were also evaluated. Porous PEEK and smooth PEEK devices showed minimal macroscopic signs of surface damage, whereas the titanium-coated devices exhibited substantial visible coating loss. Quantification of the porous PEEK deformation demonstrated that the porous structure maintained a high porosity (>65%) following impaction that would be available for bone ingrowth, and exhibited minimal changes to pore size and depth. SEM and energy dispersive X-ray spectroscopy analysis of titanium-coated devices demonstrated substantial titanium coating loss after impaction that was corroborated with a decrease in surface roughness. Smooth PEEK showed minimal signs of damage using SEM, but demonstrated a decrease in surface roughness. Although recent surface modifications to interbody fusion devices are beneficial for osseointegration, they may be susceptible to damage and wear during impaction. The current study found porous PEEK devices to show minimal damage during simulated cervical impaction, whereas titanium-coated PEEK devices lost substantial titanium coverage. Copyright © 2018 The Author(s). Published by Elsevier Inc. All rights reserved.

ARC: A compact, high-field, disassemblable fusion nuclear science facility and demonstration power plant

NASA Astrophysics Data System (ADS)

Sorbom, Brandon; Ball, Justin; Palmer, Timothy; Mangiarotti, Franco; Sierchio, Jennifer; Bonoli, Paul; Kasten, Cale; Sutherland, Derek; Barnard, Harold; Haakonsen, Christian; Goh, Jon; Sung, Choongki; Whyte, Dennis

2014-10-01

The Affordable, Robust, Compact (ARC) reactor conceptual design aims to reduce the size, cost, and complexity of a combined Fusion Nuclear Science Facility (FNSF) and demonstration fusion pilot power plant. ARC is a 270 MWe tokamak reactor with a major radius of 3.3 m, a minor radius of 1.1 m, and an on-axis magnetic field of 9.2 T. ARC has Rare Earth Barium Copper Oxide (REBCO) superconducting toroidal field coils with joints to allow disassembly, allowing for removal and replacement of the vacuum vessel as a single component. Inboard-launched current drive of 25 MW LHRF power and 13.6 MW ICRF power is used to provide a robust, steady state core plasma far from disruptive limits. ARC uses an all-liquid blanket, consisting of low pressure, slowly flowing Fluorine Lithium Beryllium (FLiBe) molten salt. The liquid blanket acts as a working fluid, coolant, and tritium breeder, and minimizes the solid material that can become activated. The large temperature range over which FLiBe is liquid permits blanket operation at 800-900 K with single phase fluid cooling and allows use of a high-efficiency Brayton cycle for electricity production in the secondary coolant loop.

Fusion product losses due to fishbone instabilities in deuterium JET plasmas

NASA Astrophysics Data System (ADS)

Kiptily, V. G.; Fitzgerald, M.; Goloborodko, V.; Sharapov, S. E.; Challis, C. D.; Frigione, D.; Graves, J.; Mantsinen, M. J.; Beaumont, P.; Garcia-Munoz, M.; Perez von Thun, C.; Rodriguez, J. F. R.; Darrow, D.; Keeling, D.; King, D.; McClements, K. G.; Solano, E. R.; Schmuck, S.; Sips, G.; Szepesi, G.; Contributors, JET

2018-01-01

During development of a high-performance hybrid scenario for future deuterium-tritium experiments on the Joint European Torus, an increased level of fast ion losses in the MeV energy range was observed during the instability of high-frequency n  =  1 fishbones. The fishbones are excited during deuterium neutral beam injection combined with ion cyclotron heating. The frequency range of the fishbones, 10-25 kHz, indicates that they are driven by a resonant interaction with the NBI-produced deuterium beam ions in the energy range  ⩽120 keV. The fast particle losses in a much higher energy range are measured with a fast ion loss detector, and the data show an expulsion of deuterium plasma fusion products, 1 MeV tritons and 3 MeV protons, during the fishbone bursts. An MHD mode analysis with the MISHKA code combined with the nonlinear wave-particle interaction code HAGIS shows that the loss of toroidal symmetry caused by the n  =  1 fishbones affects strongly the confinement of non-resonant high energy fusion-born tritons and protons by perturbing their orbits and expelling them. This modelling is in a good agreement with the experimental data.

Te and H_α Measurements on the HT-7 Tokamak: A Collaboration between the Fusion Research Center and the Institute of Plasma Physics

NASA Astrophysics Data System (ADS)

Winslow, D. L.; Carter, K. R.; Chatterjee, R.; Huang, H.; Phillips, P. E.; Rowan, W. L.; Kuang, G. L.; Li, J. G.; Luo, J. R.; Wan, B. N.; Wan, Y. X.; Xie, J. K.

1998-11-01

A team from the Fusion Research Center at the University of Texas at Austin visited the HT-7 Tokamak at the Institute of Plasma Physics at the Chinese Academy of Sciences in Hefei, Anhui, China to study the effects of lower hybrid current drive (LHCD) in the HT-7 plasma. HT-7(HT-7 Group, Fusion Energy 1996 Vol. 1, 685 (1997).) is a medium-sized (R = 1.22 m) tokamak with superconducting toroidal field coils and long--pulse capabilities utilizing LHCD to assist ohmic current drive. Core and edge diagnostics supported by a stand-alone data acquisition system were installed for the spring 1998 campaign. The diagnostics included an ECE radiometer which allows determination of both electron temperature profiles and fluctuation levels in the core plasma and an H_α array detector for measurement of turbulence in regions not easily accessible to probes. In addition, a reciprocating Langmuir probe system was developed for use on HT-7 and should be available for the next campaign. The effects of LHCD upon fluctuation levels in the plasma will be discussed.

Territorial characteristics of low frequency electrostatic fluctuations in a simple magnetized torus

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kaur, R.; Singh, R.; Sarada Sree, A.

2011-01-15

This paper presents an experimental investigation of turbulence in simple toroidal plasma devices without rotational transform. It is argued that Rayleigh-Taylor (flute interchange) mode may be one of the source mechanisms for the observed turbulence but is not sufficient to explain its observed global characteristics. Taking BETA device as an example, we show that pure Rayleigh-Taylor mode cannot explain (i) the observation of mode maximum at the location other than where density scale length is minimum, (ii) the comparable value of amplitude level of fluctuations in good curvature region, and (iii) the decrease in the mode amplitude with increasing magneticmore » field. Investigations have revealed that there exists not only poloidal plasma flow but also that it is sheared. Including this effect explains the first observation. However, modification brought about by velocity shear in the Rayleigh-Taylor mode still does not explain our second and third observations. We have taken an approach that since Rayleigh-Taylor is not excited in a good curvature region, it cannot be the source of turbulence there. Nor is it defensible to say that turbulence born in a bad curvature region is carried over through ExB rotation to the good curvature region. Consequently, we have invoked cross-field Simon-Hoh instability for this region. Experimental evidence supporting our proposal is presented. This paper concludes that toroidal devices have simultaneous existence of different self-consistent sources of turbulence in different regions of the device.« less

A study to compare the efficacy of polyether ether ketone rod device with titanium devices in posterior spinal fusion in a canine model.

PubMed

Wang, Nanxiang; Xie, Huanxin; Xi, Chunyang; Zhang, Han; Yan, Jinglong

2017-03-09

The benefits of posterior lumbar fusion surgery with orthotopic paraspinal muscle-pediculated bone flaps are well established. However, the problem of non-union due to mechanical support is not completely resolved. The aim of the study was to compare the efficacy of polyether ether ketone (PEEK) rod device with conventional titanium devices in the posterior lumbar fusion surgery with orthotopic paraspinal muscle-pediculated bone flaps. This was a randomized controlled study with an experimental animal model. Thirty-two mongrel dogs were randomly divided into two groups-control group (n = 16), which received the titanium device and the treatment group (n = 16), which received PEEK rods. The animals were sacrificed 8 or 16 weeks after surgery. Lumbar spines of dogs in both groups were removed, harvested, and assessed for radiographic, biomechanical, and histological changes. Results in the current study indicated that there was no significant difference in the lumbar spine of the control and treatment groups in terms of radiographic, manual palpation, and gross examination. However, certain parameters of biomechanical testing showed significant differences (p < 0.05) in stiffness and displacement, revealing a better fusion (treatment group showed decreased stiffness with decreased displacement) of the bone graft. Similarly, the histological analysis also revealed a significant fusion mass in both treatment and control groups (p < 0.05). These findings revealed that fixation using PEEK connecting rod could improve the union of the bone graft in the posterior lumbar spine fusion surgery compared with that of the titanium rod fixation.

One and two level posterior lumbar interbody fusion (PLIF) using an expandable, stand-alone, interbody fusion device: a VariLift® case series

PubMed Central

Barrett-Tuck, Rebecca; Del Monaco, Diana

2017-01-01

Background Surgical interventions such as posterior lumbar interbody fusion (PLIF) with and without posterior instrumentation are often employed in patients with degenerative spinal conditions that fail to respond to conservative medical management. The VariLift® Interbody Fusion System was developed as a stand-alone solution to provide the benefits of an intervertebral fusion device without the requirement of supplemental pedicle screw fixation. Methods In this retrospective case series, 25 patients underwent PLIF with a stand-alone VariLift® expandable interbody fusion device without adjunctive pedicle screw fixation. There were 12 men and 13 women, with a mean age of 57.2 years (range, 33–83 years); single level in 18 patients, 2 levels in 7 patients. Back pain severity was reported as none, mild, moderate, severe and worst imaginable at baseline, 6 and 12 months. Preoperatively, 88% (22 of 25) of patients reported severe back pain. Results All patients experienced symptomatic improvement and, by 12 months postoperatively, 71% (15 of 21) of patients reported only mild residual pain. Overall, pain scores improved significantly from baseline to 12 months (P=0.0002). There were no revision surgeries and fusion was achieved 12 of 13 patients (92%) who returned for a 12-month radiographic follow-up. There were three cases of intractable postsurgical pain which required extended hospitalization or pain management, one wound infection and one case of surgical site dehiscence, both treated and resolved during inpatient hospitalization. Conclusions In this single-physician case series, the VariLift® device used in single or two-level PLIF provided effective symptom relief and produced a high fusion rate without the need for supplemental fixation. PMID:28435912

One and two level posterior lumbar interbody fusion (PLIF) using an expandable, stand-alone, interbody fusion device: a VariLift® case series.

PubMed

Barrett-Tuck, Rebecca; Del Monaco, Diana; Block, Jon E

2017-03-01

Surgical interventions such as posterior lumbar interbody fusion (PLIF) with and without posterior instrumentation are often employed in patients with degenerative spinal conditions that fail to respond to conservative medical management. The VariLift ® Interbody Fusion System was developed as a stand-alone solution to provide the benefits of an intervertebral fusion device without the requirement of supplemental pedicle screw fixation. In this retrospective case series, 25 patients underwent PLIF with a stand-alone VariLift ® expandable interbody fusion device without adjunctive pedicle screw fixation. There were 12 men and 13 women, with a mean age of 57.2 years (range, 33-83 years); single level in 18 patients, 2 levels in 7 patients. Back pain severity was reported as none, mild, moderate, severe and worst imaginable at baseline, 6 and 12 months. Preoperatively, 88% (22 of 25) of patients reported severe back pain. All patients experienced symptomatic improvement and, by 12 months postoperatively, 71% (15 of 21) of patients reported only mild residual pain. Overall, pain scores improved significantly from baseline to 12 months (P=0.0002). There were no revision surgeries and fusion was achieved 12 of 13 patients (92%) who returned for a 12-month radiographic follow-up. There were three cases of intractable postsurgical pain which required extended hospitalization or pain management, one wound infection and one case of surgical site dehiscence, both treated and resolved during inpatient hospitalization. In this single-physician case series, the VariLift ® device used in single or two-level PLIF provided effective symptom relief and produced a high fusion rate without the need for supplemental fixation.

A Standard Mammography Unit - Standard 3D Ultrasound Probe Fusion Prototype: First Results.

PubMed

Schulz-Wendtland, Rüdiger; Jud, Sebastian M; Fasching, Peter A; Hartmann, Arndt; Radicke, Marcus; Rauh, Claudia; Uder, Michael; Wunderle, Marius; Gass, Paul; Langemann, Hanna; Beckmann, Matthias W; Emons, Julius

2017-06-01

The combination of different imaging modalities through the use of fusion devices promises significant diagnostic improvement for breast pathology. The aim of this study was to evaluate image quality and clinical feasibility of a prototype fusion device (fusion prototype) constructed from a standard tomosynthesis mammography unit and a standard 3D ultrasound probe using a new method of breast compression. Imaging was performed on 5 mastectomy specimens from patients with confirmed DCIS or invasive carcinoma (BI-RADS ™ 6). For the preclinical fusion prototype an ABVS system ultrasound probe from an Acuson S2000 was integrated into a MAMMOMAT Inspiration (both Siemens Healthcare Ltd) and, with the aid of a newly developed compression plate, digital mammogram and automated 3D ultrasound images were obtained. The quality of digital mammogram images produced by the fusion prototype was comparable to those produced using conventional compression. The newly developed compression plate did not influence the applied x-ray dose. The method was not more labour intensive or time-consuming than conventional mammography. From the technical perspective, fusion of the two modalities was achievable. In this study, using only a few mastectomy specimens, the fusion of an automated 3D ultrasound machine with a standard mammography unit delivered images of comparable quality to conventional mammography. The device allows simultaneous ultrasound - the second important imaging modality in complementary breast diagnostics - without increasing examination time or requiring additional staff.

Nuclear Fusion Award 2010 speech Nuclear Fusion Award 2010 speech

NASA Astrophysics Data System (ADS)

Rice, John

2011-01-01

Following the suggestion of Earl Marmar in 1995, I installed a compact von Hamos type x-ray spectrometer (originally built with Elisabeth Rachlew and Jan Kallne) on a tangentially viewing port on the Alcator C-Mod tokamak. The spectrometer views the plasma through a 2 cm diameter hole, and is tuned to H-like argon, suitable for passive measurement of the core toroidal rotation velocity from the Doppler shift. It soon became evident that the rotation in Ohmic L-mode discharges, while for the most part directed counter-current, depends in a very complicated fashion on plasma parameters, notably the electron density, current and magnetic configuration. The rotation can even flip sign for almost no apparent reason! In Ohmic and ion cyclotron range of frequencies (ICRF) heated H-mode plasmas the rotation is in the co-current direction and has a relatively simple dependence on plasma parameters, proportional to the stored energy normalized to the current. Rotation velocities as high as 130 km s-1 have been observed without external momentum input. In dimensionless terms this intrinsic (or spontaneous rotation) depends on the normalized plasma pressure. The association of toroidal rotation with plasma pressure in ICRF H-modes was first observed by Lars-Goran Eriksson in JET discharges. Similar results were subsequently reported for Tore Supra enhanced confinement plasmas. In the early 2000s concerns began to surface about the lack of substantial neutral beam driven rotation in ITER, and intrinsic rotation became a topic of interest in the ITPA Transport Group. Through that connection, similar observations from DIII-D, TCV and JT-60U were added to the growing list. A database of intrinsic rotation observations was assembled with the goal of extrapolating to the expected values for ITER. Both dimensional and dimensionless scalings were developed and formed the backbone of the 2007 Nuclear Fusion paper. I gratefully acknowledge the important contributions to this paper from Alex Ince-Cushman, John deGrassie, Lars-Goran Eriksson, Yoshiteru Sakamoto, Andrea Scarabosio and Yuri Podpaly, as well as the other coauthors. I would like to express my sincere appreciation to Earl Marmar, Martin Greenwald and Miklos Porkolab at MIT for continued support of this work, as well as to the entire C-Mod team. This award was made possible due to the insight of Mitsuru Kikuchi and the support of the IAEA through Werner Burkhart, and I am truly grateful to both of them. Many thanks as well to the outstanding staff at Nuclear Fusion. It is a distinct honor to be included in the group of previous winners: Tim Luce, Clemente Angioni, Todd Evans and Steve Sabbagh. It is also a great honor to be considered alongside the 2010 nominees: Phil Snyder, Sibylle Guenter, Maiko Yoshida, Hajime Urano, Fulvio Zonca, Erik Garcia, Costanza Maggi, Hartmut Zohm, Thierry Loarer and Bruce Lipschultz. Finally, I would like to thank the readers of Nuclear Fusion for the many citations. John Rice 2010 Nuclear Fusion Award winner Plasma Science and Fusion Center, MIT, Cambridge, MA, USA

Engineering Considerations for the Self-Energizing Magnetoplasmadynamic (MPD)-Type Fusion Plasma Thruster

DTIC Science & Technology

1992-02-01

Feasibility studies Of dense plasma focus (DPF) device as a fusion propulsion thruster have been performed. Both conventional and spin-polarized D...uncertainties remain in the validity of scaling laws on capacitor mass at high current beyond 1 MA. Fusion Propulsion, Dense Plasma Focus , Magnetoplasmadynamic Thruster, Advanced Fuel, D-3He Fusion, Spin-Polarized Fusion.

Toroidal ferroelectricity in PbTiO3 nanoparticles.

PubMed

Stachiotti, M G; Sepliarsky, M

2011-04-01

We report from first-principles-based atomistic simulations that ferroelectricity can be sustained in PbTiO(3) nanoparticles of only a few lattice constants in size as a result of a toroidal ordering. We find that size-induced topological transformations lead to the stabilization of a ferroelectric bubble by the alignment of vortex cores along a closed path. These transformations, which are driven by the aspect ratio of the nanostructure, change the topology of the polarization field, producing a rich variety of polar configurations. For sufficiently flat nanostructures, a multibubble state bridges the gap between 0D nanodots and 2D ultrathin films. The thermal properties of the ferroelectric bubbles indicate that this state is suitable for the development of nanometric devices. © 2011 American Physical Society

Current Status of the Gasdynamic Mirror Fusion Propulsion Experiment

NASA Technical Reports Server (NTRS)

Emrich, William J., Jr.

2002-01-01

Nuclear fusion appears to be the most promising concept for producing extremely high specific impulse rocket engines. One particular fusion concept which seems to be particularly well suited for fusion propulsion applications is the gasdynamic mirror (GDM). An experimental GDM device has been constructed at the NASA Marshall Space Flight Center to provide an initial assessment of the feasibility of this type of propulsion system. An initial shakedown of the device is currently underway with initial experiments slated to occur in late 2001. This device would operate at much higher plasma densities and with much larger L/D ratios than previous mirror machines. The high L/D ratio minimizes to a large extent certain magnetic curvature effects which lead to plasma instabilities causing a loss of plasma confinement. The high plasma density results in the plasma behaving much more like a conventional fluid with a mean free path shorter than the length of the device. This characteristic helps reduce problems associated with 'loss cone' microinstabilities. The device has been constructed to allow a considerable degree of flexibility in its configuration thus permitting the experiment to grow over time without necessitating a great deal of additional fabrication.

Self-consistent modeling of CFETR baseline scenarios for steady-state operation

NASA Astrophysics Data System (ADS)

Chen, Jiale; Jian, Xiang; Chan, Vincent S.; Li, Zeyu; Deng, Zhao; Li, Guoqiang; Guo, Wenfeng; Shi, Nan; Chen, Xi; CFETR Physics Team

2017-07-01

Integrated modeling for core plasma is performed to increase confidence in the proposed baseline scenario in the 0D analysis for the China Fusion Engineering Test Reactor (CFETR). The steady-state scenarios are obtained through the consistent iterative calculation of equilibrium, transport, auxiliary heating and current drives (H&CD). Three combinations of H&CD schemes (NB + EC, NB + EC + LH, and EC + LH) are used to sustain the scenarios with q min > 2 and fusion power of ˜70-150 MW. The predicted power is within the target range for CFETR Phase I, although the confinement based on physics models is lower than that assumed in 0D analysis. Ideal MHD stability analysis shows that the scenarios are stable against n = 1-10 ideal modes, where n is the toroidal mode number. Optimization of RF current drive for the RF-only scenario is also presented. The simulation workflow for core plasma in this work provides a solid basis for a more extensive research and development effort for the physics design of CFETR.

Kinetic Monte Carlo and cellular particle dynamics simulations of multicellular systems

NASA Astrophysics Data System (ADS)

Flenner, Elijah; Janosi, Lorant; Barz, Bogdan; Neagu, Adrian; Forgacs, Gabor; Kosztin, Ioan

2012-03-01

Computer modeling of multicellular systems has been a valuable tool for interpreting and guiding in vitro experiments relevant to embryonic morphogenesis, tumor growth, angiogenesis and, lately, structure formation following the printing of cell aggregates as bioink particles. Here we formulate two computer simulation methods: (1) a kinetic Monte Carlo (KMC) and (2) a cellular particle dynamics (CPD) method, which are capable of describing and predicting the shape evolution in time of three-dimensional multicellular systems during their biomechanical relaxation. Our work is motivated by the need of developing quantitative methods for optimizing postprinting structure formation in bioprinting-assisted tissue engineering. The KMC and CPD model parameters are determined and calibrated by using an original computational-theoretical-experimental framework applied to the fusion of two spherical cell aggregates. The two methods are used to predict the (1) formation of a toroidal structure through fusion of spherical aggregates and (2) cell sorting within an aggregate formed by two types of cells with different adhesivities.

Nuclear Fusion prize laudation Nuclear Fusion prize laudation

NASA Astrophysics Data System (ADS)

Burkart, W.

2011-01-01

Clean energy in abundance will be of critical importance to the pursuit of world peace and development. As part of the IAEA's activities to facilitate the dissemination of fusion related science and technology, the journal Nuclear Fusion is intended to contribute to the realization of such energy from fusion. In 2010, we celebrated the 50th anniversary of the IAEA journal. The excellence of research published in the journal is attested to by its high citation index. The IAEA recognizes excellence by means of an annual prize awarded to the authors of papers judged to have made the greatest impact. On the occasion of the 2010 IAEA Fusion Energy Conference in Daejeon, Republic of Korea at the welcome dinner hosted by the city of Daejeon, we celebrated the achievements of the 2009 and 2010 Nuclear Fusion prize winners. Steve Sabbagh, from the Department of Applied Physics and Applied Mathematics, Columbia University, New York is the winner of the 2009 award for his paper: 'Resistive wall stabilized operation in rotating high beta NSTX plasmas' [1]. This is a landmark paper which reports record parameters of beta in a large spherical torus plasma and presents a thorough investigation of the physics of resistive wall mode (RWM) instability. The paper makes a significant contribution to the critical topic of RWM stabilization. John Rice, from the Plasma Science and Fusion Center, MIT, Cambridge is the winner of the 2010 award for his paper: 'Inter-machine comparison of intrinsic toroidal rotation in tokamaks' [2]. The 2010 award is for a seminal paper that analyzes results across a range of machines in order to develop a universal scaling that can be used to predict intrinsic rotation. This paper has already triggered a wealth of experimental and theoretical work. I congratulate both authors and their colleagues on these exceptional papers. W. Burkart Deputy Director General Department of Nuclear Sciences and Applications International Atomic Energy Agency, Vienna, Austria References [1] Sabbagh S. et al 2006 Nucl. Fusion 46 635-44 [2] Rice J.E. et al 2007 Nucl. Fusion 47 1618-24

Suppression of energetic particle driven instabilities with HHFW heating

DOE PAGES

Fredrickson, E. D.; Taylor, G.; Bertelli, N.; ...

2015-01-01

In plasmas in the National Spherical Torus Experiment (NSTX) [Ono et al., Nucl. Fusion 40 (2000) 557] heated with neutral beams, the beam ions typically excite Energetic Particle Modes (EPMs or fishbones), and Toroidal, Global or Compressional Alfvén Eigenmodes (TAE, GAE, CAE). These modes can redistribute the energetic beam ions, altering the beam driven current profile and the plasma heating profile, or they may affect electron thermal transport or cause losses of the beam ions. In this paper we present experimental results where these instabilities, driven by the super-thermal beam ions, are suppressed with the application of High Harmonic Fastmore » Wave heating.« less

Compact toroid injection into C-2U

NASA Astrophysics Data System (ADS)

Roche, Thomas; Gota, H.; Garate, E.; Asai, T.; Matsumoto, T.; Sekiguchi, J.; Putvinski, S.; Allfrey, I.; Beall, M.; Cordero, M.; Granstedt, E.; Kinley, J.; Morehouse, M.; Sheftman, D.; Valentine, T.; Waggoner, W.; the TAE Team

2015-11-01

Sustainment of an advanced neutral beam-driven FRC for a period in excess of 5 ms is the primary goal of the C-2U machine at Tri Alpha Energy. In addition, a criteria for long-term global sustainment of any magnetically confined fusion reactor is particle refueling. To this end, a magnetized coaxial plasma-gun has been developed. Compact toroids (CT) are to be injected perpendicular to the axial magnetic field of C-2U. To simulate this environment, an experimental test-stand has been constructed. A transverse magnetic field of B ~ 1 kG is established (comparable to the C-2U axial field) and CTs are fired across it. As a minimal requirement, the CT must have energy density greater than that of the magnetic field it is to penetrate, i.e., 1/2 ρv2 >=B2 / 2μ0 . This criteria is easily met and indeed the CTs traverse the test-stand field. A preliminary experiment on C-2U shows the CT also capable of penetrating into FRC plasmas and refueling is observed resulting in a 20 - 30% increase in total particle number per single-pulsed CT injection. Results from test-stand and C-2U experiments will be presented.

Fully Parallel MHD Stability Analysis Tool

NASA Astrophysics Data System (ADS)

Svidzinski, Vladimir; Galkin, Sergei; Kim, Jin-Soo; Liu, Yueqiang

2015-11-01

Progress on full parallelization of the plasma stability code MARS will be reported. MARS calculates eigenmodes in 2D axisymmetric toroidal equilibria in MHD-kinetic plasma models. It is a powerful tool for studying MHD and MHD-kinetic instabilities and it is widely used by fusion community. Parallel version of MARS is intended for simulations on local parallel clusters. It will be an efficient tool for simulation of MHD instabilities with low, intermediate and high toroidal mode numbers within both fluid and kinetic plasma models, already implemented in MARS. Parallelization of the code includes parallelization of the construction of the matrix for the eigenvalue problem and parallelization of the inverse iterations algorithm, implemented in MARS for the solution of the formulated eigenvalue problem. Construction of the matrix is parallelized by distributing the load among processors assigned to different magnetic surfaces. Parallelization of the solution of the eigenvalue problem is made by repeating steps of the present MARS algorithm using parallel libraries and procedures. Results of MARS parallelization and of the development of a new fix boundary equilibrium code adapted for MARS input will be reported. Work is supported by the U.S. DOE SBIR program.

Theory comparison and numerical benchmarking on neoclassical toroidal viscosity torque

NASA Astrophysics Data System (ADS)

Wang, Zhirui; Park, Jong-Kyu; Liu, Yueqiang; Logan, Nikolas; Kim, Kimin; Menard, Jonathan E.

2014-04-01

Systematic comparison and numerical benchmarking have been successfully carried out among three different approaches of neoclassical toroidal viscosity (NTV) theory and the corresponding codes: IPEC-PENT is developed based on the combined NTV theory but without geometric simplifications [Park et al., Phys. Rev. Lett. 102, 065002 (2009)]; MARS-Q includes smoothly connected NTV formula [Shaing et al., Nucl. Fusion 50, 025022 (2010)] based on Shaing's analytic formulation in various collisionality regimes; MARS-K, originally computing the drift kinetic energy, is upgraded to compute the NTV torque based on the equivalence between drift kinetic energy and NTV torque [J.-K. Park, Phys. Plasma 18, 110702 (2011)]. The derivation and numerical results both indicate that the imaginary part of drift kinetic energy computed by MARS-K is equivalent to the NTV torque in IPEC-PENT. In the benchmark of precession resonance between MARS-Q and MARS-K/IPEC-PENT, the agreement and correlation between the connected NTV formula and the combined NTV theory in different collisionality regimes are shown for the first time. Additionally, both IPEC-PENT and MARS-K indicate the importance of the bounce harmonic resonance which can greatly enhance the NTV torque when E ×B drift frequency reaches the bounce resonance condition.

One-dimensional energetic particle quasilinear diffusion for realistic TAE instabilities

NASA Astrophysics Data System (ADS)

Duarte, Vinicius; Ghantous, Katy; Berk, Herbert; Gorelenkov, Nikolai

2014-10-01

Owing to the proximity of the characteristic phase (Alfvén) velocity and typical energetic particle (EP) superthermal velocities, toroidicity-induced Alfvén eigenmodes (TAEs) can be resonantly destabilized endangering the plasma performance. Thus, it is of ultimate importance to understand the deleterious effects on the confinement resulting from fast ion driven instabilities expected in fusion-grade plasmas. We propose to study the interaction of EPs and TAEs using a line broadened quasilinear model, which captures the interaction in both regimes of isolated and overlapping modes. The resonance particles diffuse in the phase space where the problem essentially reduces to one dimension with constant kinetic energy and the diffusion mainly along the canonical toroidal angular momentum. Mode structure and wave particle resonances are computed by the NOVA code and are used in a quasilinear diffusion code that is being written to study the evolution of the distribution function, under the assumption that they can be considered virtually unalterable during the diffusion. A new scheme for the resonant particle diffusion is being proposed that builds on the 1-D nature of the diffusion from a single mode, which leads to a momentum conserving difference scheme even when there is mode overlap.

Optimization of 3D Field Design

NASA Astrophysics Data System (ADS)

Logan, Nikolas; Zhu, Caoxiang

2017-10-01

Recent progress in 3D tokamak modeling is now leveraged to create a conceptual design of new external 3D field coils for the DIII-D tokamak. Using the IPEC dominant mode as a target spectrum, the Finding Optimized Coils Using Space-curves (FOCUS) code optimizes the currents and 3D geometry of multiple coils to maximize the total set's resonant coupling. The optimized coils are individually distorted in space, creating toroidal ``arrays'' containing a variety of shapes that often wrap around a significant poloidal extent of the machine. The generalized perturbed equilibrium code (GPEC) is used to determine optimally efficient spectra for driving total, core, and edge neoclassical toroidal viscosity (NTV) torque and these too provide targets for the optimization of 3D coil designs. These conceptual designs represent a fundamentally new approach to 3D coil design for tokamaks targeting desired plasma physics phenomena. Optimized coil sets based on plasma response theory will be relevant to designs for future reactors or on any active machine. External coils, in particular, must be optimized for reliable and efficient fusion reactor designs. Work supported by the US Department of Energy under DE-AC02-09CH11466.

Overview of MST Research

NASA Astrophysics Data System (ADS)

Sarff, J. S.; MST Team

2011-10-01

MST progress in advancing the RFP for (1) fusion plasma confinement with minimal external magnetization, (2) toroidal confinement physics, and (3) basic plasma physics is summarized. New tools and diagnostics are accessing physics barely studied in the RFP. Several diagnostic advances are important for ITER/burning plasma. A 1 MW neutral beam injector operates routinely for fast ion, heating, and transport investigations. Energetic ions are also created spontaneously by tearing mode reconnection, reminiscent of astrophysical plasmas. Classical confinement of impurity ions is measured in reduced-tearing plasmas. Fast ion slowing-down is also classical. Alfven-eigenmode-like activity occurs with NBI, but apparently not TAE. Stellarator-like helical structure appears in the core of high current plasmas, with improved confinement characteristics. FIR interferometry, Thomson scattering, and HIBP diagnostics are beginning to explore microturbulence scales, an opportunity to exploit the RFP's high beta and strong magnetic shear parameter space. A programmable power supply for the toroidal field flexibly explores scenarios from advanced inductive profile control to low current tokamak operation. A 1 MW 5.5 GHz source for electron Bernstein wave injection is nearly complete to investigate heating and current drive in over-dense plasmas. Supported by DOE and NSF.

Tridimensional Thermonuclear Instability in Subignited Plasmas and on the Surface of the Pulsars

NASA Astrophysics Data System (ADS)

Cardinali, A.; Coppi, B.

2016-10-01

Tridimensional modes involving an increase of the electron temperature can be excited as a result of α-particle heating in subignited D-T fusion burning plasmas when a nearly time- independent external source of heating is applied. The analyzed modes are shown to emerge from an axisymmetric toroidal configurations and are radially localized around rational magnetic surfaces corresponding to q(r =r0) =m0 /n0 where m0 and n0 are the relevant poloidal and toroidal mode numbers. The radial width of the mode is of the order of the thermal scale distances δT =D⊥e th /D∥e th 1/4 (R0 /n0) 1/2(dlnq/dr)0-1/2. The mode has a rather severe damping rate, that has to be overcome by the relevant heating rate. Thus the temperature range to be considered is that where the D-T plasma reactivity undergoes a relatively large increase as a function of temperature. This kind of theory has been applied to the plasmas that are envisioned to be associated with surface of pulsar and be subjects to (spatially) inhomogenous thermonuclear burning. Sponsored in part by the US DOE.

Transition From High Harmonic Fast Wave to Whistler/Helicon Regime in Tokamaks

NASA Astrophysics Data System (ADS)

Harris, S. P.; Pinsker, R. I.; Porkolab, M.

2014-10-01

Experiments are being prepared1 on DIII-D in which fast waves (FWs) at 0.5 GHz will be used to drive current noninductively in the mid-radius region. Previous DIII-D experiments used FWs at ~0.1 GHz to drive central current; in this work we examine the frequency dependence of wave propagation and damping in the 0.1-1.0 GHz range with the goal of identifying the optimum frequency range for a particular application. Strongly enhanced electron damping and reduced ion damping at higher frequencies must be weighed against increasing coupling difficulties at higher frequencies and more restrictive wave accessibility at low toroidal field. Wave propagation and accessibility is studied with ray tracing models in slab, cylindrical, and fully toroidal geometries. Analytic expressions for electron and ion damping will be derived with an emphasis on understanding the transition from the moderate-to-high ion cyclotron harmonic regime to the very high harmonic or ``whistler''/``helicon''/lower hybrid FW regime. Work supported in part by the National Undergraduate Fellowship Program in Plasma Physics and Fusion Energy Sciences and the US Department of Energy under DE-FC02-04ER54698.

Measurements of Impurity Particle Transport Associated with Drift-Wave Turbulence in MST

NASA Astrophysics Data System (ADS)

Nishizawa, Takashi; Nornberg, Mark; Boguski, John; Craig, Darren; den Hartog, Daniel; Pueschel, M. J.; Sarff, John; Terry, Paul; Williams, Zach; Xing, Zichuan

2017-10-01

Understanding and controlling impurity transport in a toroidal magnetized plasma is one of the critical issues that need to be addressed in order to achieve controlled fusion. Gyrokinetic modeling shows turbulence can drive impurity transport, but direct measurements of the turbulent flux have not been made. Particle balance is typically used to infer the presence of turbulent impurity transport. We report, for the first time in a toroidal plasma, direct measurements of turbulence-driven impurity transport. Trapped electron mode (TEM) turbulence appears in MST plasmas when MHD tearing fluctuations are suppressed. Impurity ion-Doppler spectroscopy is used to correlate impurity density and radial velocity fluctuations associated with TEM. Small Doppler shifts associated with the radial velocity fluctuations (rms 1km/s) are resolved with the use of a new linearized spectrum correlation analysis method, which improves the rejection of Poisson noise. The method employs frequency-domain correlation analysis to expose the fluctuation and transport spectrum. The C+ 2 impurity transport velocity driven by turbulence is found to be 48m/s (inward), which is sufficiently large to impact an impurity flux balance in MST improved-confinement plasmas. This work is supported by the US DOE.

Son of IXION: A Steady State Centrifugally Confined Plasma for Fusion*

NASA Astrophysics Data System (ADS)

Hassam, Adil

1996-11-01

A magnetic confinement scheme in which the inertial, u.grad(u), forces effect parallel confinement is proposed. The basic geometry is mirror-like as far as the poloidal field goes or, more simply, multipole (FM-1) type. The rotation is toroidal in this geometry. A supersonic rotation can effect complete parallel confinement, with the usual magnetic mirror force rendered irrelevant. The rotation shear, in addition, aids in the suppression of the flute mode. This suppression is not complete which indicates the addition of a toroidal field, at maximum of the order of the poloidal field. We show that at rotation in excess of Mach 3, the parallel particle and heat losses can be minimized to below the Lawson breakeven point. The crossfield transport can be expected to be better than tokamaks on account of the large velocity shear. Other advantages of the scheme are that it is steady state and disruption free. An exploratory experiment that tests equilibrium, parallel detachment, and MHD stability is proposed. The concept resembles earlier (Geneva, 1958) experiments on "homopolar generators" and a mirror configuration called IXION. Ixion, Greek mythological king, was forever strapped to a rotating, flaming wheel. *Work supported by DOE

Calculation of the non-inductive current profile in high-performance NSTX plasmas

NASA Astrophysics Data System (ADS)

Gerhardt, S. P.; Fredrickson, E.; Gates, D.; Kaye, S.; Menard, J.; Bell, M. G.; Bell, R. E.; Le Blanc, B. P.; Kugel, H.; Sabbagh, S. A.; Yuh, H.

2011-03-01

The constituents of the current profile have been computed for a wide range of high-performance plasmas in NSTX (Ono et al 2000 Nucl. Fusion 40 557); these include cases designed to maximize the non-inductive fraction, pulse length, toroidal-β or stored energy. In the absence of low-frequency MHD activity, good agreement is found between the reconstructed current profile and that predicted by summing the independently calculated inductive, pressure-driven and neutral beam currents, without the need to invoke any anomalous beam ion diffusion. Exceptions occur, for instance, when there are toroidal Alfvén eigenmode avalanches or coupled m/n = 1/1 + 2/1 kink-tearing modes. In these cases, the addition of a spatially and temporally dependent fast-ion diffusivity can reduce the core beam current drive, restoring agreement between the reconstructed profile and the summed constituents, as well as bringing better agreement between the simulated and measured neutron emission rate. An upper bound on the fast-ion diffusivity of ~0.5-1 m2 s-1 is found in 'MHD-free' discharges, based on the neutron emission, the time rate of change in the neutron signal when a neutral beam is stepped and reconstructed on-axis current density.

Comprehensive evaluation of the linear stability of Alfvén eigenmodes driven by alpha particles in an ITER baseline scenario

NASA Astrophysics Data System (ADS)

Figueiredo, A. C. A.; Rodrigues, P.; Borba, D.; Coelho, R.; Fazendeiro, L.; Ferreira, J.; Loureiro, N. F.; Nabais, F.; Pinches, S. D.; Polevoi, A. R.; Sharapov, S. E.

2016-07-01

The linear stability of Alfvén eigenmodes in the presence of fusion-born alpha particles is thoroughly assessed for two variants of an ITER baseline scenario, which differ significantly in their core and pedestal temperatures. A systematic approach based on CASTOR-K (Borba and Kerner 1999 J. Comput. Phys. 153 101; Nabais et al 2015 Plasma Sci. Technol. 17 89) is used that considers all possible eigenmodes for a given magnetic equilibrium and determines their growth rates due to alpha-particle drive and Landau damping on fuel ions, helium ashes and electrons. It is found that the fastest growing instabilities in the aforementioned ITER scenario are core-localized, low-shear toroidal Alfvén eigenmodes. The largest growth-rates occur in the scenario variant with higher core temperatures, which has the highest alpha-particle density and density gradient, for eigenmodes with toroidal mode numbers n≈ 30 . Although these eigenmodes suffer significant radiative damping, which is also evaluated, their growth rates remain larger than those of the most unstable eigenmodes found in the variant of the ITER baseline scenario with lower core temperatures, which have n≈ 15 and are not affected by radiative damping.

Application of adaptive gridding to magnetohydrodynamic flows

DOE Office of Scientific and Technical Information (OSTI.GOV)

Schnack, D.D.; Lotatti, I.; Satyanarayana, P.

1996-12-31

The numerical simulation of the primitive, three-dimensional, time-dependent, resistive MHD equations on an unstructured, adaptive poloidal mesh using the TRIM code has been reported previously. The toroidal coordinate is approximated pseudo-spectrally with finite Fourier series and Fast-Fourier Transforms. The finite-volume algorithm preserves the magnetic field as solenoidal to round-off error, and also conserves mass, energy, and magnetic flux exactly. A semi-implicit method is used to allow for large time steps on the unstructured mesh. This is important for tokamak calculations where the relevant time scale is determined by the poloidal Alfven time. This also allows the viscosity to be treatedmore » implicitly. A conjugate-gradient method with pre-conditioning is used for matrix inversion. Applications to the growth and saturation of ideal instabilities in several toroidal fusion systems has been demonstrated. Recently we have concentrated on the details of the mesh adaption algorithm used in TRIM. We present several two-dimensional results relating to the use of grid adaptivity to track the evolution of hydrodynamic and MHD structures. Examples of plasma guns, opening switches, and supersonic flow over a magnetized sphere are presented. Issues relating to mesh adaption criteria are discussed.« less

Gallium arsenide-gallium nitride wafer fusion and the n-aluminum gallium arsenide/p-gallium arsenide/n-gallium nitride double heterojunction bipolar transistor

NASA Astrophysics Data System (ADS)

Estrada, Sarah M.

This dissertation describes the n-AlGaAs/p-GaAs/n-GaN heterojunction bipolar transistor (HBT), the first transistor formed via wafer fusion. The fusion process was developed as a way to combine lattice-mismatched materials for high-performance electronic devices, not obtainable via conventional all-epitaxial formation methods. Despite the many challenges of wafer fusion, successful transistors were demonstrated and improved, via the optimization of material structure and fusion process conditions. Thus, this project demonstrated the integration of disparate device materials, chosen for their optimal electronic properties, unrestricted by the conventional (and very limiting) requirement of lattice-matching. By combining an AlGaAs-GaAs emitter-base with a GaN collector, the HBT benefited from the high breakdown voltage of GaN, and from the high emitter injection efficiency and low base transit time of AlGaAs-GaAs. Because the GaAs-GaN lattice mismatch precluded an all-epitaxial formation of the HBT, the GaAs-GaN heterostructure was formed via fusion. This project began with the development of a fusion process that formed mechanically robust and electrically active GaAs-GaN heterojunctions. During the correlation of device electrical performance with a systematic variation of fusion conditions over a wide range (500--750°C, 0.5--2hours), a mid-range fusion temperature was found to induce optimal HBT electrical performance. Transmission electron microscopy (TEM) and secondary ion mass spectrometry (SIMS) were used to assess possible reasons for the variations observed in device electrical performance. Fusion process conditions were correlated with electrical (I-V), structural (TEM), and chemical (SIMS) analyses of the resulting heterojunctions, in order to investigate the trade-off between increased interfacial disorder (TEM) with low fusion temperature and increased diffusion (SIMS) with high fusion temperature. The best do device results (IC ˜ 2.9 kA/cm2 and beta ˜ 3.5, at VCE = 20V and IB = 10mA) were obtained with an HBT formed via fusion at 600°C for 1 hour, with an optimized base-collector design. This was quite an improvement, as compared to an HBT with a simpler base-collector structure, also fused at 600°C for 1 hour (IC ˜ 0.83 kA/cm2 and beta ˜ 0.89, at VCE = 20V and IB = 10mA). Fused AlGaAs-GaAs-GaAs HBTs were compared to fused AlGaAs-GaAs-GaN HBTs, demonstrating that the use of a wider bandgap collector (Eg,GaN > Eg,GaAs) did indeed improve HBT performance at high applied voltages, as desired for high-power applications.

Advanced feedback control methods in EXTRAP T2R reversed field pinch

NASA Astrophysics Data System (ADS)

Yadikin, D.; Brunsell, P. R.; Paccagnella, R.

2006-07-01

Previous experiments in the EXTRAP T2R reversed field pinch device have shown the possibility of suppression of multiple resistive wall modes (RWM). A feedback system has been installed in EXTRAP T2R having 100% coverage of the toroidal surface by the active coil array. Predictions based on theory and the previous experimental results show that the number of active coils should be sufficient for independent stabilization of all unstable RWMs in the EXTRAP T2R. Experiments using different feedback schemes are performed, comparing the intelligent shell, the fake rotating shell, and the mode control with complex feedback gains. Stabilization of all unstable RWMs throughout the discharge duration of td≈10τw is seen using the intelligent shell feedback scheme. Mode rotation and the control of selected Fourier harmonics is obtained simultaneously using the mode control scheme with complex gains. Different sensor signals are studied. A feedback system with toroidal magnetic field sensors could have an advantage of lower feedback gain needed for the RWM suppression compared to the system with radial magnetic field sensors. In this study, RWM suppression is demonstrated, using also the toroidal field component as a sensor signal in the feedback system.

Comparing magnetic fluctuation dynamics in nonlinear MHD simulations of low-aspect-ratio RFPs to RELAX experiments

NASA Astrophysics Data System (ADS)

McCollam, K. J.; den Hartog, D. J.; Jacobson, C. M.; Sovinec, C. R.; Masamune, S.; Sanpei, A.

2017-10-01

We present comparisons of magnetic tearing fluctuation activity between RFP experiments on the low-aspect-ratio RELAX device (R / a 2) and nonlinear simulations of zero-beta, single-fluid MHD using the NIMROD code in both cylindrical and toroidal geometries at a Lundquist number of S =104 , nearly as high as experimental values. Time-average fluctuation amplitudes observed in the simulations are similar to those from the experiments, but more rigorous comparisons versus spectral mode numbers are in progress. We also focus on how the spatiotemporal dynamics of the fluctuations vary with RFP equilibrium parameters. Interestingly, at shallow reversal, cylindrical simulations show a relatively uncoupled spectrum of nearly quiescent modes periodically varying in time, whereas the corresponding toroidal cases show a fully chaotic spectrum of strongly nonlinearly interacting modes. We ascribe this to the geometric m = 1 coupling present in the toroidal but not the cylindrical case. We present initial results from convergence studies with increased spatial resolution for both geometries. Simulations at higher S are planned. This work is supported by the U.S. DOE and by the Japan Society for the Promotion of Science.

High efficiency and brightness fluorescent organic light emitting diode by triplet-triplet fusion

DOEpatents

Forrest, Stephen; Zhang, Yifan

2015-02-10

A first device is provided. The first device further comprises an organic light emitting device. The organic light emitting device further comprises an anode, a cathode, and an emissive layer disposed between the anode and the cathode. The emissive layer may include an organic host compound and at least one organic emitting compound capable of fluorescent emission at room temperature. Various configurations are described for providing a range of current densities in which T-T fusion dominates over S-T annihilation, leading to very high efficiency fluorescent OLEDs.

Evaluation of a polyetheretherketone (PEEK) titanium composite interbody spacer in an ovine lumbar interbody fusion model: biomechanical, microcomputed tomographic, and histologic analyses.

PubMed

McGilvray, Kirk C; Waldorff, Erik I; Easley, Jeremiah; Seim, Howard B; Zhang, Nianli; Linovitz, Raymond J; Ryaby, James T; Puttlitz, Christian M

2017-12-01

The most commonly used materials used for interbody cages are titanium metal and polymer polyetheretherketone (PEEK). Both of these materials have demonstrated good biocompatibility. A major disadvantage associated with solid titanium cages is their radiopacity, limiting the postoperative monitoring of spinal fusion via standard imaging modalities. However, PEEK is radiolucent, allowing for a temporal assessment of the fusion mass by clinicians. On the other hand, PEEK is hydrophobic, which can limit bony ingrowth. Although both PEEK and titanium have demonstrated clinical success in obtaining a solid spinal fusion, innovations are being developed to improve fusion rates and to create stronger constructs using hybrid additive manufacturing approaches by incorporating both materials into a single interbody device. The purpose of this study was to examine the interbody fusion characteristic of a PEEK Titanium Composite (PTC) cage for use in lumbar fusion. Thirty-four mature female sheep underwent two-level (L 2 -L 3 and L 4 -L 5 ) interbody fusion using either a PEEK or a PTC cage (one of each per animal). Animals were sacrificed at 0, 8, 12, and 18 weeks post surgery. Post sacrifice, each surgically treated functional spinal unit underwent non-destructive kinematic testing, microcomputed tomography scanning, and histomorphometric analyses. Relative to the standard PEEK cages, the PTC constructs demonstrated significant reductions in ranges of motion and a significant increase in stiffness. These biomechanical findings were reinforced by the presence of significantly more bone at the fusion site as well as ingrowth into the porous end plates. Overall, the results indicate that PTC interbody devices could potentially lead to a more robust intervertebral fusion relative to a standard PEEK device in a clinical setting. Copyright © 2017 The Author(s). Published by Elsevier Inc. All rights reserved.

Control of External Kink Instability

NASA Astrophysics Data System (ADS)

Navratil, Gerald

2004-11-01

A fundamental pressure and current limiting phenomenon in magnetically confined plasmas for fusion energy is the long wavelength ideal-MHD kink mode. These modes have been extensively studied in tokamak and reversed field pinch (RFP) devices. They are characterized by significant amplitude on the boundary of the confined plasma and can therefore be controlled by manipulation of the external boundary conditions. In the past ten years, the theoretically predicted stabilizing effect of a nearby conducting wall has been documented in experiments, which opens the possibility of a significant increase in maximum stable plasma pressure. While these modes are predicted to remain unstable when the stabilizing wall is resistive, their growth rates are greatly reduced from the hydrodynamic time scale to the time scale of magnetic diffusion through the resistive wall. These resistive wall slowed kink modes have been identified as limiting phenomena in tokamak (DIII-D, PBX-M, HBT-EP, JT-60U, JET, NSTX) and RFP (HBTX, Extrap, T2R) devices. The theoretical prediction of stabilization to nearly the ideal wall pressure limit by toroidal plasma rotation and/or active feedback control using coils has recently been realized experimentally. Sustained, stable operation at double the no-wall pressure limit has been achieved. Discovery of the phenomenon of resonant field amplification by marginally stable kink modes and its role in the momentum balance of rotationally stabilized plasmas has emerged as a key feature. A theoretical framework, based on an extension of the very successful treatment of the n=0 axisymmetric mode developed in the early 1990's, to understand the stabilization mechanisms and model the performance of active feedback control systems is now established. This allows design of kink control systems for burning plasma experiments like ITER.

Superconductivity and fusion energy—the inseparable companions

NASA Astrophysics Data System (ADS)

Bruzzone, Pierluigi

2015-02-01

Although superconductivity will never produce energy by itself, it plays an important role in energy-related applications both because of its saving potential (e.g., power transmission lines and generators), and its role as an enabling technology (e.g., for nuclear fusion energy). The superconducting magnet’s need for plasma confinement has been recognized since the early development of fusion devices. As long as the research and development of plasma burning was carried out on pulsed devices, the technology of superconducting fusion magnets was aimed at demonstrations of feasibility. In the latest generation of plasma devices, which are larger and have longer confinement times, the superconducting coils are a key enabling technology. The cost of a superconducting magnet system is a major portion of the overall cost of a fusion plant and deserves significant attention in the long-term planning of electricity supply; only cheap superconducting magnets will help fusion get to the energy market. In this paper, the technology challenges and design approaches for fusion magnets are briefly reviewed for past, present, and future projects, from the early superconducting tokamaks in the 1970s, to the current ITER (International Thermonuclear Experimental Reactor) and W7-X projects and future DEMO (Demonstration Reactor) projects. The associated cryogenic technology is also reviewed: 4.2 K helium baths, superfluid baths, forced-flow supercritical helium, and helium-free designs. Open issues and risk mitigation are discussed in terms of reliability, technology, and cost.

Lithium-based surfaces controlling fusion plasma behavior at the plasma-material interfacea)

NASA Astrophysics Data System (ADS)

Allain, Jean Paul; Taylor, Chase N.

2012-05-01

The plasma-material interface and its impact on the performance of magnetically confined thermonuclear fusion plasmas are considered to be one of the key scientific gaps in the realization of nuclear fusion power. At this interface, high particle and heat flux from the fusion plasma can limit the material's lifetime and reliability and therefore hinder operation of the fusion device. Lithium-based surfaces are now being used in major magnetic confinement fusion devices and have observed profound effects on plasma performance including enhanced confinement, suppression and control of edge localized modes (ELM), lower hydrogen recycling and impurity suppression. The critical spatial scale length of deuterium and helium particle interactions in lithium ranges between 5-100 nm depending on the incident particle energies at the edge and magnetic configuration. Lithium-based surfaces also range from liquid state to solid lithium coatings on a variety of substrates (e.g., graphite, stainless steel, refractory metal W/Mo/etc., or porous metal structures). Temperature-dependent effects from lithium-based surfaces as plasma facing components (PFC) include magnetohydrodynamic (MHD) instability issues related to liquid lithium, surface impurity, and deuterium retention issues, and anomalous physical sputtering increase at temperatures above lithium's melting point. The paper discusses the viability of lithium-based surfaces in future burning-plasma environments such as those found in ITER and DEMO-like fusion reactor devices.

PubMed Central

Schulz-Wendtland, Rüdiger; Jud, Sebastian M.; Fasching, Peter A.; Hartmann, Arndt; Radicke, Marcus; Rauh, Claudia; Uder, Michael; Wunderle, Marius; Gass, Paul; Langemann, Hanna; Beckmann, Matthias W.; Emons, Julius

2017-01-01

Aim The combination of different imaging modalities through the use of fusion devices promises significant diagnostic improvement for breast pathology. The aim of this study was to evaluate image quality and clinical feasibility of a prototype fusion device (fusion prototype) constructed from a standard tomosynthesis mammography unit and a standard 3D ultrasound probe using a new method of breast compression. Materials and Methods Imaging was performed on 5 mastectomy specimens from patients with confirmed DCIS or invasive carcinoma (BI-RADS ™ 6). For the preclinical fusion prototype an ABVS system ultrasound probe from an Acuson S2000 was integrated into a MAMMOMAT Inspiration (both Siemens Healthcare Ltd) and, with the aid of a newly developed compression plate, digital mammogram and automated 3D ultrasound images were obtained. Results The quality of digital mammogram images produced by the fusion prototype was comparable to those produced using conventional compression. The newly developed compression plate did not influence the applied x-ray dose. The method was not more labour intensive or time-consuming than conventional mammography. From the technical perspective, fusion of the two modalities was achievable. Conclusion In this study, using only a few mastectomy specimens, the fusion of an automated 3D ultrasound machine with a standard mammography unit delivered images of comparable quality to conventional mammography. The device allows simultaneous ultrasound – the second important imaging modality in complementary breast diagnostics – without increasing examination time or requiring additional staff. PMID:28713173

Nuclear fusion of advanced fuels using converging focused ion beams

NASA Astrophysics Data System (ADS)

Egle, Brian James

The Six Ion Gun Fusion Experiment (SIGFE) was designed and built to investigate a possible avenue to increase the reaction rate efficiency of the D-D and D-3He nuclear fusion reactions in Inertial Electrostatic Confinement (IEC) devices to the levels required for several non-electric applications of nuclear fusion. The SIGFE is based on the seminal IEC experiment published by Hirsch in 1967, and is the first experiment to recreate the results and unique features of the Hirsch device. The SIGFE used six identical ion beams to focus and converge deuterium and helium-3 ions into a sphere of less than 2 mm at nearly mono-energetic ion energies up to 150 keV. With improved ion optics and diagnostics, the SIGFE concluded that within the investigated parameter space, the region where the ion beams converged accounted for less than 0.2% of the total D-D fusion reactions. The maximum D-D fusion rates were observed when the ion beams were intentionally defocused to strike the inside surface of the cathode lenses. In this defocused state, the total D-D fusion rate increased when the chamber pressure was decreased. The maximum D-D fusion rate was 4.3 x 107 neutrons per second at a cathode voltage of -130 kV, a total cathode current of 10 mA, and a chamber pressure of 27 mPa. The D and 3He ion beams were produced in six self-contained ion gun modules. The modules were each capable of at least 4 mA of ion current while maintaining a main chamber pressure as low as 13 mPa. The theoretically calculated extractable ion current agreed with the experiment within a factor of 2. A concept was also developed and evaluated for the production of radioisotopes from the 14.7 MeV D-3He fusion protons produced in an IEC device. Monte Carlo simulations of this concept determined that a D-3He fusion rate on the order of 1011 s-1 would be required for an IEC device to produce 1 mCi of the 11C radioisotope.

Development of a magnetized coaxial plasma gun for compact toroid injection into the C-2 field-reversed configuration device

DOE Office of Scientific and Technical Information (OSTI.GOV)

Matsumoto, T., E-mail: cstd14003@g.nihon-u.ac.jp; Sekiguchi, J.; Asai, T.

A compact toroid (CT) injector was developed for the C-2 device, primarily for refueling of field-reversed configurations. The CTs are formed by a magnetized coaxial plasma gun (MCPG), which consists of coaxial cylindrical electrodes and a bias coil for creating a magnetic field. First, a plasma ring is generated by a discharge between the electrodes and is accelerated by Lorenz self-force. Then, the plasma ring is captured by an interlinkage flux (poloidal flux). Finally, the fully formed CT is ejected from the MCPG. The MCPG described herein has two gas injection ports that are arranged tangentially on the outer electrode.more » A tungsten-coated inner electrode has a head which can be replaced with a longer one to extend the length of the acceleration region for the CT. The developed MCPG has achieved supersonic CT velocities of ∼100 km/s. Plasma parameters for electron density, electron temperature, and the number of particles are ∼5 × 10{sup 21} m{sup −3}, ∼40 eV, and 0.5–1.0 × 10{sup 19}, respectively.« less

Quantifying opportunities for hospital cost control: medical device purchasing and patient discharge planning.

PubMed

Robinson, James C; Brown, Timothy T

2014-09-01

To quantify the potential reduction in hospital costs from adoption of best local practices in supply chain management and discharge planning. We performed multivariate statistical analyses of the association between total variable cost per procedure and medical device price and length of stay, controlling for patient and hospital characteristics. Ten hospitals in 1 major metropolitan area supplied patient-level administrative data on 9778 patients undergoing joint replacement, spine fusion, or cardiac rhythm management (CRM) procedures in 2008 and 2010. The impact on each hospital of matching lowest local market device prices and lowest patient length of stay (LOS) was calculated using multivariate regression analysis controlling for patient demographics, diagnoses, comorbidities, and implications. Average variable costs ranged from $11,315 for joint replacement to $16,087 for CRM and $18,413 for spine fusion. Implantable medical devices accounted for a large share of each procedure's variable costs: 44% for joint replacement, 39% for spine fusion, and 59% for CRM. Device prices and patient length-of-stay exhibited wide variation across hospitals. Total potential hospital cost savings from achieving best local practices in device prices and patient length of stay are 14.5% for joint replacement, 18.8% for spine fusion;,and 29.1% for CRM. Hospitals have opportunities for cost reduction from adoption of best local practices in supply chain management and discharge planning.

Tokamak with mechanical compression of toroidal magnetic field

DOEpatents

Ohkawa, Tihiro

1981-01-01

A tokamak apparatus includes a pressure vessel for defining a reservoir and confining liquid therein. A collapsible toroidal liner disposed within the pressure vessel defines a toroidal space within the liner. Liquid metal fills the reservoir outside said liner. A toroidal magnetic field is developed within the toroidal space about the major axis thereof. A toroidal plasma is developed within the toroidal space about the major axis thereof. Pressure is applied to the liquid metal to collapse the liner and reduce the volume of the toroidal space, thereby increasing the toroidal magnetic flux density therein.

Fluctuations and intermittent poloidal transport in a simple toroidal plasma

DOE Office of Scientific and Technical Information (OSTI.GOV)

Goud, T. S.; Ganesh, R.; Saxena, Y. C.

In a simple magnetized toroidal plasma, fluctuation induced poloidal flux is found to be significant in magnitude. The probability distribution function of the fluctuation induced poloidal flux is observed to be strongly non-Gaussian in nature; however, in some cases, the distribution shows good agreement with the analytical form [Carreras et al., Phys. Plasmas 3, 2664 (1996)], assuming a coupling between the near Gaussian density and poloidal velocity fluctuations. The observed non-Gaussian nature of the fluctuation induced poloidal flux and other plasma parameters such as density and fluctuating poloidal velocity in this device is due to intermittent and bursty nature ofmore » poloidal transport. In the simple magnetized torus used here, such an intermittent fluctuation induced poloidal flux is found to play a crucial role in generating the poloidal flow.« less

Non-axisymmetric equilibrium reconstruction of a current-carrying stellarator using external magnetic and soft x-ray inversion radius measurements

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ma, X., E-mail: xzm0005@auburn.edu; Maurer, D. A.; Knowlton, S. F.

2015-12-15

Non-axisymmetric free-boundary equilibrium reconstructions of stellarator plasmas are performed for discharges in which the magnetic configuration is strongly modified by ohmically driven plasma current. These studies were performed on the compact toroidal hybrid device using the V3FIT reconstruction code with a set of 50 magnetic diagnostics external to the plasma. With the assumption of closed magnetic flux surfaces, the reconstructions using external magnetic measurements allow accurate estimates of the net toroidal flux within the last closed flux surface, the edge safety factor, and the plasma shape of these highly non-axisymmetric plasmas. The inversion radius of standard sawteeth is used tomore » infer the current profile near the magnetic axis; with external magnetic diagnostics alone, the current density profile is imprecisely reconstructed.« less

Non-axisymmetric equilibrium reconstruction of a current-carrying stellarator using external magnetic and soft x-ray inversion radius measurements

NASA Astrophysics Data System (ADS)

Ma, X.; Maurer, D. A.; Knowlton, S. F.; ArchMiller, M. C.; Cianciosa, M. R.; Ennis, D. A.; Hanson, J. D.; Hartwell, G. J.; Hebert, J. D.; Herfindal, J. L.; Pandya, M. D.; Roberds, N. A.; Traverso, P. J.

2015-12-01

Non-axisymmetric free-boundary equilibrium reconstructions of stellarator plasmas are performed for discharges in which the magnetic configuration is strongly modified by ohmically driven plasma current. These studies were performed on the compact toroidal hybrid device using the V3FIT reconstruction code with a set of 50 magnetic diagnostics external to the plasma. With the assumption of closed magnetic flux surfaces, the reconstructions using external magnetic measurements allow accurate estimates of the net toroidal flux within the last closed flux surface, the edge safety factor, and the plasma shape of these highly non-axisymmetric plasmas. The inversion radius of standard sawteeth is used to infer the current profile near the magnetic axis; with external magnetic diagnostics alone, the current density profile is imprecisely reconstructed.

Fluid interaction with spinning toroidal tanks

NASA Technical Reports Server (NTRS)

Fester, D. A.; Anderson, J. E.

1977-01-01

An experimental study was conducted to evaluate propellant behavior in spinning torroidal tanks that could be used in a retropropulsion system of an advanced outer-planet Pioneer orbiter. Information on propellant slosh and settling and on ullage orientation and stability was obtained. The effects of axial acceleration, spin rate, spin-rate change, and spacecraft wobble, both singly and in combination, were evaluated using a one-eighth scale transparent tank in one-g and low-g environments. Liquid loadings ranged from 5% to 96% full. The impact of a surface tension acquisition device was assessed by comparison with bare-tank results. The testing simulated the behavior of the fluorine/hydrazine and nitrogen textroxide/monomethylhydrazine propellants. Results are presented that indicate that no major fluid behavior problems would be encountered with any of the four propellants in the toroidal tanks of a spin-stabilized orbiter spacecraft.

Non-axisymmetric equilibrium reconstruction of a current-carrying stellarator using external magnetic and soft x-ray inversion radius measurements

DOE PAGES

Ma, X.; Maurer, D. A.; Knowlton, Stephen F.; ...

2015-12-22

Non-axisymmetric free-boundary equilibrium reconstructions of stellarator plasmas are performed for discharges in which the magnetic configuration is strongly modified by ohmically driven plasma current. These studies were performed on the compact toroidal hybrid device using the V3FIT reconstruction code with a set of 50 magnetic diagnostics external to the plasma. With the assumption of closed magnetic flux surfaces, the reconstructions using external magnetic measurements allow accurate estimates of the net toroidal flux within the last closed flux surface, the edge safety factor, and the plasma shape of these highly non-axisymmetric plasmas. Lastly, the inversion radius of standard saw-teeth is usedmore » to infer the current profile near the magnetic axis; with external magnetic diagnostics alone, the current density profile is imprecisely reconstructed.« less

All-dielectric perforated metamaterials with toroidal dipolar response (Conference Presentation)

NASA Astrophysics Data System (ADS)

Stenishchev, Ivan; Basharin, Alexey A.

2017-05-01

We present metamaterials based on dielectric slab with perforated identical cylindrical clusters with perforated holes, which allow to support the toroidal dipolar response due to Mie-resonances in each hole. Note that proposed metamaterial is technologically simple for fabrication in optical frequency range. Metamaterial can be fabricated by several methods. For instance, we may apply the molecular beam epitaxy method for deposition of Si or GaAs layers, which have permittivity close to 16. Next step, nanometer/micrometer holes are perforated by focused ion beam method or laser cutting method. Fundamental difference of proposed metamaterial is technological fabrication process. Classically all- dielectric optical metamaterials consist of nano-spheres or nano-discs, which are complicated for fabrication, while our idea and suggested metamaterials are promising prototype of various optical/THz all-dielectic devices as sensor, nano-antennas elements for nanophotonics.

Resistive magnetohydrodynamics with toroidal rotation in toroidal plasmas

NASA Astrophysics Data System (ADS)

Cao, Jintao; Cai, Huishan

2018-01-01

Toroidal rotation has always existed in tokamak plasmas, and its Mach number can reach unity during neutral beam injection. Toroidal rotation can affect plasma equilibrium and magnetohydrodynamic instabilities significantly. Based on linearized equations including the toroidal rotation effect, the toroidal model derived by Glasser et al. [Phys. Fluids 18, 875 (1975)] is extended to include this effect, and a set of resistive equations including the toroidal rotation effect in the axi-symmetry toroidal geometry is derived. Based on these derived equations, the effect of toroidal rotation on tearing modes is considered, and the growth rate of tearing modes is obtained analytically. It is shown that the effect of toroidal rotation on tearing modes depends on both the direction of toroidal rotation flow and the sign of toroidal rotation flow shear. When they have the same sign, they play a role in stabilizing tearing modes, while when they have opposite signs, they have a destabilizing effect on tearing modes.

Asymptotic and spectral analysis of the gyrokinetic-waterbag integro-differential operator in toroidal geometry

DOE Office of Scientific and Technical Information (OSTI.GOV)

Besse, Nicolas, E-mail: Nicolas.Besse@oca.eu; Institut Jean Lamour, UMR CNRS/UL 7198, Université de Lorraine, BP 70239 54506 Vandoeuvre-lès-Nancy Cedex; Coulette, David, E-mail: David.Coulette@ipcms.unistra.fr

2016-08-15

Achieving plasmas with good stability and confinement properties is a key research goal for magnetic fusion devices. The underlying equations are the Vlasov–Poisson and Vlasov–Maxwell (VPM) equations in three space variables, three velocity variables, and one time variable. Even in those somewhat academic cases where global equilibrium solutions are known, studying their stability requires the analysis of the spectral properties of the linearized operator, a daunting task. We have identified a model, for which not only equilibrium solutions can be constructed, but many of their stability properties are amenable to rigorous analysis. It uses a class of solution to themore » VPM equations (or to their gyrokinetic approximations) known as waterbag solutions which, in particular, are piecewise constant in phase-space. It also uses, not only the gyrokinetic approximation of fast cyclotronic motion around magnetic field lines, but also an asymptotic approximation regarding the magnetic-field-induced anisotropy: the spatial variation along the field lines is taken much slower than across them. Together, these assumptions result in a drastic reduction in the dimensionality of the linearized problem, which becomes a set of two nested one-dimensional problems: an integral equation in the poloidal variable, followed by a one-dimensional complex Schrödinger equation in the radial variable. We show here that the operator associated to the poloidal variable is meromorphic in the eigenparameter, the pulsation frequency. We also prove that, for all but a countable set of real pulsation frequencies, the operator is compact and thus behaves mostly as a finite-dimensional one. The numerical algorithms based on such ideas have been implemented in a companion paper [D. Coulette and N. Besse, “Numerical resolution of the global eigenvalue problem for gyrokinetic-waterbag model in toroidal geometry” (submitted)] and were found to be surprisingly close to those for the original gyrokinetic-Vlasov equations. The purpose of the present paper is to make these new ideas accessible to two readerships: applied mathematicians and plasma physicists.« less

Development of Long-Pulse Heating and Current Drive Actuators and Operational Techniques Compatible with a High-Z Divertor and First Wall

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wang, Guiding

Accurate measurement of the edge electron density profile is essential to optimizing antenna coupling and assessment of impurity contamination in studying long-pulse plasma heating and current drive in fusion devices. Measurement of the edge density profile has been demonstrated on the US fusion devices such as C-Mod, DIII-D, and TFTR amongst many devices, and has been used for RF loading and impurity modeling calculations for many years. University of Science and Technology of China (USTC) has recently installed a density profile reflectometer system on the EAST fusion device at the Institute of Plasma Physics, Chinese Academy of Sciences in Chinamore » based on the University of California Los Angeles (UCLA)-designed reflectometer system on the DIII-D fusion device at General Atomics Company in San Diego, California. UCLA has been working with USTC to optimize the existing microwave antenna, waveguide system, microwave electronics, and data analysis to produce reliable edge density profiles. During the past budget year, progress has been made in all three major areas: effort to achieve reliable system operations under various EAST operational conditions, effort to optimize system performance, and effort to provide quality density profiles into EAST’s database routinely.« less

Development of Laser Based Plasma Diagnostics for Fusion Research on NSTX-U

NASA Astrophysics Data System (ADS)

Barchfeld, Robert Adam

Worldwide demand for power, and in particular electricity, is growing. Increasing population, expanding dependence on electrical devices, as well as the development of emerging nations, has created significant challenges for the power production. Compounding the issue are concerns over pollution, natural resource supplies, and political obstacles in troubled parts of the world. Many believe that investment in renewable energy will solve the expected energy crisis; however, renewable energy has many shortfalls. Consequently, additional sources of energy should be explored to provide the best options for the future. Electricity from fusion power offers many advantages over competing technologies. It can potentially produce large amounts of clean energy, without the serious concerns of fission power plant safety and nuclear waste. Fuel supplies for fusion are plentiful. Fusion power plants can be operated as needed, without dependence on location, or local conditions. However, there are significant challenges before fusion can be realized. Many factors currently limit the effectiveness of fusion power, which prevents a commercial power plant from being feasible. Scientists in many countries have built, and operate, experimental fusion plants to study the fusion process. The leading examples are magnetic confinement reactors known as tokamaks. At present, reactor gain is near unity, where the fusion power output is nearly the same as the power required to operate the reactor. A tenfold increase in gain is what reactors such as ITER hope to achieve, where 50 MW will be used for plasma heating, magnetic fields, and so forth, with a power output of 500 MW. Before this can happen, further research is required. Loss of particle and energy confinement is a principal cause of low performance; therefore, increasing confinement time is key. There are many causes of thermal and particle transport that are being researched, and the prime tools for conducting this research are plasma diagnostics. Plasma diagnostics collect data from fusion reactors in a number of different ways. Among these are far infrared (FIR) laser based systems. By probing a fusion plasma with FIR lasers, many properties can be measured, such as density and density fluctuations. This dissertation discusses the theory and design of two laser based diagnostic instruments: 1) the Far Infrared Tangential Interferometer and Polarimeter (FIReTIP) systems, and 2) the High-ktheta Scattering System. Both of these systems have been designed and fabricated at UC Davis for use on the National Spherical Torus Experiment - Upgrade (NSTX-U), located at Princeton Plasma Physics Laboratory (PPPL). These systems will aid PPPL scientists in fusion research. The FIReTIP system uses 119 ?m methanol lasers to pass through the plasma core to measure a chord averaged plasma density through interferometry. It can also measure the toroidal magnetic field strength by the way of polarimetery. The High-ktheta Scattering System uses a 693 GHz formic acid laser to measure electron scale turbulence. Through collective Thomson scattering, as the probe beam passes through the plasma, collective electron motion will scatter power to a receiver with the angle determined by the turbulence wavenumber. This diagnostic will measure ktheta from 7 to 40 cm-1 with a 4-channel receiver array. The High-ktheta Scattering system was designed to facilitate research on electron temperature gradient (ETG) modes, which are believed to be a major contributor to anomalous transport on NSTX-U. The design and testing of these plasma diagnostics are described in detail. There are a broad range of components detailed including: optically pumped gas FIR lasers, overmoded low loss waveguide, launching and receiving optical designs, quasi-optical mixers, electronics, and monitoring and control systems. Additionally, details are provided for laser maintenance, alignment techniques, and the fundamentals of nano-CNC-machining.

A non-linear 4-wave resonant model for non-perturbative fast ion interactions with Alfv'enic modes in burning plasmas

NASA Astrophysics Data System (ADS)

Zonca, Fulvio; Chen, Liu

2007-11-01

We adopt the 4-wave modulation interaction model, introduced by Chen et al [1] for analyzing modulational instabilities of the radial envelope of Ion Temperature Gradient driven modes in toroidal geometry, extending it to the modulations on the fast particle distribution function due to nonlinear Alfv'enic mode dynamics, as proposed in Ref. [2]. In the case where the wave-particle interactions are non-perturbative and strongly influence the mode evolution, as in the case of Energetic Particle Modes (EPM) [3], radial distortions (redistributions) of the fast ion source dominate the mode nonlinear dynamics. In this work, we show that the resonant particle motion is secular with a time-scale inversely proportional to the mode amplitude [4] and that the time evolution of the EPM radial envelope can be cast into the form of a nonlinear Schr"odinger equation a la Ginzburg-Landau [5]. [1] L. Chen et al, Phys. Plasmas 7 3129 (2000) [2] F. Zonca et al, Theory of Fusion Plasmas (Bologna: SIF) 17 (2000) [3] L. Chen, Phys. Plasmas 1, 1519 (1994).[4] F. Zonca et al, Nucl. Fusion 45 477 (2005) [5] F. Zonca et al, Plasma Phys. Contr. Fusion 48 B15 (2006)

Influence of the Al wire placed in the anode axis on the transformation of the deuterium plasma column in the plasma focus discharge

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kubes, P.; Cikhardtova, B.; Cikhardt, J.

In this paper, we describe the influence of an Al wire of 270 μm in diameter placed along the anode axis on the transformation of the deuterium pinch column in a megaampere (MA) plasma focus device. The evolution of the pinched column and of the wire corona was investigated by means of the multiframe interferometry, neutron and X-ray diagnostics. The wire corona did not influence considerably on the evolution of dense plasma structures and neutron production, but it increased the plasma density and consequently, the currents around its surface. The distribution of the closed internal currents (ranging hundreds of kA) andmore » associated magnetic fields amounting to 5 T were also estimated in the dense plasma column and in plasmoidal structures at the near-equilibrium state. The description is based on the balance of the plasma pressure and the pressure of the internal poloidal and toroidal current components compressed by the external pinched column. The dominant number of fusion deuterium-deuterium (D-D) neutrons is produced during the evolution of instabilities, when the uninterrupted wire corona (containing deuterium) connects the dense structures of the pinch, and it did not allow the formation of a constriction of the sub-millimeter diameter.« less

Flux amplification and sustainment of ST plasmas by multi-pulsed coaxial helicity injection on HIST

NASA Astrophysics Data System (ADS)

Higashi, T.; Ishihara, M.; Kikuchi, Y.; Fukumoto, N.; Nagata, M.

2010-11-01

The Helicity Injected Spherical Torus (HIST) device has been developed towards high-current start up and sustainment by Multi-pulsed Coaxial Helicity Injection (M-CHI) method. Multiple pulses operation of the coaxial plasma gun can build the magnetic field of STs and spheromak plasmas in a stepwise manner. So far, successive gun pulses on SSPX at LLNL were demonstrated to maintain the magnetic field of spheromak in a quasi-steady state against resistive decay [1]. The resistive 3D-MHD numerical simulation [2] for STs reproduced the current amplification by the M-CHI method and confirmed that stochastic magnetic field was reduced during the decay phase. By double pulsed operation on HIST, the plasma current was effectively amplified against the resistive decay. The life time increases up to 10 ms which is longer than that in the single CHI case (4 ms). The edge poloidal fields last between 0.5 ms and 6 ms like a repetitive manner. During the second driven phase, the toroidal ion flow is driven in the same direction as the plasma current as well as in the initial driven phase. At the meeting, we will discuss a current amplification mechanism based on the merging process with the plasmoid injected secondly from the gun. [1] B. Hudson et al., Phys. Plasmas Vol.15, 056112 (2008). [2] Y. Kagei et al., J. Plasma Fusion Res. Vol.79, 217 (2003).

Non-perturbative measurement of cross-field thermal diffusivity reduction at the O-point of 2/1 neoclassical tearing mode islands in the DIII-D tokamak

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bardóczi, L.; Rhodes, T. L.; Carter, T. A.

2016-05-15

Neoclassical tearing modes (NTMs) often lead to the decrease of plasma performance and can lead to disruptions, which makes them a major impediment in the development of operating scenarios in present toroidal fusion devices. Recent gyrokinetic simulations predict a decrease of plasma turbulence and cross-field transport at the O-point of the islands, which in turn affects the NTM dynamics. In this paper, a heat transport model of magnetic islands employing spatially non-uniform cross-field thermal diffusivity (χ{sub ⊥}) is presented. This model is used to derive χ{sub ⊥} at the O-point from electron temperature data measured across 2/1 NTM islands inmore » DIII-D. It was found that χ{sub ⊥} at the O-point is 1 to 2 orders of magnitude smaller than the background plasma transport, in qualitative agreement with gyrokinetic predictions. As the anomalously large values of χ{sub ⊥} are often attributed to turbulence driven transport, the reduction of the O-point χ{sub ⊥} is consistent with turbulence reduction found in recent experiments. Finally, the implication of reduced χ{sub ⊥} at the O-point on NTM dynamics was investigated using the modified Rutherford equation that predicts a significant effect of reduced χ{sub ⊥} at the O-point on NTM saturation.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Madduri, Kamesh; Im, Eun-Jin; Ibrahim, Khaled Z.

The next decade of high-performance computing (HPC) systems will see a rapid evolution and divergence of multi- and manycore architectures as power and cooling constraints limit increases in microprocessor clock speeds. Understanding efficient optimization methodologies on diverse multicore designs in the context of demanding numerical methods is one of the greatest challenges faced today by the HPC community. In this paper, we examine the efficient multicore optimization of GTC, a petascale gyrokinetic toroidal fusion code for studying plasma microturbulence in tokamak devices. For GTC’s key computational components (charge deposition and particle push), we explore efficient parallelization strategies across a broadmore » range of emerging multicore designs, including the recently-released Intel Nehalem-EX, the AMD Opteron Istanbul, and the highly multithreaded Sun UltraSparc T2+. We also present the first study on tuning gyrokinetic particle-in-cell (PIC) algorithms for graphics processors, using the NVIDIA C2050 (Fermi). Our work discusses several novel optimization approaches for gyrokinetic PIC, including mixed-precision computation, particle binning and decomposition strategies, grid replication, SIMDized atomic floating-point operations, and effective GPU texture memory utilization. Overall, we achieve significant performance improvements of 1.3–4.7× on these complex PIC kernels, despite the inherent challenges of data dependency and locality. Finally, our work also points to several architectural and programming features that could significantly enhance PIC performance and productivity on next-generation architectures.« less

Non-perturbative measurement of cross-field thermal diffusivity reduction at the O-point of 2/1 neoclassical tearing mode islands in the DIII-D tokamak

DOE PAGES

Bardóczi, Laszlo; Rhodes, Terry L.; Carter, Troy A.; ...

2016-05-09

Neoclassical Tearing Modes (NTMs) often lead to the decrease of plasma performance and can lead to disruptions, which makes them a major impediment in the development of operating scenarios in present toroidal fusion devices. Recent gyrokinetic simulations predict a decrease of plasma turbulence and cross- eld transport at the O-point of the islands, which in turn affects the NTM dynamics. In this paper a heat transport model of magnetic islands employing spatially non-uniform cross-field thermal diffusivity (χ more » $$\\perp$$) is presented. This model is used to derive χ $$\\perp$$ at the O-point from electron temperature data measured across 2/1 NTM islands in DIII-D. It was found that χ $$\\perp$$ at the O-point is 1 to 2 orders of magnitude smaller than the background plasma transport, in qualitative agreement with gyrokinetic predictions. As the anomalously large values of χ $$\\perp$$ are often attributed to turbulence driven transport, the reduction of the O-point χ $$\\perp$$ is consistent with turbulence reduction found in recent experiments. Lastly, the implication of reduced χ $$\\perp$$ at the O-point on NTM dynamics was investigated using the modi ed Rutherford equation that predicts a significant effect of reduced χ $$\\perp$$ at the O-point on NTM saturation.« less

Competition of Perpendicular and Parallel Flows in a Straight Magnetic Field

NASA Astrophysics Data System (ADS)

Li, Jiacong; Diamond, Patrick; Hong, Rongjie; Tynan, George

2017-10-01

In tokamaks, intrinsic rotations in both toroidal and poloidal directions are important for the stability and confinement. Since they compete for energy from background turbulence, the coupling of them is the key to understanding the physics of turbulent state and transport bifurcations, e.g. L-H transition. V⊥ can affect the parallel Reynolds stress via cross phase and energetics, and thus regulates the parallel flow generation. In return, the turbulence driven V∥ plays a role in the mean vorticity flux, influencing the generation of V⊥. Also, competition of intrinsic azimuthal and axial flows is observed in CSDX-a linear plasma device with straight magnetic fields. CSDX is a well diagnosed venue to study the basic physics of turbulence-flow interactions in straight magnetic fields. Here, we study the turbulent energy branching between the turbulence driven parallel flow and perpendicular flow. Specifically, the ratio between parallel and perpendicular Reynolds power decreases when the mean perpendicular flow increases. As the mean parallel flow increases, this ratio first increases and then decreases before the parallel flow shear hits the parallel shear flow instability threshold. We seek to understand the flow states and compare with CSDX experiments. This material is based upon work supported by the U.S. Department of Energy, Office of Science, Office of Fusion Energy Sciences, under Award Number DE-FG02-04ER54738.

Toroidal Rotation and 3D Nonlinear Dynamics in the Peeling-Ballooning Model of ELMs

NASA Astrophysics Data System (ADS)

Snyder, P. B.

2004-11-01

Maximizing the height of the edge transport barrier (or ``pedestal'') while maintaining acceptably small edge localized modes (ELMs) is a critical issue for tokamak performance. The peeling-ballooning model proposes that intermediate wavelength MHD instabilities are responsible for ELMs and impose constraints on the pedestal. Recent studies of linear peeling-ballooning stability have found encouraging agreement with observations [e.g. 1]. To allow more detailed prediction of mode characteristics, including eventually predictions of the ELM energy loss and its deposition, we consider effects of sheared toroidal rotation, as well as 3D nonlinear dynamics. An eigenmode formulation for toroidal rotation shear is developed and incorporated into the framework of the ELITE stability code [2], resolving the low rotation discontinuity in previous high-n results. Rotation shear is found to impact the structure of peeling-ballooning modes, causing radial narrowing and mode shearing. The calculated mode frequency is found to agree with observed rotation in the edge region in the early stages of the ELM crash. Nonlinear studies with the 3D BOUT and NIMROD codes reveal detailed characteristics of the early evolution of these edge instabilities, including the impact of non-ideal effects. The expected linear growth phase is followed by a fast crash event in which poloidally narrow, filamentary structures propagate radially outward from the pedestal region, closely resembling observed ELM events. Comparisons with ELM observations will be discussed. \\vspace0.25em [1] P.B. Snyder et al., Nucl. Fusion 44, 320 (2004); P.B. Snyder et al., Phys. Plasmas 9, 2037 (2002). [2] H.R. Wilson et al., Phys. Plasmas 9, 1277 (2002).

Effects of density gradient caused by multi-pulsing CHI on two-fluid flowing equilibria of spherical torus plasmas

NASA Astrophysics Data System (ADS)

Kanki, T.; Nagata, M.

2014-10-01

Two-fluid dynamo relaxation is examined to understand sustainment mechanism of spherical torus (ST) plasmas by multi-pulsing CHI (M-CHI) in the HIST device. The steeper density gradient between the central open flux column (OFC) and closed flux regions by applying the second CHI pulse is observed to cause not only the E × B drift but also the ion diamagnetic drift, leading the two-fluid dynamo. The purpose of this study is to investigate the effects of the steep change in the density gradient on the ST equilibria by using the two-fluid equilibrium calculations. The toroidal magnetic field becomes from a diamagnetic to a paramagnetic profile in the closed flux region while it remains a diamagnetic profile in the OFC region. The toroidal ion flow velocity is increased from negative to positive values in the closed flux region. Here, the negative ion flow velocity is the opposite direction to the toroidal current. The poloidal ion flow velocity between the OFC and closed flux regions is increased, because the ion diamagnetic drift velocity is changed in the same direction as the E × B drift velocity through the steeper ion pressure gradient. As a result, the strong shear flow and the paramagnetic toroidal field are generated in the closed flux region. Here, the ion flow velocity is the same direction as the poloidal current. The radial electric field shear between the OFC and closed flux regions is enhanced due to the strong dependence on the magnetic force through the interaction of toroidal ion flow velocity and axial magnetic field. The two-fluid effect is significant there due to the ion diamagnetic effect.

Tokamak with liquid metal toroidal field coil

DOEpatents

Ohkawa, Tihiro; Schaffer, Michael J.

1981-01-01

Tokamak apparatus includes a pressure vessel for defining a reservoir and confining liquid therein. A toroidal liner disposed within the pressure vessel defines a toroidal space within the liner. Liquid metal fills the reservoir outside said liner. Electric current is passed through the liquid metal over a conductive path linking the toroidal space to produce a toroidal magnetic field within the toroidal space about the major axis thereof. Toroidal plasma is developed within the toroidal space about the major axis thereof.

Fusion Energy Division progress report, 1 January 1990--31 December 1991

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sheffield, J.; Baker, C.C.; Saltmarsh, M.J.

1994-03-01

The Fusion Program of the Oak Ridge National Laboratory (ORNL), a major part of the national fusion program, encompasses nearly all areas of magnetic fusion research. The program is directed toward the development of fusion as an economical and environmentally attractive energy source for the future. The program involves staff from ORNL, Martin Marietta Energy systems, Inc., private industry, the academic community, and other fusion laboratories, in the US and abroad. Achievements resulting from this collaboration are documented in this report, which is issued as the progress report of the ORNL Fusion Energy Division; it also contains information from componentsmore » for the Fusion Program that are external to the division (about 15% of the program effort). The areas addressed by the Fusion Program include the following: experimental and theoretical research on magnetic confinement concepts; engineering and physics of existing and planned devices, including remote handling; development and testing of diagnostic tools and techniques in support of experiments; assembly and distribution to the fusion community of databases on atomic physics and radiation effects; development and testing of technologies for heating and fueling fusion plasmas; development and testing of superconducting magnets for containing fusion plasmas; development and testing of materials for fusion devices; and exploration of opportunities to apply the unique skills, technology, and techniques developed in the course of this work to other areas (about 15% of the Division`s activities). Highlights from program activities during 1990 and 1991 are presented.« less

Simple map in action-angle coordinates

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kerwin, Olivia; Punjabi, Alkesh; Ali, Halima

A simple map [A. Punjabi, A. Verma, and A. Boozer, Phys. Rev. Lett. 69, 3322 (1992)] is the simplest map that has the topology of divertor tokamaks [A. Punjabi, H. Ali, T. Evans, and A. Boozer, Phys. Lett. A 364, 140 (2007)]. Here, action-angle coordinates, the safety factor, and the equilibrium generating function for the simple map are calculated analytically. The simple map in action-angle coordinates is derived from canonical transformations. This map cannot be integrated across the separatrix surface because of the singularity in the safety factor there. The stochastic broadening of the ideal separatrix surface in action-angle representationmore » is calculated by adding a perturbation to the simple map equilibrium generating function. This perturbation represents the spatial noise and field errors typical of the DIII-D [J. L. Luxon and L. E. Davis, Fusion Technol. 8, 441 (1985)] tokamak. The stationary Fourier modes of the perturbation have poloidal and toroidal mode numbers (m,n,)=((3,1),(4,1),(6,2),(7,2),(8,2),(9,3),(10,3),(11,3)) with amplitude {delta}=0.8x10{sup -5}. Near the X-point, about 0.12% of toroidal magnetic flux inside the separatrix, and about 0.06% of the poloidal flux inside the separatrix is lost. When the distance from the O-point to the X-point is 1 m, the width of stochastic layer near the X-point is about 1.4 cm. The average value of the action on the last good surface is 0.19072 compared to the action value of 3/5{pi} on the separatrix. The average width of stochastic layer in action coordinate is 2.7x10{sup -4}, while the average area of the stochastic layer in action-angle phase space is 1.69017x10{sup -3}. On average, about 0.14% of action or toroidal flux inside the ideal separatrix is lost due to broadening. Roughly five times more toroidal flux is lost in the simple map than in DIII-D for the same perturbation [A. Punjabi, H. Ali, A. Boozer, and T. Evans, Bull. Amer. Phys. Soc. 52, 124 (2007)].« less

Simple map in action-angle coordinates

NASA Astrophysics Data System (ADS)

Kerwin, Olivia; Punjabi, Alkesh; Ali, Halima

2008-07-01

A simple map [A. Punjabi, A. Verma, and A. Boozer, Phys. Rev. Lett. 69, 3322 (1992)] is the simplest map that has the topology of divertor tokamaks [A. Punjabi, H. Ali, T. Evans, and A. Boozer, Phys. Lett. A 364, 140 (2007)]. Here, action-angle coordinates, the safety factor, and the equilibrium generating function for the simple map are calculated analytically. The simple map in action-angle coordinates is derived from canonical transformations. This map cannot be integrated across the separatrix surface because of the singularity in the safety factor there. The stochastic broadening of the ideal separatrix surface in action-angle representation is calculated by adding a perturbation to the simple map equilibrium generating function. This perturbation represents the spatial noise and field errors typical of the DIII-D [J. L. Luxon and L. E. Davis, Fusion Technol. 8, 441 (1985)] tokamak. The stationary Fourier modes of the perturbation have poloidal and toroidal mode numbers (m,n,)={(3,1),(4,1),(6,2),(7,2),(8,2),(9,3),(10,3),(11,3)} with amplitude δ =0.8×10-5. Near the X-point, about 0.12% of toroidal magnetic flux inside the separatrix, and about 0.06% of the poloidal flux inside the separatrix is lost. When the distance from the O-point to the X-point is 1m, the width of stochastic layer near the X-point is about 1.4cm. The average value of the action on the last good surface is 0.19072 compared to the action value of 3/5π on the separatrix. The average width of stochastic layer in action coordinate is 2.7×10-4, while the average area of the stochastic layer in action-angle phase space is 1.69017×10-3. On average, about 0.14% of action or toroidal flux inside the ideal separatrix is lost due to broadening. Roughly five times more toroidal flux is lost in the simple map than in DIII-D for the same perturbation [A. Punjabi, H. Ali, A. Boozer, and T. Evans, Bull. Amer. Phys. Soc. 52, 124 (2007)].

Salvage of infected total knee fusion: the last option.

PubMed

Wiedel, Jerome D

2002-11-01

Currently the most common indication for an arthrodesis of the knee is a failed infected total knee prosthesis. Other causes of a failed total knee replacement that might necessitate a knee fusion include aseptic loosening, deficient extensor mechanism, poor soft tissues, and Charcot joint. Techniques available for achieving a knee fusion are external fixation and internal fixation methods. The external fixation compression devices have been the most widely used for knee fusion and have been successful until the indications for fusion changed to mostly failed prosthetic knee replacement. With failed total knee replacement, the problem of severe bone loss became an issue, and the external fixation compression devices, even including the biplane external fixators, have been the least successful method reported for gaining fusion. The Ilizarov technique has been shown to achieve rigid fixation despite this bone loss, and a review of reports are showing high fusion rates using this method. Internal fixation methods including plate fixation and intramedullary nails have had the best success in gaining fusion in the face of this bone loss and have replaced external fixation methods as the technique of choice for knee fusion when severe bone loss is present. A review of the literature and a discussion of different fusion techniques are presented including a discussion of the influence that infection has on the success of fusion.

Gasdynamic Mirror Fusion Propulsion Experiment

NASA Technical Reports Server (NTRS)

Emrich, William J., Jr.; Rodgers, Stephen L. (Technical Monitor)

2001-01-01

Nuclear fusion appears to be the most promising concept for producing extremely high specific impulse rocket engines. One particular fusion concept which seems to be particularly well suited for fusion propulsion applications is the gasdynamic mirror (GDM). This device would operate at much higher plasma densities and with much larger LD ratios than previous mirror machines. Several advantages accrue from such a design. First, the high LA:) ratio minimizes to a large extent certain magnetic curvature effects which lead to plasma instabilities causing a loss of plasma confinement. Second, the high plasma density will result in the plasma behaving much more Re a conventional fluid with a mean free path shorter than the length of the device. This characteristic helps reduce problems associated with "loss cone" microinstabilities. An experimental GDM device is currently being constructed at the NASA Marshall Space Flight Center to provide an initial assessment of the feasibility of this type of propulsion system. Initial experiments are expected to commence in the late fall of 2000.

Prediction and realization of ITER-like pedestal pressure in the high- B tokamak Alcator C-Mod

NASA Astrophysics Data System (ADS)

Hughes, Jerry

2017-10-01

Fusion power in a burning plasma will scale as the square of the plasma pressure, which is increased in a straightforward way by increasing magnetic field: Pfus p2 B4 . Experiments on Alcator C-Mod, a compact high- B tokamak, have tested predictive capability for pedestal pressure, at toroidal field BT up to 8T , and poloidal field BP up to 1T . These reactor-like fields enable C-Mod to approach an ITER predicted value of 90kPa . This is expected if, as in the EPED model, the pedestal is constrained by onset of kinetic ballooning modes (KBMs) and peeling-ballooning modes (PMB), yielding a pressure pedestal approximately as pped BT ×BP . One successful path to high confinement on C-Mod is the high-density (ne > 3 ×1020m-3) approach, pursued using enhanced D-alpha (EDAs) H-mode. In EDA H-mode, transport regulates both the pedestal profiles and the core impurity content, holding the pedestal stationary, at just below the PBM stability boundary. We have extended this stationary ELM-suppressed regime to the highest magnetic fields achievable on C-Mod, and used it to approach the maximum pedestal predicted by EPED at high density: pped 60kPa . Another approach to high pressure utilizes a pedestal limited by PBMs at low collisionality, where pressure increases with density and EPED predicts access to a higher ``Super H'' solution for pped. Experiments at reduced density (ne < 2 ×1020m-3) and strong plasma shaping (δ > 0.5) accessed these regimes on C-Mod, producing pedestals with world record pped 80kPa , at Tped 2keV . In both the high and low density approaches, the impact of the pedestal on core performance is substantial. Our exploration of high pedestal regimes yielded a volume-averaged pressure 〈 p 〉 > 2atm , a world record value for a magnetic fusion device. The results hold promise for the projection of pedestal pressure and overall performance of high field burning plasma devices. Supported by U.S. Department of Energy awards DE-FC02-99ER54512, DE-FG02-95ER54309, DE-FC02-06ER54873, DE-AC02-09CH11466, DE-SC0007880 using Alcator C-Mod, a DOE Office of Science User Facility.

Tokamak with in situ magnetohydrodynamic generation of toroidal magnetic field

DOEpatents

Schaffer, Michael J.

1986-01-01

A tokamak apparatus includes an electrically conductive metal pressure vessel for defining a chamber and confining liquid therein. A liner disposed within said chamber defines a toroidal space within the liner and confines gas therein. The metal vessel provides an electrically conductive path linking the toroidal space. Liquid metal is forced outwardly through the chamber outside of the toroidal space to generate electric current in the conductive path and thereby generate a toroidal magnetic field within the toroidal space. Toroidal plasma is developed within the toroidal space about the major axis thereof.

77 FR 35112 - Petition To Modify an Exemption of a Previously Approved Antitheft Device; Ford Motor Company

Federal Register 2010, 2011, 2012, 2013, 2014

2012-06-12

... stated that the Fusion Titanium trim and Fusion Hybrid vehicles will be equipped with the IAwPB system as... Standard for the Ford Fusion vehicle line beginning with its model year (MY) 2012 vehicles. On February 16, 2012, Ford submitted a petition to modify its previously approved exemption for the Ford Fusion vehicle...

DIII-D research advancing the scientific basis for burning plasmas and fusion energy

NASA Astrophysics Data System (ADS)

W. M. SolomonThe DIII-D Team

2017-10-01

The DIII-D tokamak has addressed key issues to advance the physics basis for ITER and future steady-state fusion devices. In work related to transient control, magnetic probing is used to identify a decrease in ideal stability, providing a basis for active instability sensing. Improved understanding of 3D interactions is emerging, with RMP-ELM suppression correlated with exciting an edge current driven mode. Should rapid plasma termination be necessary, shattered neon pellet injection has been shown to be tunable to adjust radiation and current quench rate. For predictive simulations, reduced transport models such as TGLF have reproduced changes in confinement associated with electron heating. A new wide-pedestal variant of QH-mode has been discovered where increased edge transport is found to allow higher pedestal pressure. New dimensionless scaling experiments suggest an intrinsic torque comparable to the beam-driven torque on ITER. In steady-state-related research, complete ELM suppression has been achieved that is relatively insensitive to q 95, having a weak effect on the pedestal. Both high-q min and hybrid steady-state plasmas have avoided fast ion instabilities and achieved increased performance by control of the fast ion pressure gradient and magnetic shear, and use of external control tools such as ECH. In the boundary, experiments have demonstrated the impact of E× B drifts on divertor detachment and divertor asymmetries. Measurements in helium plasmas have found that the radiation shortfall can be eliminated provided the density near the X-point is used as a constraint in the modeling. Experiments conducted with toroidal rings of tungsten in the divertor have indicated that control of the strike-point flux is important for limiting the core contamination. Future improvements are planned to the facility to advance physics issues related to the boundary, transients and high performance steady-state operation.

DIII-D research advancing the scientific basis for burning plasmas and fusion energy

DOE PAGES

Solomon, Wayne M.

2017-07-12

The DIII-D tokamak has addressed key issues to advance the physics basis for ITER and future steady-state fusion devices. In work related to transient control, magnetic probing is used to identify a decrease in ideal stability, providing a basis for active instability sensing. Improved understanding of 3D interactions is emerging, with RMP-ELM suppression correlated with exciting an edge current driven mode. Should rapid plasma termination be necessary, shattered neon pellet injection has been shown to be tunable to adjust radiation and current quench rate. For predictive simulations, reduced transport models such as TGLF have reproduced changes in confinement associated withmore » electron heating. A new wide- pedestal variant of QH-mode has been discovered where increased edge transport is found to allow higher pedestal pressure. New dimensionless scaling experiments suggest an intrinsic torque comparable to the beam-driven torque on ITER. In steady-state-related research, complete ELM suppression has been achieved that is relatively insensitive to q 95, having a weak effect on the pedestal. Both high-q min and hybrid steady-state plasmas have avoided fast ion instabilities and achieved increased performance by control of the fast ion pressure gradient and magnetic shear, and use of external control tools such as ECH. In the boundary, experiments have demonstrated the impact of E × B drifts on divertor detachment and divertor asymmetries. Measurements in helium plasmas have found that the radiation shortfall can be eliminated provided the density near the X-point is used as a constraint in the modeling. Experiments conducted with toroidal rings of tungsten in the divertor have indicated that control of the strike-point flux is important for limiting the core contamination. In conclusion, future improvements are planned to the facility to advance physics issues related to the boundary, transients and high performance steady-state operation.« less

DIII-D research advancing the scientific basis for burning plasmas and fusion energy

DOE Office of Scientific and Technical Information (OSTI.GOV)

Solomon, Wayne M.

The DIII-D tokamak has addressed key issues to advance the physics basis for ITER and future steady-state fusion devices. In work related to transient control, magnetic probing is used to identify a decrease in ideal stability, providing a basis for active instability sensing. Improved understanding of 3D interactions is emerging, with RMP-ELM suppression correlated with exciting an edge current driven mode. Should rapid plasma termination be necessary, shattered neon pellet injection has been shown to be tunable to adjust radiation and current quench rate. For predictive simulations, reduced transport models such as TGLF have reproduced changes in confinement associated withmore » electron heating. A new wide- pedestal variant of QH-mode has been discovered where increased edge transport is found to allow higher pedestal pressure. New dimensionless scaling experiments suggest an intrinsic torque comparable to the beam-driven torque on ITER. In steady-state-related research, complete ELM suppression has been achieved that is relatively insensitive to q 95, having a weak effect on the pedestal. Both high-q min and hybrid steady-state plasmas have avoided fast ion instabilities and achieved increased performance by control of the fast ion pressure gradient and magnetic shear, and use of external control tools such as ECH. In the boundary, experiments have demonstrated the impact of E × B drifts on divertor detachment and divertor asymmetries. Measurements in helium plasmas have found that the radiation shortfall can be eliminated provided the density near the X-point is used as a constraint in the modeling. Experiments conducted with toroidal rings of tungsten in the divertor have indicated that control of the strike-point flux is important for limiting the core contamination. In conclusion, future improvements are planned to the facility to advance physics issues related to the boundary, transients and high performance steady-state operation.« less

Helical flow in RFX-mod tokamak plasmas

NASA Astrophysics Data System (ADS)

Piron, L.; Zaniol, B.; Bonfiglio, D.; Carraro, L.; Kirk, A.; Marrelli, L.; Martin, R.; Piron, C.; Piovesan, P.; Zuin, M.

2017-05-01

This work presents the first evidence of helical flow in RFX-mod q(a)  <  2 tokamak plasmas. The flow pattern is characterized by the presence of convective cells with m  =  1 and n  =  1 periodicity in the poloidal and toroidal directions, respectively. A similar helical flow deformation has been observed in the same device when operated as a reversed field pinch (RFP). In RFP plasmas, the flow dynamic is tailored by the innermost resonant m  =  1, n  =  7 tearing mode, which sustains the magnetic field configuration through the dynamo mechanism (Bonomo et al 2011 Nucl. Fusion 51 123007). By contrast, in the tokamak experiments presented here, it is strongly correlated with the m  =  1, n  =  1 MHD activity. A helical deformation of the flow pattern, associated with the deformation of the magnetic flux surfaces, is predicted by several codes, such as Specyl (Bonfiglio et al 2005 Phys. Rev. Lett. 94 145001), PIXIE3D (Chacón et al 2008 Phys. Plasmas 15 056103), NIMROD (King et al 2012 Phys. Plasmas 19 055905) and M3D-C1 (Jardin et al 2015 Phys. Rev. Lett. 115 215001). Among them, the 3D fully non-linear PIXIE3D has been used to calculate synthetic flow measurements, using a 2D flow modelling code. Inputs to the code are the PIXIE3D flow maps, the ion emission profiles as calculated by a 1D collisional radiative impurity transport code (Carraro et al 2000 Plasma Phys. Control. Fusion 42 731) and a synthetic diagnostic with the same geometry installed in RFX-mod. Good agreement between the synthetic and the experimental flow behaviour has been obtained, confirming that the flow oscillations observed with the associated convective cells are a signature of helical flow.

Confined ion energy >200 keV and increased fusion yield in a DPF with monolithic tungsten electrodes and pre-ionization

NASA Astrophysics Data System (ADS)

Lerner, Eric J.; Hassan, Syed M.; Karamitsos, Ivana; Von Roessel, Fred

2017-10-01

To reduce impurities in the dense plasma focus FF-1 device, we used monolithic tungsten electrodes with pre-ionization. With this new set-up, we demonstrated a three-fold reduction of impurities by mass and a ten-fold reduction by ion number. FF-1 produced a 50% increase in fusion yield over our previous copper electrodes, both for a single shot and for a mean of ten consecutive shots with the same conditions. These results represent a doubling of fusion yield as compared with any other plasma focus device with the same 60 kJ energy input. In addition, FF-1 produced a new single-shot record of 240 ± 20 keV for mean ion energy, a record for any confined fusion plasma, using any device, and a 50% improvement in ten-shot mean ion energy. With a deuterium-nitrogen mix and corona-discharge pre-ionization, we were also able to reduce the standard deviation in the fusion yield to about 15%, a four-fold reduction over the copper-electrode results. We intend to further reduce impurities with new experiments using microwave treatment of tungsten electrodes, followed by the use of beryllium electrodes.

NIMROD modeling of quiescent H-mode: reconstruction considerations and saturation mechanism

NASA Astrophysics Data System (ADS)

King, J. R.; Burrell, K. H.; Garofalo, A. M.; Groebner, R. J.; Kruger, S. E.; Pankin, A. Y.; Snyder, P. B.

2017-02-01

The extended-MHD NIMROD code (Sovinec and King 2010 J. Comput. Phys. 229 5803) models broadband-MHD activity from a reconstruction of a quiescent H-mode shot on the DIII-D tokamak (Luxon 2002 Nucl. Fusion 42 614). Computations with the reconstructed toroidal and poloidal ion flows exhibit low-{{n}φ} perturbations ({{n}φ}≃ 1 -5) that grow and saturate into a turbulent-like MHD state. The workflow used to project the reconstructed state onto the NIMROD basis functions re-solves the Grad-Shafranov equation and extrapolates profiles to include scrape-off-layer currents. Evaluation of the transport from the turbulent-like MHD state leads to a relaxation of the density and temperature profiles.

Study of Plasma Liner Driven Magnetized Target Fusion Via Advanced Simulations

DOE Office of Scientific and Technical Information (OSTI.GOV)

Samulyak, Roman V.; Brookhaven National Lab.; Parks, Paul

The feasibility of the plasma liner driven Magnetized Target Fusion (MTF) via terascale numerical simulations will be assessed. In the MTF concept, a plasma liner, formed by merging of a number (60 or more) of radial, highly supersonic plasma jets, implodes on the target in the form of two compact plasma toroids, and compresses it to conditions of the fusion ignition. By avoiding major difficulties associated with both the traditional laser driven inertial confinement fusion and solid liner driven MTF, the plasma liner driven MTF potentially provides a low-cost and fast R&D path towards the demonstration of practical fusion energy.more » High fidelity numerical simulations of full nonlinear models associated with the plasma liner MTF using state-of-art numerical algorithms and terascale computing are necessary in order to resolve uncertainties and provide guidance for future experiments. At Stony Brook University, we have developed unique computational capabilities that ideally suite the MTF problem. The FronTier code, developed in collaboration with BNL and LANL under DOE funding including SciDAC for the simulation of 3D multi-material hydro and MHD flows, has beenbenchmarked and used for fundamental and engineering problems in energy science applications. We have performed 3D simulations of converging supersonic plasma jets, their merger and the formation of the plasma liner, and a study of the corresponding oblique shock problem. We have studied the implosion of the plasma liner on the magnetized plasma target by resolving Rayleigh-Taylor instabilities in 2D and 3D and other relevant physics and estimate thermodynamic conditions of the target at the moment of maximum compression and the hydrodynamic efficiency of the method.« less

Fusion energy division annual progress report, period ending December 31, 1980

DOE Office of Scientific and Technical Information (OSTI.GOV)

Not Available

1981-11-01

The ORNL Program encompasses most aspects of magnetic fusion research including research on two magnetic confinement programs (tokamaks and ELMO bumpy tori); the development of the essential technologies for plasma heating, fueling, superconducting magnets, and materials; the development of diagnostics; the development of atomic physics and radiation effect data bases; the assessment of the environmental impact of magnetic fusion; the physics and engineering of present-generation devices; and the design of future devices. The integration of all of these activities into one program is a major factor in the success of each activity. An excellent example of this integration is themore » extremely successful application of neutral injection heating systems developed at ORNL to tokamaks both in the Fusion Energy Division and at Princeton Plasma Physics Laboratory (PPPL). The goal of the ORNL Fusion Program is to maintain this balance between plasma confinement, technology, and engineering activities.« less

BOOK REVIEW: Stellarator and Heliotron Devices

NASA Astrophysics Data System (ADS)

Johnson, John L.

1999-02-01

Stellarators and tokamaks are the most advanced devices that have been developed for magnetic fusion applications. The two approaches have much in common; tokamaks have received the most attention because their axisymmetry justifies the use of simpler models and provides a more forgiving geometry. However, recent advances in treating more complicated three dimensional systems have made it possible to design stellarators that are not susceptible to disruptions and do not need plasma current control. This has excited interest recently. The two largest new magnetic experiments in the world are the LHD device, which commenced operation in Toki, Japan, in 1998 and W7-X, which should become operational in Greifswald, Germany, in 2004. Other recently commissioned stellarators, including H-1 in Canberra, Australia, TJ-II in Madrid, Spain, and IMS in Madison, Wisconsin, have joined these in rejuvenating the stellarator programme. Thus, it is most appropriate that the author has made the lecture material that he presents to his students in the Graduate School of Energy Science at Kyoto University available to everyone. Stellarator and Heliotron Devices provides an excellent treatment of stellarator theory. It is aimed at graduate students who have a good understanding of classical mechanics and mathematical techniques. It contains good descriptions and derivations of essentially every aspect of fusion theory. The author provides an excellent qualitative introduction to each subject, pointing out the strengths and weaknesses of the models that are being used and describing our present understanding. He judiciously uses simple models which illustrate the similarities and differences between stellarators and tokamaks. To some extent the treatment is uneven, rigorous derivations starting with basic principles being given in some cases and relations and equations taken from the original papers being used as a starting point in others. This technique provides an excellent training ground for students without detracting from the usefulness of the book for knowledgeable fusion physicists. After a short, somewhat historical, introduction, Chapter 2 contains a good treatment of the basic properties of a toroidal magnetic configuration (the concepts of magnetic surfaces, rotational transform, shear and magnetic wells), averaging techniques which can often be used to simplify the calculations, helically invariant configurations, magnetic islands and line tracing techniques. Derivations and discussions of the basic tools of plasma theory, including the Vlasov equation, magnetohydrodynamic equations and their reduced form for low-β, large aspect ratio systems, properties of MHD waves, the drift kinetic equation and transport equations, are given in Chapter 3. Chapter 4 contains a good treatment of MHD equilibria, including a derivation of the three dimensional Grad-Shafranov equation, a discussion of the calculation of equilibria with a planar magnetic axis with both averaged equations and a variational approach, a comparison of the results of the two techniques, a formulation for stellarators with a helical magnetic axis and a good discussion of the Pfirsch-Schlüter current. The treatment of MHD instabilities in Chapter 5 is also excellent. It starts with a good derivation and discussion of the energy principle, gives a detailed treatment of ballooning modes where the wavelengths of the perturbation perpendicular to the field are short while those along B are long and derives the Mercier criterion from the ballooning mode equation. I personally prefer to obtain this criterion by making the low mode number assumption that dξ/dΨ>>dξ/dθ approx dξ/dζ, since non-ideal effects such as finite gyration radius corrections may provide less stabilization to these modes. A careful treatment of the resistive interchange mode is followed by a discussion of the role of localized stability criteria in the analysis of experiment and design studies, a study of Pfirsch-Schlüter current driven magnetic islands and the interpretation of sawtooth instabilities in Heliotron E. The treatment of particle orbits in Chapter 6 includes a derivation of drift equations, a discussion of the characteristics of trapped particle confinement in a heliotron and one of the Monte Carlo method for studying transport phenomena. A good treatment of neoclassical transport in a stellarator, with emphasis on the relation between parallel viscosity driven fluxes and bootstrap current, is given in Chapter 7. This is the best treatment I have found, outside of the original references, but it is still demanding. In addition, a radial electric field is introduced into the energy transport equations. The treatment of heating and confinement of heliotron plasmas in Chapter 8 is a good combination of providing results from experiments on the Heliotron E and DR heliotrons and the ATF and CHS stellarators and showing how theoretical interpretation is formulated. The discussions of ray tracing and energy absorption for both ECRH and ICRF heating techniques, as well as a treatment of neutral beam injection, are very clear. Measurements of bootstrap current and plasma rotation, as well as the density limits associated with pellet injection, are discussed. The chapter ends with a discussion of what may be the author's favourite topic, pressure gradient driven turbulence, in which he describes mixing length and scale invariance techniques. Finally, a discussion of the characteristics of a steady state fusion reactor, including a treatment of the containment, slowing down and energy transfer of the alpha particles, one of the toroidal Alfvén modes driven by these particles and some physics of divertors are given in Chapter 9. A reviewer is usually expected to find some faults. I had no problem in finding one as soon as I received the book: indeed, I did not like its title. I have always maintained that Lyman Spitzer defined a stellarator as any toroidal device in which the rotational transform is generated by coils outside the plasma, either through imposition of a helical magnetic axis as in a figure-8 stellarator or a heliac, or through the generation of helical magnetic fields, as in a classical stellarator, a torsatron or a quasi-helical stellarator such as W7-X. The author notes that the heliotron (as it was invented by Uo in Japan) is the same as the torsatron (first proposed by Gourdon and his colleagues in Europe) in his introduction, but cannot bring himself to ignore Uo's desire to maintain a distinction between stellarators and heliotrons. Enough typographical errors are present to make one have to be careful before relying on the book for specific formulas. Nevertheless, it will prove to be a useful reference. I have always respected the author for the quality of students he produces. He provides a list of some of them in the preface, which justifies this opinion. These students are a good demonstration of the usefulness of this book.

Health physics measurement of Princeton Tokamaks, 1977-1987.

PubMed

Stencel, J R; Gilbert, J D; Couch, J G; Griesbach, O A; Fennimore, J J; Greco, J M

1989-06-01

The Princeton Plasma Physics Laboratory (PPPL) began fusion experiments in 1951. In the early years, the major health physics concerns were associated with x radiation produced by energetic electrons in the plasma. Within the past year, neutron and 3H production from 2H-2H (represented hereafter as D-D) reactions has increased significantly on the larger fusion devices. Tritium retention noted in graphite tiles underscores the significance of material selection in present and future 3H-fueled fusion devices. This paper reports on operational health physics radiation measurements made on various PPPL machines over the past 10 y.

Study of the electromagnetic characteristics of multiple HTSPPT modules based on the configuration of toroidal structure for inductive pulsed power supply

NASA Astrophysics Data System (ADS)

Zhang, Cunshan; Zheng, Xinxin; Li, Haitao; Li, Zhenmei; Zhang, Tao; Jiao, Can

2018-04-01

High temperature superconducting pulsed power transformer (HTSPPT) is an important device for pulsed power supplies. It consists of a superconducting primary and a normal conducting secondary, which is used for energy storage and current amplification. The critical current density, the energy storage, and the coupling coefficient are three main performance indexes. They are affected by the geometry parameters of HTSPPT modules, such as the height and the width of the superconducting coils. In addition, the hoop stress of the HTSPPT coils is limited by the maximum tensile strength of high temperature superconducting (HTS) tapes. In this paper, Bi-2223/Ag HTS tapes are selected as the wire of primary inductor and the toroidal structure model is selected for multiple HTSPPT modules. The relationships between the geometry parameters of HTSPPT modules and the electrical performance are studied.

Control of magnetohydrodynamic stability by phase space engineering of energetic ions in tokamak plasmas.

PubMed

Graves, J P; Chapman, I T; Coda, S; Lennholm, M; Albergante, M; Jucker, M

2012-01-10

Virtually collisionless magnetic mirror-trapped energetic ion populations often partially stabilize internally driven magnetohydrodynamic disturbances in the magnetosphere and in toroidal laboratory plasma devices such as the tokamak. This results in less frequent but dangerously enlarged plasma reorganization. Unique to the toroidal magnetic configuration are confined 'circulating' energetic particles that are not mirror trapped. Here we show that a newly discovered effect from hybrid kinetic-magnetohydrodynamic theory has been exploited in sophisticated phase space engineering techniques for controlling stability in the tokamak. These theoretical predictions have been confirmed, and the technique successfully applied in the Joint European Torus. Manipulation of auxiliary ion heating systems can create an asymmetry in the distribution of energetic circulating ions in the velocity orientated along magnetic field lines. We show the first experiments in which large sawtooth collapses have been controlled by this technique, and neoclassical tearing modes avoided, in high-performance reactor-relevant plasmas.

KTX circuit model and discharge waveform prediction

NASA Astrophysics Data System (ADS)

Bai, Wei; Lan, T.; Mao, W. Z.; You, W.; Li, H.; Liu, A. D.; Xie, J. L.; Wan, S. D.; Liu, W. D.; Yang, L.; Fu, P.; Xiao, C. J.; Ding, W. X.

2013-10-01

The Keda Torus eXperiment (KTX) is a constructing reversed field pinch (RFP) device in University of Science and Technology of China. The KTX power supply system includes the Ohmic heating, field shaping and toroidal power supply systems, which produce the Ohmic field, equilibrium field and toroidal field, respectively. The detailed circuit model will be introduced in this poster. Another purpose is to predict its discharge waveforms using the modified Bessel function mode (MBFM), which describes the evolution of plasma current and magnetic flux in RFP base on Taylor theory. Furthermore, the power supply requirements of external field shaping winding are also predicted in the model, which will be very helpful for the design of plasma equilibrium controlling system. Supported by ITER-China program (No. 2011GB106000), NNSFC (Nos. 10990210, 10990211, 10335060 and 10905057), CPSF (No. 20080440104), YIF (No. WK2030040019) and KIPCAS (No. kjcx-yw-n28).

Characteristics of Low-q(a) Disruptions in the Compact Toroidal Hybrid

NASA Astrophysics Data System (ADS)

Pandya, M. D.; Archmiller, M. C.; Ennis, D. A.; Hartwell, G. J.; Maurer, D. A.

2014-10-01

Tokamak disruptions are dramatic events that lead to a sudden loss of plasma confinement. Disruptions that occur at low edge safety-factor, q (a) , limit the operation of tokamaks to q (a) >= 2 . The Compact Toroidal Hybrid (CTH) is a torsatron-tokamak hybrid with a helical field coil and vertical field coils to establish a stellartor equilibrium, while an ohmic coil induces plasma current. A feature of the CTH device is the ability to adjust the vacuum rotational transform, tvac (t =1/q ), by varying the ratio of current in the helical and toroidal field coils. The value of edge tvac can be varied from about 0.02 to 0.3 (qvac (a) ~ 50 to 3.3). Plasma discharges in CTH are routinely observed to operate with q (a) < 2 , and in some cases as low as q (a) ~ 1 . 1 . In CTH, low-q(a) disruptions are observed with a dominant m/n=3/2 precursor. The disruptivity of plasma discharges is over 80% when tvac (a) < 0 . 04 (qvac (a) < 25) and as tvac (a) is increased further, the disruptivity of the plasma discharges decreases. The disruptions are completely suppressed for tvac (a) > 0 . 07 (qvac (a) ~ 14) . This work is supported by US Department of Energy Grant No. DE-FG02-00ER54610.

The effect of a sheared flow on magnetic islands in plasmas with non-axisymetric geometry

NASA Astrophysics Data System (ADS)

Cancino, M. Stefany; Martinell, Julio J.

2018-02-01

The stability of a magnetic island in a toroidal magnetic confinement device depends on various factors besides the usual tearing-mode stability parameter ?, determined by the local current profile.The presence of a sheared flow in the vicinity of the rational surface that supports the island is one of the factors that affects its stability since it can give rise to a polarization current around the island position. The contribution of the polarization current to the stability has been computed for a tokamak geometry. Here, we consider the case of magnetic islands with a shear flow in a stellarator which has a non-axisymmetric magnetic geometry. The main difference is a contribution to the polarization current from the toroidal electrostatic oscillation. A correction due to the global toroidal magnetic geometry is also present. It is found that the regime where the stability is affected corresponds to the large island width relative to the ion gyroradius. Thus, the contribution is relevant for low-temperature regimes. In that case, the polarization current is destabilizing for frequencies larger than the ion diamagnetic frequency. Our results imply that the sheared flow can produce a growth of the magnetic island in a cold plasma but it can become narrower as the temperature rises.

Biomechanical evaluation of CIBOR spine interbody fusion device.

PubMed

Chong, Alexander C M; Harrer, Seth W; Heggeness, Michael H; Wooley, Paul H

2017-07-01

The CIBOR PEEK spinal interbody fusion device is an anterior lumbar interbody fusion construct with a hollow center designed to accommodate an osteoinductive carbon foam insert to promote bony ingrowth to induce fusion where rigid stabilization is needed. Three different sizes of the device were investigated. Part-I: implants were tested under axial compression and rotation using polyurethane foam blocks. Part-II: simulated 2-legged stance using cadaveric specimen using the L5-S1 lumbar spine segment. Part-III: a survey feedback form was used to investigate two orthopedic surgeons concern regarding the implant. In Part-I, the subsidence hysteresis under axial compression loading was found to be statistical significant difference between these three implant sizes. It was noted that the implants had migration as rotation applied, and the amount of subsidence was a factor of the axial compression loads applied. In Part-II, a minor subsidence and carbon foam debris were observed when compared to each implant size. Poor contact surface of the implant with the end plates of the L5 or S1 vertebrae from the anterior view under maximum loads was observed; however, the implant seemed to be stable. Each surgeon has their own subjective opinion about the CIBOR implant. Two out of the three different sizes of the device (medium and large sizes) provided appropriate rigid stabilization at the physiological loads. Neither orthopedic surgeon was 100% satisfied with overall performance of the implant, but felt potential improvement could be made. This study indicates an option for operative treatment of spine interbody fusion, as the CIBOR spine interbody fusion device has a hollow center. This hollow center is designed to accommodate a carbon foam insert to promote bony ingrowth. © 2016 Wiley Periodicals, Inc. J Biomed Mater Res Part B: Appl Biomater, 105B: 1157-1168, 2017. © 2016 Wiley Periodicals, Inc.

Dense Plasma Focus - From Alternative Fusion Source to Versatile High Energy Density Plasma Source for Plasma Nanotechnology

NASA Astrophysics Data System (ADS)

Rawat, R. S.

2015-03-01

The dense plasma focus (DPF), a coaxial plasma gun, utilizes pulsed high current electrical discharge to heat and compress the plasma to very high density and temperature with energy densities in the range of 1-10 × 1010 J/m3. The DPF device has always been in the company of several alternative magnetic fusion devices as it produces intense fusion neutrons. Several experiments conducted on many different DPF devices ranging over several order of storage energy have demonstrated that at higher storage energy the neutron production does not follow I4 scaling laws and deteriorate significantly raising concern about the device's capability and relevance for fusion energy. On the other hand, the high energy density pinch plasma in DPF device makes it a multiple radiation source of ions, electron, soft and hard x-rays, and neutrons, making it useful for several applications in many different fields such as lithography, radiography, imaging, activation analysis, radioisotopes production etc. Being a source of hot dense plasma, strong shockwave, intense energetic beams and radiation, etc, the DPF device, additionally, shows tremendous potential for applications in plasma nanoscience and plasma nanotechnology. In the present paper, the key features of plasma focus device are critically discussed to understand the novelties and opportunities that this device offers in processing and synthesis of nanophase materials using, both, the top-down and bottom-up approach. The results of recent key experimental investigations performed on (i) the processing and modification of bulk target substrates for phase change, surface reconstruction and nanostructurization, (ii) the nanostructurization of PLD grown magnetic thin films, and (iii) direct synthesis of nanostructured (nanowire, nanosheets and nanoflowers) materials using anode target material ablation, ablated plasma and background reactive gas based synthesis and purely gas phase synthesis of various different types of nanostructured materials using DPF device will discussed to establish this device as versatile tool for plasma nanotechnology.

Modeling an unmitigated thermal quench event in a large field magnet in a DEMO reactor

DOE PAGES

Merrill, Brad J.

2015-03-25

The superconducting magnet systems of future fusion reactors, such as a Demonstration Power Plant (DEMO), will produce magnetic field energies in the 10 s of GJ range. The release of this energy during a fault condition could produce arcs that can damage the magnets of these systems. The public safety consequences of such events must be explored for a DEMO reactor because the magnets are located near the DEMO's primary radioactive confinement barrier, the reactor's vacuum vessel (VV). Great care will be taken in the design of DEMO's magnet systems to detect and provide a rapid field energy dump tomore » avoid any accidents conditions. During an event when a fault condition proceeds undetected, the potential of producing melting of the magnet exists. If molten material from the magnet impinges on the walls of the VV, these walls could fail, resulting in a pathway for release of radioactive material from the VV. A model is under development at Idaho National Laboratory (INL) called MAGARC to investigate the consequences of this accident in a large toroidal field (TF) coil. Recent improvements to this model are described in this paper, along with predictions for a DEMO relevant event in a toroidal field magnet.« less

Theory comparison and numerical benchmarking on neoclassical toroidal viscosity torque

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wang, Zhirui; Park, Jong-Kyu; Logan, Nikolas

Systematic comparison and numerical benchmarking have been successfully carried out among three different approaches of neoclassical toroidal viscosity (NTV) theory and the corresponding codes: IPEC-PENT is developed based on the combined NTV theory but without geometric simplifications [Park et al., Phys. Rev. Lett. 102, 065002 (2009)]; MARS-Q includes smoothly connected NTV formula [Shaing et al., Nucl. Fusion 50, 025022 (2010)] based on Shaing's analytic formulation in various collisionality regimes; MARS-K, originally computing the drift kinetic energy, is upgraded to compute the NTV torque based on the equivalence between drift kinetic energy and NTV torque [J.-K. Park, Phys. Plasma 18, 110702more » (2011)]. The derivation and numerical results both indicate that the imaginary part of drift kinetic energy computed by MARS-K is equivalent to the NTV torque in IPEC-PENT. In the benchmark of precession resonance between MARS-Q and MARS-K/IPEC-PENT, the agreement and correlation between the connected NTV formula and the combined NTV theory in different collisionality regimes are shown for the first time. Additionally, both IPEC-PENT and MARS-K indicate the importance of the bounce harmonic resonance which can greatly enhance the NTV torque when E×B drift frequency reaches the bounce resonance condition.« less

Deuterium charge exchange recombination spectroscopy from the top of the pedestal to the scrape off layer in H-mode plasmas

NASA Astrophysics Data System (ADS)

Haskey, S. R.; Grierson, B. A.; Stagner, L.; Burrell, K. H.; Chrystal, C.; Groebner, R. J.; Ashourvan, A.; Pablant, N. A.

2017-10-01

Recent completion of the thirty two channel main-ion (deuterium) charge exchange recombination spectroscopy (CER) diagnostic on DIII-D [J.L. Luxon, Nucl. Fusion 42 (2002) 614] enables detailed comparisons between impurity and main-ion temperature, density, and toroidal rotation. Sixteen sightlines cover the core of the plasma and another sixteen are densely packed towards the edge, providing high resolution measurements of the pedestal and steep gradient edge region of H-mode plasmas. The complexities of the Dα spectrum require fitting with a comprehensive model, as well as using iterative collisional radiative modeling to determine the underlying thermal deuterium ion properties. Large differences in the structure and magnitude of impurity (C6+) and main-ion (D+) toroidal rotation profiles are seen in the H-mode pedestal. Additionally the D+ temperature can be half the value of the C6+ temperature at the separatrix and shows more of a pedestal structure. Typically only the impurity properties are measured and the main-ion properties are either assumed to be the same, or inferred using neoclassical models, which require validation in the steep gradient region. These measured differences have implications for transport model validation, intrinsic rotation studies, pedestal stability, and the boundary conditions for scrape off layer and plasma material interactions studies.

Plasma wall interaction, a key issue on the way to a steady state burning fusion device

NASA Astrophysics Data System (ADS)

Philipps, V.

2006-04-01

The International Tokamak Experimental Reactor (ITER), the first burning fusion plasma experiment based on the tokamak principle, is ready for construction. It is based on many years of fusion research resulting in a robust design in most of the areas. Present day fusion research concentrates on the remaining critical issues which are, to a large extent, connected with processes of plasma wall interaction. This is mainly due to extended duty cycle and the increase of the plasma stored energy in comparison with present-day machines. Critical topics are the lifetime of the plasma facing components (PFC) and the long-term tritium retention. These processes are controlled mainly by material erosion, both during steady state operation and transient power losses (disruptions and edge localized modes (ELMs)) and short- and long-range material migration and re-deposition. The extrapolation from present-day 'full carbon wall' devices suggests that the long-term tritium retention in a burning fusion device would be unacceptably high under these conditions allowing for only an unacceptable limited number of pulses in a D T mixture. As a consequence of this, research activities have been strengthened to understand in more detail the underlying processes of material erosion and re-deposition, to develop methods to remove retained tritium from the PFCs and remote areas of a fusion device and to explore these processes and the plasma performance in more detail with metallic PFC, such as beryllium (Be) and tungsten (W), which are foreseen for the ITER experiment. This paper outlines the main physical mechanisms leading to material erosion, migration and re-deposition and the associated fuel retention. It addresses the experimental database in these areas and describes the further research strategies that will be needed to tackle critical issues.

Predictions of H-mode performance in ITER

NASA Astrophysics Data System (ADS)

Budny, Robert

2008-11-01

Time-dependent integrated predictions of performance metrics such as the fusion power PDT, QDT≡ PDT/Pext, and alpha profiles are presented. The PTRANSP [1] code is used, along with GLF23 to predict plasma profiles, NUBEAM for NNBI and alpha heating, TORIC for ICRH, and TORAY for ECRH. Effects of sawteeth mixing, beam steering, beam shine-through, radiation loss, ash accumulation, and toroidal rotation are included. A total heating of Pext=73MW is assumed to achieve H-mode during the density and current ramp-up phase. Various mixes of NNBI, ICRH, and ECRH heating schemes are compared. After steady state conditions are achieved, Pext is stepped down to lower values to explore high QDT. Physics and computation uncertainties lead to ranges in predictions for PDT and QDT. Physics uncertainties include the L->H and H->L threshold powers, pedestal height, impurity and ash transport, and recycling. There are considerably more uncertainties predicting the peak value for QDT than for PDT. [0pt] [1] R.V. Budny, R. Andre, G. Bateman, F. Halpern, C.E. Kessel, A. Kritz, and D. McCune, Nuclear Fusion 48 (2008) 075005.

Tokamak Operation with Safety Factor q 95 < 2 via Control of MHD Stability

DOE PAGES

Piovesan, Paolo; Hanson, Jeremy M.; Martin, Piero; ...

2014-07-24

Magnetic feedback control of the resistive-wall mode has enabled DIII-D to access stable operation at safety factor q95 = 1:9 in divertor plasmas for 150 instability growth times. Magnetohydrodynamic stability sets a hard, disruptive limit on the minimum edge safety factor achievable in a tokamak, or on the maximum plasma current at given toroidal magnetic eld. In tokamaks with a divertor, the limit occurs at q95 = 2, as con rmed in DIII-D. Since the energy con cement time scales linearly with current, this also bounds the performance of a fusion reactor. DIII-D has overcome this limit, opening a wholemore » new high-current regime not accessible before. This result brings signi cant possible bene ts in terms of fusion performance, but it also extends resistive wall mode physics and its control to conditions never explored before. In present experiments, q95 < 2 operation is eventually halted by voltage limits reached in the feedback power supplies, not by intrinsic physics issues. Improvements to power supplies and to control algorithms have the potential to further extend this regime.« less

Effects of compressional magnetic perturbation on kinetic Alfven waves

NASA Astrophysics Data System (ADS)

Dong, Ge; Bhattacharjee, Amitava; Lin, Zhihong

2016-10-01

Kinetic Alfven waves play a very important role in the dynamics of fusion as well as space and astrophysical plasmas. The compressional magnetic perturbation δB|| can play important role in kinetic Alfven waves (KAW) and various instabilities at large plasma β. It could affect the nonlinear behavior of these modes significantly even at small β. In this study, we have implemented δB|| in gyrokinetic toroidal code (GTC). The perpendicular Ampere's law is solved as a force balance equation. Double gyroaveraging is incorporated in the code to treat the finite Larmor radius effects related to δB|| terms. KAW is studied in slab geometry as a benchmark case. A scan in β for the KAW dispersion relation shows that as β approaches 1 (>0.3), the effects of δB|| becomes important. Connections are made with other existing studies of KAWs in the fusion and space plasma literature. This new capability of including δB|| in GTC could be applied to nonlinear simulations of modes such as kinetic ballooning and tearing modes. This research is supported by DOE Contract No. DE-AC02-09CH11466.

Interpretation of frequency sweeping of n=0 mode in JET

NASA Astrophysics Data System (ADS)

Berk, H. L.

2006-04-01

Persistent rapid up and down frequency chirping modes with a toroidal mode number of zero (n=0) are observed in the JET tokamak when energetic ions, in the range of several hundred keV, are created by high field side ion cyclotron resonance frequency heating. Fokker-Planck calculations demonstrate that the heating method enables the formation of an energetically inverted ion distribution which supplies the free energy for the ions to excite a global geodesic acoustic mode (GGAM). The large frequency shifts of this mode are attributed to the formation of phase space structures whose frequencies, which are locked to an ion orbit resonance frequency, are forced to continually shift so that energetic particle energy can be released to counterbalance the energy dissipation present in the background plasma. In collaboration with C.J. Boswell, MIT; D. Borba, A.C.A. Figueiredo, Center for Nuclear Fusion Association; T. Johnson, Alfven Laboratory, KTH; M.F.F. Nave, Center for Nuclear Fusion Association; S.D. Pinches, Max Planck Institute for Plasma Physics; S.E. Sharapov, UKEA Culham Science Centre; and T. Zhou, University of Texas at Austin.

Theoretical issues on the spontaneous rotation of axisymmetric plasmas

NASA Astrophysics Data System (ADS)

Coppi, B.; Zhou, T.

2014-09-01

An extensive series of experiments have confirmed that the observed ‘spontaneous rotation’ phenomenon in axisymmetric plasmas is related to the confinement properties of these plasmas and connected to the excitation of collective modes associated with these properties (Coppi 2000 18th IAEA Fusion Energy Conf. (Sorrento, Italy, 2000) THP 1/17, www-pub.iaea.org/MTCD/publications/PDF/csp_008c/html/node343.htm and Coppi 2002 Nucl. Fusion 42 1). In particular, radially localized modes can extract angular momentum from the plasma column from which they grow while the background plasma has to recoil in the direction opposite to that of the mode phase velocity. In the case of the excitation of the plasma modes at the edge, the loss of their angular momentum can be connected to the directed particle ejection to the surrounding medium. The recoil angular momentum is then redistributed inside the plasma column mainly by the combination of an effective viscous diffusion and an inward angular momentum transport velocity that is connected, for instance, to ion temperature gradient (ITG) driven modes. The linear and quasi-linear theories of the collisionless trapped electron modes and of the toroidal ITG driven modes are re-examined in the context of their influence on angular momentum transport. Internal modes that produce magnetic reconnection and are electromagnetic in nature, acquire characteristic phase velocity directions in high temperature regimes and become relevant to the ‘generation’ of angular momentum. The drift-tearing mode, the ‘complex’ reconnecting mode and the m0 = 1 internal mode belong to this category, the last mode acquiring different features depending on the strength of its driving factor. Toroidal velocity profiles that reproduce the experimental observations are obtained considering a global angular momentum balance equation that includes the localized sources associated with the excited internal electrostatic and electromagnetic modes besides the appropriate diffusive and the inward angular momentum transparent terms.

Experiment to investigate current drive by fast Alfven waves in a small tokamak

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gahl, J.; Ishihara, O.; Wong, K.

1985-07-01

An experiment has been carried out to study current generation by Doppler shifted cyclotron resonance heating of minority ions with a unidirectional wave in the small tokamak at Texas Tech University. One of the objectives of the experiment is to understand in detail the wave-particle interactions through which fast (compressional) Alfven waves in the ion cyclotron range of frequencies drive currents in toroidal devices.

Efficient use of mobile devices for quantification of pressure injury images.

PubMed

Garcia-Zapirain, Begonya; Sierra-Sosa, Daniel; Ortiz, David; Isaza-Monsalve, Mariano; Elmaghraby, Adel

2018-01-01

Pressure Injuries are chronic wounds that are formed due to the constriction of the soft tissues against bone prominences. In order to assess these injuries, the medical personnel carry out the evaluation and diagnosis using visual methods and manual measurements, which can be inaccurate and may generate discomfort in the patients. By using segmentation techniques, the Pressure Injuries can be extracted from an image and accurately parameterized, leading to a correct diagnosis. In general, these techniques are based on the solution of differential equations and the involved numerical methods are demanding in terms of computational resources. In previous work, we proposed a technique developed using toroidal parametric equations for image decomposition and segmentation without solving differential equations. In this paper, we present the development of a mobile application useful for the non-contact assessment of Pressure Injuries based on the toroidal decomposition from images. The usage of this technique allows us to achieve an accurate segmentation almost 8 times faster than Active Contours without Edges (ACWE) and Dynamic Contours methods. We describe the techniques and the implementation for Android devices using Python and Kivy. This application allows for the segmentation and parameterization of injuries, obtain relevant information for the diagnosis and tracking the evolution of patient's injuries.

Initial results from the rebuilt EXTRAP T2R RFP device

NASA Astrophysics Data System (ADS)

Brunsell, P. R.; Bergsåker, H.; Cecconello, M.; Drake, J. R.; Gravestijn, R. M.; Hedqvist\\ad{2 }, A.; Malmberg, J.-A.

2001-11-01

The EXTRAP T2R thin shell reversed-field pinch (RFP) device has recently resumed operation after a major rebuild including the replacement of the graphite armour with molybdenum limiters, a fourfold increase of the shell time constant, and the replacement of the helical coil used for the toroidal field with a conventional solenoid-type coil. Wall-conditioning using hydrogen glow discharge cleaning was instrumental for successful RFP operation. Carbon was permanently removed from the walls during the first week of operation. The initial results from RFP operation with relatively low plasma currents in the range Ip = 70-100 kA are reported. RFP discharges are sustained for more than three shell times. Significant improvements in plasma parameters are observed, compared to operation before the rebuild. There is a substantial reduction in the carbon impurity level. The electron density behaviour is more shot-to-shot reproducible. The typical density is ne = 0.5-1×1019 m-3. Monitors of Hα line radiation indicate that the plasma wall interaction is more toroidally symmetric and that there is less transient gas release from the wall. The minimum loop voltage is in the range Vt = 28-35 V, corresponding to a reduction by a factor of two to three compared to the value before the rebuild.

A laser device for fusion of nasal mucosa

NASA Astrophysics Data System (ADS)

Sooklal, Valmiki; McClure, Jesse; Hooper, Luke; Larson, Michael

2010-02-01

A prototype device has been created to fuse septal tissue membranes as an alternative to sutures or staples through the controlled application of laser heating and pressure to induce protein denaturation and subsequent tissue fusion, through renaturation and intertwining, across the interface. Lasers have been used to close wounds in controlled laboratory tests over the last 15 years. Many encouraging results have been obtained; however, no commercial delivery systems are currently available. This is due primarily to two factors: requiring an inordinate amount of experience on the part of the operator, and attempting to achieve general applicability for multiple tissue systems. The present device overcomes these barriers as it is tailored for the particular application of septal laser fusion, namely for the coaptation of mucoperichondrial membranes. The important parameters involved in fusing biological tissues are identified. The development of the device followed from computational modeling based on Monte Carlo simulation of photon transport and on engineering firstprinciples. Experiments were designed and analyzed using orthogonal arrays, employing a subset of the relevant parameters, i.e., laser irradiance, dwell time and spot size, for a range of wavelengths. The in vitro fusion experiments employed 1cm by 1cm sections of equine nasal mucosa having a nominal thickness of 1mm.

Compact fusion energy based on the spherical tokamak

NASA Astrophysics Data System (ADS)

Sykes, A.; Costley, A. E.; Windsor, C. G.; Asunta, O.; Brittles, G.; Buxton, P.; Chuyanov, V.; Connor, J. W.; Gryaznevich, M. P.; Huang, B.; Hugill, J.; Kukushkin, A.; Kingham, D.; Langtry, A. V.; McNamara, S.; Morgan, J. G.; Noonan, P.; Ross, J. S. H.; Shevchenko, V.; Slade, R.; Smith, G.

2018-01-01

Tokamak Energy Ltd, UK, is developing spherical tokamaks using high temperature superconductor magnets as a possible route to fusion power using relatively small devices. We present an overview of the development programme including details of the enabling technologies, the key modelling methods and results, and the remaining challenges on the path to compact fusion.

Shuttleless toroid winder

DOEpatents

Lindenmeyer, Carl W.

1981-01-01

A lower support receives a toroid at a winding station with the axis of the toroid aligned with a slot in the support. An upper guide member applies an axial force to hold the toroid against the lower support. A pair of movable jaws carried by an indexing mechanism engage the outer surface of the toroid to apply a radial holding force. While the toroid is thus held, a wire is placed axially through the toroid, assisted by a funnel-shaped surface in the upper guide member, and is drawn tight about the toroid by a pair of cooperating draw rollers. When operated in the "full cycle" mode, the operator then actuates a switch which energizes a power drive to release the axial clamp and to drive the indexing mechanism and the jaws to rotate the toroid about its axis. At the same time, the wire is ejected from the draw rollers beneath the toroid so that the operator may grasp it to form another loop. When the toroid is fully indexed, the jaws release it, and the upper guide member is returned to clamp the toroid axially while the indexing mechanism is returned to its starting position. The apparatus may also be operated in a "momentary contact" mode in which the mechanism is driven only for the time a switch is actuated.

Development progresses of radio frequency ion source for neutral beam injector in fusion devices.

PubMed

Chang, D H; Jeong, S H; Kim, T S; Park, M; Lee, K W; In, S R

2014-02-01

A large-area RF (radio frequency)-driven ion source is being developed in Germany for the heating and current drive of an ITER device. Negative hydrogen ion sources are the major components of neutral beam injection systems in future large-scale fusion experiments such as ITER and DEMO. RF ion sources for the production of positive hydrogen (deuterium) ions have been successfully developed for the neutral beam heating systems at IPP (Max-Planck-Institute for Plasma Physics) in Germany. The first long-pulse ion source has been developed successfully with a magnetic bucket plasma generator including a filament heating structure for the first NBI system of the KSTAR tokamak. There is a development plan for an RF ion source at KAERI to extract the positive ions, which can be applied for the KSTAR NBI system and to extract the negative ions for future fusion devices such as the Fusion Neutron Source and Korea-DEMO. The characteristics of RF-driven plasmas and the uniformity of the plasma parameters in the test-RF ion source were investigated initially using an electrostatic probe.

A multicenter evaluation of comprehensive analysis of MLL translocations and fusion gene partners in acute leukemia using the MLL FusionChip device.

PubMed

Harrison, Christine J; Griffiths, Mike; Moorman, Fìona; Schnittger, Susanne; Cayuela, Jean-Michel; Shurtleff, Sheila; Gottardi, Enrico; Mitterbauer, Gerlinde; Colomer, Dolores; Delabesse, Eric; Castéras, Vincent; Maroc, Nicolas

2007-02-01

Rearrangements of the MLL gene are significant in acute leukemia. Among the most frequent translocations are t(4;11)(q21;q23) and t(9;11)(p22;q23), which give rise to the MLL-AFF1 and MLL-MLLT3 fusion genes (alias MLL-AF4 and MLL-AF9) in acute lymphoblastic and acute myeloid leukemia, respectively. Current evidence suggests that determining the MLL status of acute leukemia, including precise identification of the partner gene, is important in defining appropriate treatment. This underscores the need for accurate detection methods. A novel molecular diagnostic device, the MLL FusionChip, has been successfully used to identify MLL fusion gene translocations in acute leukemia, including the precise breakpoint location. This study evaluated the performance of the MLL FusionChip within a routine clinical environment, comprising nine centers worldwide, in the analysis of 21 control and 136 patient samples. It was shown that the assay allowed accurate detection of the MLL fusion gene, regardless of the breakpoint location, and confirmed that this multiplex approach was robust in a global multicenter trial. The MLL FusionChip was shown to be superior to other detection methods. The type of molecular information provided by MLL FusionChip gave an indication of the appropriate primers to design for disease monitoring of MLL patients following treatment.

Instrumented posterior lumbar interbody fusion (PLIF) with interbody fusion device (Cage) in degenerative disc disease (DDD): 3 years outcome.

PubMed

Ahsan, M K; Hossain, M A; Sakeb, N; Khan, S I; Zaman, N

2013-10-01

This prospective interventional study carried out at Bangabandhu Sheikh Mujib Medical University and a private hospital in Dhaka, Bangladesh during the period from October 2003 to September 2011. Surgical treatment of degenerative disc disease (DDD) should aim to re-expand the interbody space and stabilize until fusion is complete. The present study conducted to find out the efficacy of using interbody fusion device (Cage) to achieve interbody space re-expansion and fusion in surgical management of DDD. We have performed the interventional study on 53 patients, 42 female and 11 male, with age between 40 to 67 years. All the patients were followed up for 36 to 60 months (average 48 months). Forty seven patients were with spondylolisthesis and 06 with desiccated disc. All subjects were evaluated with regard to immediate and long term complications, radiological fusion and interbody space re-expansion and maintenance. The clinical outcome (pain and disability) was scored by standard pre and postoperative questionnaires. Intrusion, extrusion and migration of the interbody fusion cage were also assessed. Forty seven patients were considered to have satisfactory outcome in at least 36 months follow up. Pseudoarthrosis developed in 04 cases and 06 patients developed complications. In this series posterior lumbar interbody fusion (PLIF) with interbody cage and instrumentation in DDD showed significant fusion rate and maintenance of interbody space. Satisfactory outcome observed in 88.68% cases.

Steady State Advanced Tokamak (SSAT): The mission and the machine

NASA Astrophysics Data System (ADS)

Thomassen, K.; Goldston, R.; Nevins, B.; Neilson, H.; Shannon, T.; Montgomery, B.

1992-03-01

Extending the tokamak concept to the steady state regime and pursuing advances in tokamak physics are important and complementary steps for the magnetic fusion energy program. The required transition away from inductive current drive will provide exciting opportunities for advances in tokamak physics, as well as important impetus to drive advances in fusion technology. Recognizing this, the Fusion Policy Advisory Committee and the U.S. National Energy Strategy identified the development of steady state tokamak physics and technology, and improvements in the tokamak concept, as vital elements in the magnetic fusion energy development plan. Both called for the construction of a steady state tokamak facility to address these plan elements. Advances in physics that produce better confinement and higher pressure limits are required for a similar unit size reactor. Regimes with largely self-driven plasma current are required to permit a steady-state tokamak reactor with acceptable recirculating power. Reliable techniques of disruption control will be needed to achieve the availability goals of an economic reactor. Thus the central role of this new tokamak facility is to point the way to a more attractive demonstration reactor (DEMO) than the present data base would support. To meet the challenges, we propose a new 'Steady State Advanced Tokamak' (SSAT) facility that would develop and demonstrate optimized steady state tokamak operating mode. While other tokamaks in the world program employ superconducting toroidal field coils, SSAT would be the first major tokamak to operate with a fully superconducting coil set in the elongated, divertor geometry planned for ITER and DEMO.

Tokamak with liquid metal for inducing toroidal electrical field

DOEpatents

Ohkawa, Tihiro

1981-01-01

A tokamak apparatus includes a vessel for defining a reservoir and confining liquid therein. A toroidal liner disposed within said vessel defines a toroidal space within the liner confines gas therein. Liquid metal fills the reservoir outside the liner. A magnetic field is established in the liquid metal to develop magnetic flux linking the toroidal space. The gas is ionized. The liquid metal and the toroidal space are moved relative to one another transversely of the space to generate electric current in the ionized gas in the toroidal space about its major axis and thereby heat plasma developed in the toroidal space.

Electromagnetic toroidal excitations in matter and free space.

PubMed

Papasimakis, N; Fedotov, V A; Savinov, V; Raybould, T A; Zheludev, N I

2016-03-01

The toroidal dipole is a localized electromagnetic excitation, distinct from the magnetic and electric dipoles. While the electric dipole can be understood as a pair of opposite charges and the magnetic dipole as a current loop, the toroidal dipole corresponds to currents flowing on the surface of a torus. Toroidal dipoles provide physically significant contributions to the basic characteristics of matter including absorption, dispersion and optical activity. Toroidal excitations also exist in free space as spatially and temporally localized electromagnetic pulses propagating at the speed of light and interacting with matter. We review recent experimental observations of resonant toroidal dipole excitations in metamaterials and the discovery of anapoles, non-radiating charge-current configurations involving toroidal dipoles. While certain fundamental and practical aspects of toroidal electrodynamics remain open for the moment, we envision that exploitation of toroidal excitations can have important implications for the fields of photonics, sensing, energy and information.

User's guide to the Residual Gas Analyzer (RGA)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Artman, S.A.

1988-08-04

The Residual Gas Analyzer (RGA), a Model 100C UTI quadrupole mass spectrometer, measures the concentrations of selected masses in the Fusion Energy Division's (FED) Advanced Toroidal Facility (ATF). The RGA software is a VAX FORTRAN computer program which controls the experimental apparatus, records the raw data, performs data reduction, and plots the data. The RGA program allows data to be collected from an RGA on ATF or from either of two RGAs in the laboratory. In the laboratory, the RGA diagnostic plays an important role in outgassing studied on various candidate materials for fusion experiments. One such material, graphite, ismore » being used more often in fusion experiments due to its ability to withstand high power loads. One of the functions of the RGA diagnostic is aid in the determination of the best grade of graphite to be used in these experiments and to study the procedures used to condition it. A procedure of particular interest involves baking the graphite sample in order to remove impurities that may be present in it. These impurities can be studied while in the ATF plasma or while being baked and outgassed in the laboratory. The Residual Gas Analyzer is a quadrupole mass spectrometer capable of scanning masses ranging in size from 1 atomic mass unit (amu) to 300 amu while under computer control. The procedure for collecting data for a particular mass is outlined.« less