Use of reconstructed 3D equilibria to determine onset conditions of helical cores in tokamaks for extrapolation to ITER

DOE PAGES

Wingen, A.; Wilcox, R. S.; Seal, S. K.; ...

2018-01-15

In this paper, large, spontaneous m/n = 1/1 helical cores are shown to be expected in tokamaks such as ITER with extended regions of low- or reversed- magnetic shear profiles and q near 1 in the core. The threshold for this spontaneous symmetry breaking is determined using VMEC scans, beginning with reconstructed 3D equilibria from DIII-D and Alcator C-Mod based on observed internal 3D deformations. The helical core is a saturated internal kink mode (Wesson 1986 Plasma Phys. Control. Fusion 28 243); its onset threshold is shown to be proportional tomore » $$({\\rm d}p/{\\rm d}\\rho)/B_t^2$$ around q = 1. Below the threshold, applied 3D fields can drive a helical core to finite size, as in DIII-D. The helical core size thereby depends on the magnitude of the applied perturbation. Above it, a small, random 3D kick causes a bifurcation from axisymmetry and excites a spontaneous helical core, which is independent of the kick size. Systematic scans of the q-profile show that the onset threshold is very sensitive to the q-shear in the core. Helical cores occur frequently in Alcator C-Mod during ramp-up when slow current penetration results in a reversed shear q-profile, which is favorable for helical core formation. In conclusion, a comparison of the helical core onset threshold for discharges from DIII-D, Alcator C-Mod and ITER confirms that while DIII-D is marginally stable, Alcator C-Mod and especially ITER are highly susceptible to helical core formation without being driven by an externally applied 3D magnetic field.« less

Structure of chaotic magnetic field lines in IR-T1 tokamak due to ergodic magnetic limiter

NASA Astrophysics Data System (ADS)

Ahmadi, S.; Salar Elahi, A.; Ghorannevis, M.

2018-03-01

In this paper we have studied an Ergodic Magnetic Limiter (EML) based chaotic magnetic field for transport control in the edge plasma of IR-T1 tokamak. The resonance created by the EML causes perturbation of the equilibrium field line in tokamak and as a result, the field lines are chaotic in the vicinity of the dimerized island chains. Transport barriers are formed in the chaotic field line and actually observe in tokamak with reverse magnetic shear. We used area-preserving non-twist (and twist) Poincaré maps to describe the formation of transport barriers, which are actually features of Hamiltonian systems. This transport barrier is useful in reducing radial diffusion of the field line and thus improving the plasma confinement.

Energetic electrons, hard x-ray emission and MHD activity studies in the IR-T1 tokamak.

PubMed

Agah, K Mikaili; Ghoranneviss, M; Elahi, A Salar

2015-01-01

Determinations of plasma parameters as well as the Magnetohydrodynamics (MHD) activity, energetic electrons energy and energy confinement time are essential for future fusion reactors experiments and optimized operation. Also some of the plasma information can be deduced from these parameters, such as plasma equilibrium, stability, and MHD instabilities. In this contribution we investigated the relation between energetic electrons, hard x-ray emission and MHD activity in the IR-T1 Tokamak. For this purpose we used the magnetic diagnostics and a hard x-ray spectroscopy in IR-T1 tokamak. A hard x-ray emission is produced by collision of the runaway electrons with the plasma particles or limiters. The mean energy was calculated from the slope of the energy spectrum of hard x-ray photons.

Effect of ECRH and resonant magnetic fields on formation of magnetic islands in the T-10 tokamak plasma

NASA Astrophysics Data System (ADS)

Shestakov, E. A.; Savrukhin, P. V.

2017-10-01

Experiments in the T-10 tokamak demonstrated possibility of controlling the plasma current during disruption instability using the electron cyclotron resonance heating (ECRH) and the controlled operation of the ohmic current-holding system. Quasistable plasma discharge with repeating sawtooth oscillations can be restored after energy quench using auxiliary ECRH power when PEC / POH > 2-5. The external magnetic field generation system consisted of eight saddle coils that were arranged symmetrically relative to the equatorial plane of the torus outside of the vacuum vessel of the T-10 tokamak to study the possible resonant magnetic field effects on the rotation frequency of magnetic islands. The saddle coils power supply system is based on four thyristor converters with a total power of 300 kW. The power supply control system is based on Siemens S7 controllers. As shown by preliminary experiments, the interaction efficiency of external magnetic fields with plasma depends on the plasma magnetic configuration. Optimal conditions for slowing the rotation of magnetic islands were determined. Additionally, the direction of the error magnetic field in the T-10 tokamak was determined, and the threshold value of the external magnetic field was determined.

Magnetic evaluation of hydrogen pressures changes on MHD fluctuations in IR-T1 tokamak plasma

NASA Astrophysics Data System (ADS)

Alipour, Ramin; Ghanbari, Mohamad R.

2018-04-01

Identification of tokamak plasma parameters and investigation on the effects of each parameter on the plasma characteristics is important for the better understanding of magnetohydrodynamic (MHD) activities in the tokamak plasma. The effect of different hydrogen pressures of 1.9, 2.5 and 2.9 Torr on MHD fluctuations of the IR-T1 tokamak plasma was investigated by using of 12 Mirnov coils, singular value decomposition and wavelet analysis. The parameters such as plasma current, loop voltage, power spectrum density, energy percent of poloidal modes, dominant spatial structures and temporal structures of poloidal modes at different plasma pressures are plotted. The results indicate that the MHD activities at the pressure of 2.5 Torr are less than them at other pressures. It also has been shown that in the stable area of plasma and at the pressure of 2.5 Torr, the magnetic force and the force of plasma pressure are in balance with each other and the MHD activities are at their lowest level.

Advanced Divertor Design and Application under Modern Superconducting Tokamak Constraints

NASA Astrophysics Data System (ADS)

Covele, Brent; Kotschenreuther, Mike; Mahajan, Swadesh; Valanju, Prashant

2013-10-01

With current ITER projections already predicting divertor exhaust heat loads in the 5-10 MW/m2 range, i.e. at the maximum tolerance, it is clear that the divertor heat load problem will only be exacerbated for future superconducting tokamaks, as well as perhaps some modern tokamaks today. Thus, an advanced divertor, such as the X-Divertor (XD), Super-X Divertor (SXD), or Snowflake (SF) will become a virtual necessity to reduce incident heat flux at the target plates. Using the 2D magnetic equilibrium code CORSICA, we explore the possibilities of creating an advanced divertor for a next-generation superconducting tokamak (Ip = 15 MA, BT = 5.3 T, R = 6.2 m) under nominal engineering constraints. Advanced divertors were achieved with no in-vessel PF coils, PF current densities below 30 MA/m2, and vertical maintenance access, all of which are favorable conditions for tokamaks today. Both the XD and SF divertors are readily achievable while maintaining core plasma performance, and the advantages and disadvantages of each are discussed in turn. Some thought is given as to how the divertor cassette will need to be modified to accommodate advanced divertors. Work supported under US-DOE projects DE-FG02-04ER54742 and DE-FG02-04ER54754.

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.

Simulation of the Plasma Density Evolution during Electron Cyclotron Resonance Heating at the T-10 Tokamak

NASA Astrophysics Data System (ADS)

Dnestrovskij, Yu. N.; Vershkov, V. A.; Danilov, A. V.; Dnestrovskij, A. Yu.; Zenin, V. N.; Lysenko, S. E.; Melnikov, A. V.; Shelukhin, D. A.; Subbotin, G. F.; Cherkasov, S. V.

2018-01-01

In ohmically heated (OH) plasma with low recycling, an improved particle confinement (IPC) mode is established during gas puffing. However, after gas puffing is switched off, this mode is retained only for about 100 ms, after which an abrupt phase transition into the low particle confinement (LPC) mode occurs in the entire plasma cross section. During such a transition, energy transport due to heat conduction does not change. The phase transition in OH plasma is similar to the effect of density pump-out from the plasma core, which occurs after electron cyclotron heating (ECH) is switched on. Analysis of the measured plasma pressure profiles in the T-10 tokamak shows that, after gas puffing in the OH mode is switched off, the plasma pressure profile in the IPC stage becomes more peaked and, after the peakedness exceeds a certain critical value, the IPC-LPC transition occurs. Similar processes are also observed during ECH. If the pressure profile is insufficiently peaked during ECH, then the density pump-out effect comes into play only after the critical peakedness of the pressure profile is reached. In the plasma core, the density and pressure profiles are close to the corresponding canonical profiles. This allows one to derive an expression for the particle flux within the canonical profile model and formulate a criterion for the IPC-LPC transition. The time evolution of the plasma density profile during phase transitions was simulated for a number of T-10 shots with ECH and high recycling. The particle transport coefficients in the IPC and LPC phases, as well as the dependences of these coefficients on the ECH power, are determined.

Tangential System of Thomson Scattering for Tokamak T-15

NASA Astrophysics Data System (ADS)

Asadulin, G. M.; Bel'bas, I. S.; Gorshkov, A. V.

2017-12-01

Two systems of Thomson scattering diagnostics, with vertical and tangential probing, are used in the D-shaped plasma cross section in tokamak T-15. The tangential system allows measuring plasma temperature and density profiles along the major radius of the tokamak. This paper presents the tangential system project. The system is based on a Nd:YAG laser with wavelength of 1064 nm, pulse energy of 3 J, pulse duration of 10 ns, and repetition rate of 100 Hz. The chosen geometry allows collecting light from ten uniformly spaced points. Optimization of the registration system has been accomplished. The collected light will be transmitted through an optical fiber bundle with diameter of 3 mm and quartz fibers (numerical aperture is 0.22). Six-channel polychromators based on high-contrast interference filters have been chosen as spectral equipment. The radiation will be registered by avalanche photodiodes. The technique of electron temperature and density measurement is described, and estimation of its accuracy is carried out. The proposed system allows measuring the electron temperature with accuracy not worse than 10% within the range of 50 eV to 10 keV on the pinch edge over the internal contour, from 20 eV to 9 keV in the plasma central region, and from 2 eV to 400 eV on the pinch edge over the outer contour. The estimation is made for electron density of not less than 2.6 × 1013 cm-3.

Nanoparticle Plasma Jet as Fast Probe for Runaway Electrons in Tokamak Disruptions

NASA Astrophysics Data System (ADS)

Bogatu, I. N.; Galkin, S. A.

2017-10-01

Successful probing of runaway electrons (REs) requires fast (1 - 2 ms) high-speed injection of enough mass able to penetrate through tokamak toroidal B-field (2 - 5 T) over 1 - 2 m distance with large assimilation fraction in core plasma. A nanoparticle plasma jet (NPPJ) from a plasma gun is a unique combination of millisecond trigger-to-delivery response and mass-velocity of 100 mg at several km/s for deep direct injection into current channel of rapidly ( 1 ms) cooling post-TQ core plasma. After C60 NPPJ test bed demonstration we started to work on ITER-compatible boron nitride (BN) NPPJ. Once injected into plasma, BN NP undergoes ablative sublimation, thermally decomposes into B and N, and releases abundant B and N high-charge ions along plasma-traversing path and into the core. We present basic characteristics of our BN NPPJ concept and first results from B and N ions on Zeff > 1 effect on REs dynamics by using a self-consistent model for RE current density. Simulation results of BNQ+ NPPJ penetration through tokamak B-field to RE beam location performed with Hybrid Electro-Magnetic code (HEM-2D) are also presented. Work supported by U.S. DOE SBIR Grant.

ADX: a high field, high power density, advanced divertor and RF tokamak

NASA Astrophysics Data System (ADS)

LaBombard, B.; Marmar, E.; Irby, J.; Terry, J. L.; Vieira, R.; Wallace, G.; Whyte, D. G.; Wolfe, S.; Wukitch, S.; Baek, S.; Beck, W.; Bonoli, P.; Brunner, D.; Doody, J.; Ellis, R.; Ernst, D.; Fiore, C.; Freidberg, J. P.; Golfinopoulos, T.; Granetz, R.; Greenwald, M.; Hartwig, Z. S.; Hubbard, A.; Hughes, J. W.; Hutchinson, I. H.; Kessel, C.; Kotschenreuther, M.; Leccacorvi, R.; Lin, Y.; Lipschultz, B.; Mahajan, S.; Minervini, J.; Mumgaard, R.; Nygren, R.; Parker, R.; Poli, F.; Porkolab, M.; Reinke, M. L.; Rice, J.; Rognlien, T.; Rowan, W.; Shiraiwa, S.; Terry, D.; Theiler, C.; Titus, P.; Umansky, M.; Valanju, P.; Walk, J.; White, A.; Wilson, J. R.; Wright, G.; Zweben, S. J.

2015-05-01

The MIT Plasma Science and Fusion Center and collaborators are proposing a high-performance Advanced Divertor and RF tokamak eXperiment (ADX)—a tokamak specifically designed to address critical gaps in the world fusion research programme on the pathway to next-step devices: fusion nuclear science facility (FNSF), fusion pilot plant (FPP) and/or demonstration power plant (DEMO). This high-field (⩾6.5 T, 1.5 MA), high power density facility (P/S ˜ 1.5 MW m-2) will test innovative divertor ideas, including an ‘X-point target divertor’ concept, at the required performance parameters—reactor-level boundary plasma pressures, magnetic field strengths and parallel heat flux densities entering into the divertor region—while simultaneously producing high-performance core plasma conditions that are prototypical of a reactor: equilibrated and strongly coupled electrons and ions, regimes with low or no torque, and no fuelling from external heating and current drive systems. Equally important, the experimental platform will test innovative concepts for lower hybrid current drive and ion cyclotron range of frequency actuators with the unprecedented ability to deploy launch structures both on the low-magnetic-field side and the high-magnetic-field side—the latter being a location where energetic plasma-material interactions can be controlled and favourable RF wave physics leads to efficient current drive, current profile control, heating and flow drive. This triple combination—advanced divertors, advanced RF actuators, reactor-prototypical core plasma conditions—will enable ADX to explore enhanced core confinement physics, such as made possible by reversed central shear, using only the types of external drive systems that are considered viable for a fusion power plant. Such an integrated demonstration of high-performance core-divertor operation with steady-state sustainment would pave the way towards an attractive pilot plant, as envisioned in the ARC concept

Li experiments at the tokamak T-11 M in field of steady state PFC investigations

NASA Astrophysics Data System (ADS)

Mirnov, S. V.; Lazarev, V. B.

2011-08-01

The renewable plasma facing components (PFCs) of steady state tokamak-reactor can be created in framework of Lithium emitter-collector concept, which suggests Li-loop development close the Li-PFC and plasma periphery. It should ensure: Li-emission from PFC into the plasma, plasma periphery cooling by non-coronal Li radiation, Li ions collection before their loss on the wall and Li return into emitter. The subjects of the last T-11 M investigations were the Lithium collection by limiters and Lithium removal from the wall during tokamak conditioning. The Lithium behavior was studied with witness samples and mobile graphite probe. It was shown that Li-deposit on the sides of rail Li-limiter (collector) is proportional to the Li-emission from the Li-limiter (emitter). Lithium deposit on the ion-drift side of Li-limiter is up to 2-3 times more than on the electron-side. The efficiency of Li-collection by T-11 M limiters can be 60 ± 20% of total Lithium emission from Li-limiter during plasma discharges.

Edge Thomson scattering diagnostic on COMPASS tokamak: Installation, calibration, operation, improvements

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

Bohm, P., E-mail: bohm@ipp.cas.cz; Bilkova, P.; Melich, R.

2014-11-15

The core Thomson scattering diagnostic (TS) on the COMPASS tokamak was put in operation and reported earlier. Implementation of edge TS, with spatial resolution along the laser beam up to ∼1/100 of the tokamak minor radius, is presented now. The procedure for spatial calibration and alignment of both core and edge systems is described. Several further upgrades of the TS system, like a triggering unit and piezo motor driven vacuum window shutter, are introduced as well. The edge TS system, together with the core TS, is now in routine operation and provides electron temperature and density profiles.

The conceptual design of a robust, compact, modular tokamak reactor based on high-field superconductors

NASA Astrophysics Data System (ADS)

Whyte, D. G.; Bonoli, P.; Barnard, H.; Haakonsen, C.; Hartwig, Z.; Kasten, C.; Palmer, T.; Sung, C.; Sutherland, D.; Bromberg, L.; Mangiarotti, F.; Goh, J.; Sorbom, B.; Sierchio, J.; Ball, J.; Greenwald, M.; Olynyk, G.; Minervini, J.

2012-10-01

Two of the greatest challenges to tokamak reactors are 1) large single-unit cost of each reactor's construction and 2) their susceptibility to disruptions from operation at or above operational limits. We present an attractive tokamak reactor design that substantially lessens these issues by exploiting recent advancements in superconductor (SC) tapes allowing peak field on SC coil > 20 Tesla. A R˜3.3 m, B˜9.2 T, ˜ 500 MW fusion power tokamak provides high fusion gain while avoiding all disruptive operating boundaries (no-wall beta, kink, and density limits). Robust steady-state core scenarios are obtained by exploiting the synergy of high field, compact size and ideal efficiency current drive using high-field side launch of Lower Hybrid waves. The design features a completely modular replacement of internal solid components enabled by the demountability of the coils/tapes and the use of an immersion liquid blanket. This modularity opens up the possibility of using the device as a nuclear component test facility.

Modulation of Core Turbulent Density Fluctuations by Large-Scale Neoclassical Tearing Mode Islands in the DIII-D Tokamak

DOE PAGES

Bardóczi, L.; Rhodes, T. L.; Carter, T. A.; ...

2016-05-26

We report the first observation of localized modulation of turbulent density uctuations en (via Beam Emission Spectroscopy) by neoclassical tearing modes (NTMs) in the core of the DIII-D tokamak. NTMs are important as they often lead to severe degradation of plasma confinement and disruptions in high-confinement fusion experiments. Magnetic islands associated with NTMs significantly modify the profiles and turbulence drives. In this experiment n was found to be modulated by 14% across the island. Gyrokinetic simulations suggest that en could be dominantly driven by the ion temperature gradient (ITG) instability.

Scoping study for compact high-field superconducting net energy tokamaks

NASA Astrophysics Data System (ADS)

Mumgaard, R. T.; Greenwald, M.; Freidberg, J. P.; Wolfe, S. M.; Hartwig, Z. S.; Brunner, D.; Sorbom, B. N.; Whyte, D. G.

2016-10-01

The continued development and commercialization of high temperature superconductors (HTS) may enable the construction of compact, net-energy tokamaks. HTS, in contrast to present generation low temperature superconductors, offers improved performance in high magnetic fields, higher current density, stronger materials, higher temperature operation, and simplified assembly. Using HTS along with community-consensus confinement physics (H98 =1) may make it possible to achieve net-energy (Q>1) or burning plasma conditions (Q>5) in DIII-D or ASDEX-U sized, conventional aspect ratio tokamaks. It is shown that, by operating at high plasma current and density enabled by the high magnetic field (B>10T), the required triple products may be achieved at plasma volumes under 20m3, major radii under 2m, with external heating powers under 40MW. This is at the scale of existing devices operated by laboratories, universities and companies. The trade-offs in the core heating, divertor heat exhaust, sustainment, stability, and proximity to known plasma physics limits are discussed in the context of the present tokamak experience base and the requirements for future devices. The resulting HTS-based design space is compared and contrasted to previous studies on high-field copper experiments with similar missions. The physics exploration conducted with such HTS devices could decrease the real and perceived risks of ITER exploitation, and aid in quickly developing commercially-applicable tokamak pilot plants and reactors.

Public Data Set: Non-inductively Driven Tokamak Plasmas at Near-Unity βt in the Pegasus Toroidal Experiment

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

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

This public data set contains openly-documented, machine readable digital research data corresponding to figures published in J.A. Reusch et al., 'Non-inductively Driven Tokamak Plasmas at Near-Unity βt in the Pegasus Toroidal Experiment,' Phys. Plasmas 25, 056101 (2018).

Edge-core interaction of ITG turbulence in Tokamaks: Is the Tail Wagging the Dog?

NASA Astrophysics Data System (ADS)

Ku, S.; Chang, C. S.; Dif-Pradalier, G.; Diamond, P. H.

2010-11-01

A full-f XGC1 gyrokinetic simulation of ITG turbulence, together with the neoclassical dynamics without scale separation, has been performed for the whole-volume plasma in realistic diverted DIII-D geometry. The simulation revealed that the global structure of the turbulence and transport in tokamak plasmas results from a synergy between edge-driven inward propagation of turbulence intensity and the core-driven outward heat transport. The global ion confinement and the ion temperature gradient then self-organize quickly at turbulence propagation time scale. This synergy results in inward-outward pulse scattering leading to spontaneous production of strong internal shear layers in which the turbulent transport is almost suppressed over several radial correlation lengths. Co-existence of the edge turbulence source and the strong internal shear layer leads to radially increasing turbulence intensity and ion thermal transport profiles.

Boundary perturbations coupled to core 3/2 tearing modes on the DIII-D tokamak

NASA Astrophysics Data System (ADS)

Tobias, B.; Yu, L.; Domier, C. W.; Luhmann, N. C., Jr.; Austin, M. E.; Paz-Soldan, C.; Turnbull, A. D.; Classen, I. G. J.; the DIII-D Team

2013-09-01

High confinement (H-mode) discharges on the DIII-D tokamak are routinely subject to the formation of long-lived, non-disruptive magnetic islands that degrade confinement and limit fusion performance. Simultaneous, 2D measurement of electron temperature fluctuations in the core and edge regions allows for reconstruction of the radially resolved poloidal mode number spectrum and phase of the global plasma response associated with these modes. Coherent, n = 2 excursions of the plasma boundary are found to be the result of coupling to an n = 2, kink-like mode which arises locked in phase to the 3/2 island chain. This coupling dictates the relative phase of the displacement at the boundary with respect to the tearing mode. This unambiguous phase relationship, for which no counter-examples are observed, is presented as a test for modeling of the perturbed fields to be expected outside the confined plasma.

Realizing steady-state tokamak operation for fusion energy

NASA Astrophysics Data System (ADS)

Luce, T. C.

2011-03-01

Continuous operation of a tokamak for fusion energy has clear engineering advantages but requires conditions beyond those sufficient for a burning plasma. The fusion reactions and external sources must support both the pressure and the current equilibrium without inductive current drive, leading to demands on stability, confinement, current drive, and plasma-wall interactions that exceed those for pulsed tokamaks. These conditions have been met individually, and significant progress has been made in the past decade to realize scenarios where the required conditions are obtained simultaneously. Tokamaks are operated routinely without disruptions near pressure limits, as needed for steady-state operation. Fully noninductive sustainment with more than half of the current from intrinsic currents has been obtained for a resistive time with normalized pressure and confinement approaching those needed for steady-state conditions. One remaining challenge is handling the heat and particle fluxes expected in a steady-state tokamak without compromising the core plasma performance.

Remote network control plasma diagnostic system for Tokamak T-10

NASA Astrophysics Data System (ADS)

Troynov, V. I.; Zimin, A. M.; Krupin, V. A.; Notkin, G. E.; Nurgaliev, M. R.

2016-09-01

The parameters of molecular plasma in closed magnetic trap is studied in this paper. Using the system of molecular diagnostics, which was designed by the authors on the «Tokamak T-10» facility, the radiation of hydrogen isotopes at the plasma edge is investigated. The scheme of optical radiation registration within visible spectrum is described. For visualization, identification and processing of registered molecular spectra a new software is developed using MatLab environment. The software also includes electronic atlas of electronic-vibrational-rotational transitions for molecules of protium and deuterium. To register radiation from limiter cross-section a network control system is designed using the means of the Internet/Intranet. Remote control system diagram and methods are given. The examples of web-interfaces for working out equipment control scenarios and viewing of results are provided. After test run in Intranet, the remote diagnostic system will be accessible through Internet.

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.

Experimental identification of nonlinear coupling between (intermediate, small)-scale microturbulence and an MHD mode in the core of a superconducting tokamak

NASA Astrophysics Data System (ADS)

Sun, P. J.; Li, Y. D.; Ren, Y.; Zhang, X. D.; Wu, G. J.; Xu, L. Q.; Chen, R.; Li, Q.; Zhao, H. L.; Zhang, J. Z.; Shi, T. H.; Wang, Y. M.; Lyu, B.; Hu, L. Q.; Li, J.; The EAST Team

2018-01-01

In this paper, we present clear experimental evidence of core region nonlinear coupling between (intermediate, small)-scale microturbulence and an magnetohydrodynamics (MHD) mode during the current ramp-down phase in a set of L-mode plasma discharges in the experimental advanced superconducting tokamak (EAST, Wan et al (2006 Plasma Sci. Technol. 8 253)). Density fluctuations of broadband microturbulence (k\\perpρi˜2{-}5.2 ) and the MHD mode (toroidal mode number m = -1 , poloidal mode number n = 1 ) are measured simultaneously, using a four-channel tangential CO2 laser collective scattering diagnostic in core plasmas. The nonlinear coupling between the broadband microturbulence and the MHD mode is directly demonstrated by showing a statistically significant bicoherence and modulation of turbulent density fluctuation amplitude by the MHD mode.

Wavelength calibration of x-ray imaging crystal spectrometer on Joint Texas Experimental Tokamak

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

Yan, W.; Chen, Z. Y., E-mail: zychen@hust.edu.cn; Jin, W.

2014-11-15

The wavelength calibration of x-ray imaging crystal spectrometer is a key issue for the measurements of plasma rotation. For the lack of available standard radiation source near 3.95 Å and there is no other diagnostics to measure the core rotation for inter-calibration, an indirect method by using tokamak plasma itself has been applied on joint Texas experimental tokamak. It is found that the core toroidal rotation velocity is not zero during locked mode phase. This is consistent with the observation of small oscillations on soft x-ray signals and electron cyclotron emission during locked-mode phase.

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.

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

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

Modernized active spectroscopic diagnostics (CXRS) of the T-10 tokamak

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

Krupin, V. A., E-mail: Krupin-VA@nrcki.ru; Klyuchnikov, L. A., E-mail: Lklyuchnikov@list.ru; Korobov, K. V., E-mail: Korobov-KV@nrcki.ru

2015-12-15

This work presents the results of modernization of the CXRS (charge exchange recombination spectroscopy) diagnostics [1] at the T-10 tokamak. The relevance of this work is due to the importance of measurements of the ion temperature and nuclei density of the working gas and impurities for analysis of transport processes in the plasma ion component. Measurements of radial profiles of the ion temperature are extremely important for investigating the geodesic acoustic mode behavior which is conducted at the T-10 [2]. The modernized scheme of CXRS measurements, as well as the design and operational features of the spectrometer created for themore » new diagnostics, is described. Principles of data recording and further processing are considered in detail; attention is given to the problem of calibration of the whole complex of equipment. The performed changes in diagnostics allow the measurements to be taken simultaneously in three spectral intervals: in the region of the beam line H{sub α}, the CXRS line of carbon ion C{sup 5+}, and the CXRS line of one of the hydrogen-like ions: He{sup 1+}, Li{sup 2+}, N{sup 6+}, O{sup 7+} or Ne{sup 9+}. This makes it possible to measure the density profiles of two plasma impurities simultaneously, as well as the ion temperature from CXRS lines of different elements. The modernized diagnostics significantly broadens the possibilities of studying the physics of transport processes and quasi-coherent modes of plasma oscillations at the T-10.« less

Tank 241-T-204, core 188 analytical results for the final report

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

Nuzum, J.L.

TANK 241-T-204, CORE 188, ANALYTICAL RESULTS FOR THE FINAL REPORT. This document is the final laboratory report for Tank 241 -T-204. Push mode core segments were removed from Riser 3 between March 27, 1997, and April 11, 1997. Segments were received and extruded at 222-8 Laboratory. Analyses were performed in accordance with Tank 241-T-204 Push Mode Core Sampling and analysis Plan (TRAP) (Winkleman, 1997), Letter of instruction for Core Sample Analysis of Tanks 241-T-201, 241- T-202, 241-T-203, and 241-T-204 (LAY) (Bell, 1997), and Safety Screening Data Qual@ Objective (DO) ODukelow, et al., 1995). None of the subsamples submitted for totalmore » alpha activity (AT) or differential scanning calorimetry (DC) analyses exceeded the notification limits stated in DO. The statistical results of the 95% confidence interval on the mean calculations are provided by the Tank Waste Remediation Systems Technical Basis Group and are not considered in this report.« less

Realizing Steady State Tokamak Operation for Fusion Energy

NASA Astrophysics Data System (ADS)

Luce, T. C.

2009-11-01

Continuous operation of a tokamak for fusion energy has obvious engineering advantages, but also presents physics challenges beyond the achievement of conditions needed for a burning plasma. The power from fusion reactions and external sources must support both the pressure and the current equilibrium without inductive current drive, leading to demands on stability, confinement, current drive, and plasma-wall interactions that exceed those for pulsed tokamaks. These conditions have been met individually in the present generation of tokamaks, and significant progress has been made in the last decade to realize scenarios where the required conditions are obtained simultaneously. Tokamaks are now operated routinely without disruptions close to the ideal MHD pressure limit, as needed for steady-state operation. Scenarios that project to high fusion gain have been demonstrated where more than half of the current is supplied by the ``bootstrap'' current generated by the pressure gradient in the plasma. Fully noninductive sustainment has been obtained for about a resistive time (the longest intrinsic time scale in the confined plasma) with normalized pressure and confinement approaching those needed for demonstration of steady-state conditions in ITER. One key challenge remaining to be addressed is how to handle the demanding heat and particle fluxes expected in a steady-state tokamak without compromising the high level of core plasma performance. Rather than attempt a comprehensive historical survey, this review will start from the plasma requirements of a steady-state tokamak powerplant, illustrate with examples the progress made in both experimental and theoretical understanding, and point to the remaining physics challenges.

3D Field Modifications of Core Neutral Fueling In the EMC3-EIRENE Code

NASA Astrophysics Data System (ADS)

Waters, Ian; Frerichs, Heinke; Schmitz, Oliver; Ahn, Joon-Wook; Canal, Gustavo; Evans, Todd; Feng, Yuehe; Kaye, Stanley; Maingi, Rajesh; Soukhanovskii, Vsevolod

2017-10-01

The application of 3-D magnetic field perturbations to the edge plasmas of tokamaks has long been seen as a viable way to control damaging Edge Localized Modes (ELMs). These 3-D fields have also been correlated with a density drop in the core plasmas of tokamaks; known as `pump-out'. While pump-out is typically explained as the result of enhanced outward transport, degraded fueling of the core may also play a role. By altering the temperature and density of the plasma edge, 3-D fields will impact the distribution function of high energy neutral particles produced through ion-neutral energy exchange processes. Starved of the deeply penetrating neutral source, the core density will decrease. Numerical studies carried out with the EMC3-EIRENE code on National Spherical Tokamak eXperiment-Upgrade (NSTX-U) equilibria show that this change to core fueling by high energy neutrals may be a significant contributor to the overall particle balance in the NSTX-U tokamak: deep core (Ψ < 0.5) fueling from neutral ionization sources is decreased by 40-60% with RMPs. This work was funded by the US Department of Energy under Grant DE-SC0012315.

Dust-Particle Transport in Tokamak Edge Plasmas

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

Pigarov, A Y; Krasheninnikov, S I; Soboleva, T K

2005-09-12

Dust particulates in the size range of 10nm-100{micro}m are found in all fusion devices. Such dust can be generated during tokamak operation due to strong plasma/material-surface interactions. Some recent experiments and theoretical estimates indicate that dust particles can provide an important source of impurities in the tokamak plasma. Moreover, dust can be a serious threat to the safety of next-step fusion devices. In this paper, recent experimental observations on dust in fusion devices are reviewed. A physical model for dust transport simulation, and a newly developed code DUSTT, are discussed. The DUSTT code incorporates both dust dynamics due to comprehensivemore » dust-plasma interactions as well as the effects of dust heating, charging, and evaporation. The code tracks test dust particles in realistic plasma backgrounds as provided by edge-plasma transport codes. Results are presented for dust transport in current and next-step tokamaks. The effect of dust on divertor plasma profiles and core plasma contamination is examined.« less

Development of lithium and tungsten limiters for test on T-10 tokamak at high heat load condition

NASA Astrophysics Data System (ADS)

Lyublinski, I. E.; Vertkov, A. V.; Zharkov, M. Yu; Vershkov, V. A.; Mirnov, S. V.

2016-04-01

Application of a complex of powerful (up to 3 MW) ECR plasma heating in T-10 tokamak is pulled down with a problem of the strong plasma pollution at power input more than 2 MW. For the solution of these problems the new W and Li limiters is developed and prepared to implementation. As it is supposed, application of W as a plasma facing material will allow excluding carbon influx into vacuum chamber. An additional Li limiter arranged in a shadow of W one will be used as a Li source for plasma periphery cooling due to a reradiation on Li that will lead to decrease in power deposition on W limiters. Parameters and design of limiters are presented. Plasma facing surface of a limiter is made of capillary-porous system (CPS) with Li. Porous matrix of CPS (W felt) provides stability of liquid Li surface under MHD force effect and an opportunity of its constant renewal due to capillary forces. The necessary Li flux from a Li limiter surface is estimated for maintenance of normal operation mode of W limiters at ECRH power of 3 MW during 400 ms. It is shown, that upgrade of limiters in tokamak T-10 will allow providing of ECR plasma heating with power up to 3 MW at reasonable Li flux.

Sub-Alfvénic reduced magnetohydrodynamic equations for tokamaks

NASA Astrophysics Data System (ADS)

Sengupta, W.; Hassam, A. B.; Antonsen, T. M.

2017-06-01

A reduced set of magnetohydrodynamic (MHD) equations is derived, applicable to large aspect ratio tokamaks and relevant for dynamics that is sub-Alfvénic with respect to ideal ballooning modes. This ordering optimally allows sound waves, Mercier modes, drift modes, geodesic-acoustic modes (GAM), zonal flows and shear Alfvén waves. Wavelengths long compared to the gyroradius but comparable to the minor radius of a typical tokamak are considered. With the inclusion of resistivity, tearing modes, resistive ballooning modes, Pfirsch-Schluter cells and the Stringer spin-up are also included. A major advantage is that the resulting system is two-dimensional in space, and the system incorporates self-consistent and dynamic Shafranov shifts. A limitation is that the system is valid only in radial domains where the tokamak safety factor, , is close to rational. In the tokamak core, the system is well suited to study the sawtooth discharge in the presence of Mercier modes. The systematic ordering scheme and methodology developed are versatile enough to reduce the more general collisional two-fluid equations or possibly the Vlasov-Maxwell system in the MHD ordering.

Compact Torus Fueling of the STOR-M Tokamak

NASA Astrophysics Data System (ADS)

Xiao, C.; Hirose, A.; Zawalski, W.; White, D.; Raman, R.; Decoste, R.; Gregory, B. C.; Martin, F.

1996-11-01

Tangential injection of accelerated compact torus (CT) has been performed on the STOR-M tokamak (R/a=46/12 cm, B_t<1 T, I_p<= 50 kA, barn_e=(0.5 - 1)×10^13 cm-3) using the University of Saskatchewan Compact Torus Injector (USCTI). The CT parameters are: m~=1 μg, v=120 km/sec, B=0.1 T and n=(2 - 4)×10^15 cm-3. After CT injection, the electron density in tokamak doubles and the poloidal β-value increases. Indications of reduction in the loop voltage and H_α emission level have also been observed. Currently, following efforts are being made: (a) to coat chromium on the electrode surface, (b) to increase the on-line baking temperature, and (c) to reduce the neutral gas load which follows the CT plasma. In addition, numerical calculation of CT motion in a tokamak magnetic field has been carried out. For horizontal injection, the initial CT magnetic dipole direction should be aligned with the CT velocity for deeper penetration. In the case of vertical injection, the CT trajectory is independent of the initial magnetic dipole direction and central penetration is facilitated by off-axis injection.

Influence of molecular clustering on the interpretation of diffractograms of hydrocarbon films from tokamak T-10

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

Neverov, V. S., E-mail: vs-never@hotmail.com; Voloshinov, V. V., E-mail: vladimir.voloshinov@gmail.com; Kukushkin, A. B., E-mail: kukushkin-ab@nrcki.ru

2015-12-15

The influence of molecular clustering on the formerly suggested interpretation of diffraction patterns of hydrocarbon films formed in the vacuum vessel of the tokamak T-10 is analyzed numerically. The simulation of clustering of simple hydrocarbon molecules C(D, H){sub 4}, C{sub 2}(D, H){sub 4}, and C{sub 6}(D, H){sub 6} and molecules composed of curved graphene (fullerenes and toroidal nanotubes) is carried out with the rigid body molecular dynamics method. It is shown that formerly neglected atomic correlations C–C and C–D(H) in the amorphous hydrocarbon component decrease the calculated values of the scattered intensity in the range of scattering vector modulus 5more » < q < 20 nm{sup –1} because of homogenization of scatters on the spatial scale of ∼1 nm. The allowance for these correlations does not change the diffraction patterns in the range q > 20 nm{sup –1}. The results suggest the necessity to introduce to the procedure of determining the structural content of the films, similar to those from the tokamak T-10, the clusters formed by the van der Waals adhesion of hydrocarbon molecules to “graphene” nanoparticles. This simplifies the mathematical optimization to the former level of complexity—but for an extended ensemble of objects—and makes it possible to calculate the diffraction patterns of these objects using the distributed computing resources. A modified algorithm of structural content identification on the basis of joint X-ray and neutron diffractometry is suggested.« less

Final technical report for DE-SC00012633 AToM (Advanced Tokamak Modeling)

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

Holland, Christopher; Orlov, Dmitri; Izzo, Valerie

This final report for the AToM project documents contributions from University of California, San Diego researchers over the period of 9/1/2014 – 8/31/2017. The primary focus of these efforts was on performing validation studies of core tokamak transport models using the OMFIT framework, including development of OMFIT workflow scripts. Additional work was performed to develop tools for use of the nonlinear magnetohydrodynamics code NIMROD in OMFIT, and its use in the study of runaway electron dynamics in tokamak disruptions.

ADX - Advanced Divertor and RF Tokamak Experiment

NASA Astrophysics Data System (ADS)

Greenwald, Martin; Labombard, Brian; Bonoli, Paul; Irby, Jim; Terry, Jim; Wallace, Greg; Vieira, Rui; Whyte, Dennis; Wolfe, Steve; Wukitch, Steve; Marmar, Earl

2015-11-01

The Advanced Divertor and RF Tokamak Experiment (ADX) is a design concept for a compact high-field tokamak that would address boundary plasma and plasma-material interaction physics challenges whose solution is critical for the viability of magnetic fusion energy. This device would have two crucial missions. First, it would serve as a Divertor Test Tokamak, developing divertor geometries, materials and operational scenarios that could meet the stringent requirements imposed in a fusion power plant. By operating at high field, ADX would address this problem at a level of power loading and other plasma conditions that are essentially identical to those expected in a future reactor. Secondly, ADX would investigate the physics and engineering of high-field-side launch of RF waves for current drive and heating. Efficient current drive is an essential element for achieving steady-state in a practical, power producing fusion device and high-field launch offers the prospect of higher efficiency, better control of the current profile and survivability of the launching structures. ADX would carry out this research in integrated scenarios that simultaneously demonstrate the required boundary regimes consistent with efficient current drive and core performance.

Startup and stability of a small spherical tokamak

NASA Astrophysics Data System (ADS)

Garstka, Gregory Douglas

1997-09-01

The spherical tokamak (ST) is an evolutionary extension of the conventional tokamak concept where the aspect ratio is less than 2. These devices may possess significant advantages over standard tokamaks-they are capable of achieving higher values of /beta, seem to be more resilient to disruptions, and are significantly smaller than conventional tokamaks. Two important questions for the next generation of spherical tokamaks concern startup and internal reconnection events (IREs). Understanding startup is crucial due to the limited amount of ohmic flux in an ST. The IREs are disruption- like events observed on STs that do not result in termination of the current channel. Experiments have been conducted on the Madison EDUcational Small Aspect-ratio (MEDUSA) tokamak to answer some of the questions about startup and IREs in STs. MEDUSA is a small ohmic tokamak with an insulating vacuum vessel. Major parameters are R=12 cm, a=8 cm, Ip=10-40 kA, BT=0.2-0.45 T, /Delta tpulse=1-2 ms, /langle ne/rangle/approx5×1019/ m-3, and Te0/approx100 eV. The experiments in this work were aided by an internal magnetic probe array that constrained the reconstruction of MHD equilibria. It was found that startup efficiency, measured by the Ejima coefficient CE, improved with increasing loop voltage and toroidal field. Double tearing modes were found to be an important mechanism for current penetration in MEDUSA; their presence early in the discharge can improve the magnetic flux consumption. The lowest achieved value of the Ejima coefficient was 0.61 (0.13 for 'OH only') for a discharge with 0.375 T toroidal field and 9.4 V startup loop voltage. The study of internal reconnection events revealed the presence of a heretofore undiscovered precursor, which in MEDUSA was manifested as coherent oscillations in the internal poloidal field at 65-75 kHz for 100 μs prior to the IRE. These events were found to result in decreased /ell i and /beta, inward movement of the magnetic axis, dramatically

Numerical modelling of geodesic acoustic mode relaxation in a tokamak edge

DOE PAGES

Dorf, M. A.; Cohen, R. H.; Dorr, M.; ...

2013-05-08

Here, the edge of a tokamak in a high confinement (H mode) regime is characterized by steep density gradients and a large radial electric field. Recent analytical studies demonstrated that the presence of a strong radial electric field consistent with a subsonic pedestal equilibrium modifies the conventional results of the neoclassical formalism developed for the core region. In the present work we make use of the recently developed gyrokinetic code COGENT to numerically investigate neoclassical transport in a tokamak edge including the effects of a strong radial electric field. The results of numerical simulations are found to be in goodmore » qualitative agreement with the theoretical predictions and the quantitative discrepancy is discussed. In addition, the present work investigates the effects of a strong radial electric field on the relaxation of geodesic acoustic modes (GAMs) in a tokamak edge. Numerical simulations demonstrate that the presence of a strong radial electric field characteristic of a tokamak pedestal can enhance the GAM decay rate, and heuristic arguments elucidating this finding are provided.« less

High power heating of magnetic reconnection in merging tokamak experiments

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

Ono, Y.; Tanabe, H.; Gi, K.

2015-05-15

Significant ion/electron heating of magnetic reconnection up to 1.2 keV was documented in two spherical tokamak plasma merging experiment on MAST with the significantly large Reynolds number R∼10{sup 5}. Measured 1D/2D contours of ion and electron temperatures reveal clearly energy-conversion mechanisms of magnetic reconnection: huge outflow heating of ions in the downstream and localized heating of electrons at the X-point. Ions are accelerated up to the order of poloidal Alfven speed in the reconnection outflow region and are thermalized by fast shock-like density pileups formed in the downstreams, in agreement with recent solar satellite observations and PIC simulation results. The magneticmore » reconnection efficiently converts the reconnecting (poloidal) magnetic energy mostly into ion thermal energy through the outflow, causing the reconnection heating energy proportional to square of the reconnecting (poloidal) magnetic field B{sub rec}{sup 2}  ∼  B{sub p}{sup 2}. The guide toroidal field B{sub t} does not affect the bulk heating of ions and electrons, probably because the reconnection/outflow speeds are determined mostly by the external driven inflow by the help of another fast reconnection mechanism: intermittent sheet ejection. The localized electron heating at the X-point increases sharply with the guide toroidal field B{sub t}, probably because the toroidal field increases electron confinement and acceleration length along the X-line. 2D measurements of magnetic field and temperatures in the TS-3 tokamak merging experiment also reveal the detailed reconnection heating mechanisms mentioned above. The high-power heating of tokamak merging is useful not only for laboratory study of reconnection but also for economical startup and heating of tokamak plasmas. The MAST/TS-3 tokamak merging with B{sub p} > 0.4 T will enables us to heat the plasma to the alpha heating regime: T{sub i} > 5 keV without using any additional heating

Quiescent double barrier regime in the DIII-D tokamak.

PubMed

Greenfield, C M; Burrell, K H; DeBoo, J C; Doyle, E J; Stallard, B W; Synakowski, E J; Fenzi, C; Gohil, P; Groebner, R J; Lao, L L; Makowski, M A; McKee, G R; Moyer, R A; Rettig, C L; Rhodes, T L; Pinsker, R I; Staebler, G M; West, W P

2001-05-14

A new sustained high-performance regime, combining discrete edge and core transport barriers, has been discovered in the DIII-D tokamak. Edge localized modes (ELMs) are replaced by a steady oscillation that increases edge particle transport, thereby allowing particle control with no ELM-induced pulsed divertor heat load. The core barrier resembles those usually seen with a low (L) mode edge, without the degradation often associated with ELMs. The barriers are separated by a narrow region of high transport associated with a zero crossing in the E x B shearing rate.

Probing spherical tokamak plasmas using charged fusion products

NASA Astrophysics Data System (ADS)

Boeglin, Werner U.; Perez, Ramona V.; Darrow, Douglass S.; Cecconello, Marco; Klimek, Iwona; Allan, Scott Y.; Akers, Rob J.; Jones, Owen M.; Keeling, David L.; McClements, Ken G.; Scannell, Rory

2015-11-01

The detection of charged fusion products, such as protons and tritons resulting from D(d,p)t reactions, can be used to determine the fusion reaction rate profile in large spherical tokamak plasmas with neutral beam heating. The time resolution of a diagnostic of this type makes it possible to study the slowly-varying beam density profile, as well as rapid changes resulting from MHD instabilities. A 4-channel prototype proton detector (PD) was installed and operated on the MAST spherical tokamak in August/September 2013, and a new 6-channel system for the NSTX-U spherical tokamak is under construction. PD and neutron camera measurements obtained on MAST will be compared with TRANSP calculations, and the design of the new NSTX-U system will be presented, together with the first results from this diagnostic, if available. Supported in part by DOE DE-SC0001157.

Plasma Turbulence Imaging via Beam Emission Spectroscopy in the Core of the DIII-D Tokamak

NASA Astrophysics Data System (ADS)

McKee, George R.; Fonck, Raymond J.; Gupta, Deepak K.; Schlossberg, David J.; Shafer, Morgan W.; Boivin, Réjean L.; Solomon, Wayne

Beam Emission Spectroscopy (BES), a high-sensitivity, good spatial resolution imaging diagnostic system, has been deployed and recently upgraded and expanded at the DIII-D tokamak to better understand density fluctuations arising from plasma turbulence. The currently deployed system images density fluctuations over an approximately 5 × 7 cm region at the plasma mid-plane (radially scannable over 0.2 < r/a ≤ 1) with a 5 × 6 (radial × poloidal) grid of rectangular detection channels, with one microsecond time resolution. BES observes collisionally-induced, Doppler-shifted Dα fluorescence (λ = 652-655 nm) of injected deuterium neutral beam atoms. The diagnostic wavenumber sensitivity is approximately k⊥ < 2.5 cm-1, allowing measurement of longwavelength (k⊥ρI < 1) density fluctuations. The recent upgrade includes expanded fiber optics bundles, customdesigned high-transmission, sharp-edge interference filters, ultra fast collection optics, and enlarged photodiode detectors that together provide nearly an order of magnitude increase in sensitivity relative to an earlier generation BES system. The high sensitivity allows visualization of turbulence at normalized density fluctuation amplitudes of ‾n/n < 1%, typical of fluctuation levels in the core region. The imaging array allows for sampling over 2-3 turbulent eddy scale lengths, which captures the essential dynamics of eddy evolution, interaction and shearing.

Transport and reconnection in tokamak sawteeth.

PubMed

Gentle, K W; Austin, M E; Phillips, P E

2003-12-19

The core of a tokamak discharge often undergoes periodic relaxation oscillations, sawteeth, as the steepening current and temperature profiles are flattened by fast reconnection events. Careful analysis of the electron temperature evolution over this cycle gives an estimate of the energy dissipated in the electrons during reconnection and a measure of the transport characteristic (energy flux versus temperature gradient) over the range of parameters occurring over the remainder of the cycle. The energy dissipated is consistent with estimates of the loss of poloidal magnetic energy. The transport characteristics exhibit a wide range of behaviors.

Integrated fusion simulation with self-consistent core-pedestal coupling

DOE PAGES

Meneghini, O.; Snyder, P. B.; Smith, S. P.; ...

2016-04-20

In this study, accurate prediction of fusion performance in present and future tokamaks requires taking into account the strong interplay between core transport, pedestal structure, current profile and plasma equilibrium. An integrated modeling workflow capable of calculating the steady-state self- consistent solution to this strongly-coupled problem has been developed. The workflow leverages state-of-the-art components for collisional and turbulent core transport, equilibrium and pedestal stability. Validation against DIII-D discharges shows that the workflow is capable of robustly pre- dicting the kinetic profiles (electron and ion temperature and electron density) from the axis to the separatrix in good agreement with the experiments.more » An example application is presented, showing self-consistent optimization for the fusion performance of the 15 MA D-T ITER baseline scenario as functions of the pedestal density and ion effective charge Z eff.« less

Status of the tokamak program

NASA Astrophysics Data System (ADS)

Sheffield, J.

1981-08-01

For a specific configuration of magnetic field and plasma to be economically attractive as a commercial source of energy, it must contain a high-pressure plasma in a stable fashion while thermally isolating the plasma from the walls of the containment vessel. The tokamak magnetic configuration is presently the most successful in terms of reaching the considered goals. Tokamaks were developed in the USSR in a program initiated in the mid-1950s. By the early 1970s tokamaks were operating not only in the USSR but also in the U.S., Australia, Europe, and Japan. The advanced state of the tokamak program is indicated by the fact that it is used as a testbed for generic fusion development - for auxiliary heating, diagnostics, materials - as well as for specific tokamak advancement. This has occurred because it is the most economic source of a large, reproducible, hot, dense plasma. The basic tokamak is considered along with tokamak improvements, impurity control, additional heating, particle and power balance in a tokamak, aspects of microscopic transport, and macroscopic stability.

Plasma detachment in divertor tokamaks

NASA Astrophysics Data System (ADS)

Leonard, A. W.

2018-04-01

Observations of divertor plasma detachment in tokamaks are reviewed. Plasma detachment is characterized in terms of transport and dissipation of power, momentum and particle flux along the open field lines from the midplane to the divertor. Asymmetries in detachment onset and other characteristics between the inboard and outboard divertor plasmas is found to be primarily driven by plasma E× B drifts. The effect of divertor plate geometry and magnetic configuration on divertor detachment is summarized. Control of divertor detachment has progressed with a development of a number of diagnostics to characterize the detached state in real-time. Finally the compatibility of detached divertor operation with high performance core plasmas is examined.

Explaining Cold-Pulse Dynamics in Tokamak Plasmas Using Local Turbulent Transport Models

DOE PAGES

Rodriguez-Fernandez, P.; White, A. E.; Howard, N. T.; ...

2018-02-16

A long-standing enigma in plasma transport has been resolved by modeling of cold-pulse experiments conducted on the Alcator C-Mod tokamak. Controlled edge cooling of fusion plasmas triggers core electron heating on time scales faster than an energy confinement time, which has long been interpreted as strong evidence of nonlocal transport. Here, this Letter shows that the steady-state profiles, the cold-pulse rise time, and disappearance at higher density as measured in these experiments are successfully captured by a recent local quasilinear turbulent transport model, demonstrating that the existence of nonlocal transport phenomena is not necessary for explaining the behavior and timemore » scales of cold-pulse experiments in tokamak plasmas.« less

Explaining Cold-Pulse Dynamics in Tokamak Plasmas Using Local Turbulent Transport Models

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

Rodriguez-Fernandez, P.; White, A. E.; Howard, N. T.

A long-standing enigma in plasma transport has been resolved by modeling of cold-pulse experiments conducted on the Alcator C-Mod tokamak. Controlled edge cooling of fusion plasmas triggers core electron heating on time scales faster than an energy confinement time, which has long been interpreted as strong evidence of nonlocal transport. Here, this Letter shows that the steady-state profiles, the cold-pulse rise time, and disappearance at higher density as measured in these experiments are successfully captured by a recent local quasilinear turbulent transport model, demonstrating that the existence of nonlocal transport phenomena is not necessary for explaining the behavior and timemore » scales of cold-pulse experiments in tokamak plasmas.« less

Explaining Cold-Pulse Dynamics in Tokamak Plasmas Using Local Turbulent Transport Models

NASA Astrophysics Data System (ADS)

Rodriguez-Fernandez, P.; White, A. E.; Howard, N. T.; Grierson, B. A.; Staebler, G. M.; Rice, J. E.; Yuan, X.; Cao, N. M.; Creely, A. J.; Greenwald, M. J.; Hubbard, A. E.; Hughes, J. W.; Irby, J. H.; Sciortino, F.

2018-02-01

A long-standing enigma in plasma transport has been resolved by modeling of cold-pulse experiments conducted on the Alcator C-Mod tokamak. Controlled edge cooling of fusion plasmas triggers core electron heating on time scales faster than an energy confinement time, which has long been interpreted as strong evidence of nonlocal transport. This Letter shows that the steady-state profiles, the cold-pulse rise time, and disappearance at higher density as measured in these experiments are successfully captured by a recent local quasilinear turbulent transport model, demonstrating that the existence of nonlocal transport phenomena is not necessary for explaining the behavior and time scales of cold-pulse experiments in tokamak plasmas.

Toroidal gyrofluid equations for simulations of tokamak turbulence

NASA Astrophysics Data System (ADS)

Beer, M. A.; Hammett, G. W.

1996-11-01

A set of nonlinear gyrofluid equations for simulations of tokamak turbulence are derived by taking moments of the nonlinear toroidal gyrokinetic equation. The moment hierarchy is closed with approximations that model the kinetic effects of parallel Landau damping, toroidal drift resonances, and finite Larmor radius effects. These equations generalize the work of Dorland and Hammett [Phys. Fluids B 5, 812 (1993)] to toroidal geometry by including essential toroidal effects. The closures for phase mixing from toroidal ∇B and curvature drifts take the basic form presented in Waltz et al. [Phys. Fluids B 4, 3138 (1992)], but here a more rigorous procedure is used, including an extension to higher moments, which provides significantly improved accuracy. In addition, trapped ion effects and collisions are incorporated. This reduced set of nonlinear equations accurately models most of the physics considered important for ion dynamics in core tokamak turbulence, and is simple enough to be used in high resolution direct numerical simulations.

Simulation of MST tokamak discharges with resonant magnetic perturbations

NASA Astrophysics Data System (ADS)

Cornille, B. S.; Sovinec, C. R.; Chapman, B. E.; Dubois, A.; McCollam, K. J.; Munaretto, S.

2016-10-01

Nonlinear MHD modeling of MST tokamak plasmas with an applied resonant magnetic perturbation (RMP) reveals degradation of flux surfaces that may account for the experimentally observed suppression of runaway electrons with the RMP. Runaway electrons are routinely generated in MST tokamak discharges with low plasma density. When an m = 3 RMP is applied these electrons are strongly suppressed, while an m = 1 RMP of comparable amplitude has little effect. The computations are performed using the NIMROD code and use reconstructed equilibrium states of MST tokamak plasmas with q (0) < 1 and q (a) = 2.2 . Linear computations show that the (1 , 1) -kink and (2 , 2) -tearing modes are unstable, and nonlinear simulations produce sawtoothing with a period of approximately 0.5 ms, which is comparable to the period of MHD activity observed experimentally. Adding an m = 3 RMP in the computation degrades flux surfaces in the outer region of the plasma, while no degradation occurs with an m = 1 RMP. The outer flux surface degradation with the m = 3 RMP, combined with the sawtooth-induced distortion of flux surfaces in the core, may account for the observed suppression of runaway electrons. Work supported by DOE Grant DE-FC02-08ER54975.

Transition to subcritical turbulence in a tokamak plasma

NASA Astrophysics Data System (ADS)

van Wyk, F.; Highcock, E. G.; Schekochihin, A. A.; Roach, C. M.; Field, A. R.; Dorland, W.

2016-12-01

Tokamak turbulence, driven by the ion-temperature gradient and occurring in the presence of flow shear, is investigated by means of local, ion-scale, electrostatic gyrokinetic simulations (with both kinetic ions and electrons) of the conditions in the outer core of the Mega-Ampere Spherical Tokamak (MAST). A parameter scan in the local values of the ion-temperature gradient and flow shear is performed. It is demonstrated that the experimentally observed state is near the stability threshold and that this stability threshold is nonlinear: sheared turbulence is subcritical, i.e. the system is formally stable to small perturbations, but, given a large enough initial perturbation, it transitions to a turbulent state. A scenario for such a transition is proposed and supported by numerical results: close to threshold, the nonlinear saturated state and the associated anomalous heat transport are dominated by long-lived coherent structures, which drift across the domain, have finite amplitudes, but are not volume filling; as the system is taken away from the threshold into the more unstable regime, the number of these structures increases until they overlap and a more conventional chaotic state emerges. Whereas this appears to represent a new scenario for transition to turbulence in tokamak plasmas, it is reminiscent of the behaviour of other subcritically turbulent systems, e.g. pipe flows and Keplerian magnetorotational accretion flows.

Development of frequency modulation reflectometer for Korea Superconducting Tokamak Advanced Research tokamak

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

Seo, Seong-Heon; Wi, H. M.; Lee, W. R.

2013-08-15

Frequency modulation reflectometer has been developed to measure the plasma density profile of the Korea Superconducting Tokamak Advanced Research tokamak. Three reflectometers are operating in extraordinary polarization mode in the frequency range of Q band (33.6–54 GHz), V band (48–72 GHz), and W band (72–108 GHz) to measure the density up to 7 × 10{sup 19} m{sup −3} when the toroidal magnetic field is 2 T on axis. The antenna is installed inside of the vacuum vessel. A new vacuum window is developed by using 50 μm thick mica film and 0.1 mm thick gold gasket. The filter bank ofmore » low pass filter, notch filter, and Faraday isolator is used to reject the electron cyclotron heating high power at attenuation of 60 dB. The full frequency band is swept in 20 μs. The mixer output is directly digitized with sampling rate of 100 MSamples/s. The phase is obtained by using wavelet transform. The whole hardware and software system is described in detail and the measured density profile is presented as a result.« less

A fission-fusion hybrid reactor in steady-state L-mode tokamak configuration with natural uranium

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

Reed, Mark; Parker, Ronald R.; Forget, Benoit

2012-06-19

This work develops a conceptual design for a fusion-fission hybrid reactor operating in steady-state L-mode tokamak configuration with a subcritical natural or depleted uranium pebble bed blanket. A liquid lithium-lead alloy breeds enough tritium to replenish that consumed by the D-T fusion reaction. The fission blanket augments the fusion power such that the fusion core itself need not have a high power gain, thus allowing for fully non-inductive (steady-state) low confinement mode (L-mode) operation at relatively small physical dimensions. A neutron transport Monte Carlo code models the natural uranium fission blanket. Maximizing the fission power gain while breeding sufficient tritiummore » allows for the selection of an optimal set of blanket parameters, which yields a maximum prudent fission power gain of approximately 7. A 0-D tokamak model suffices to analyze approximate tokamak operating conditions. This fission blanket would allow the fusion component of a hybrid reactor with the same dimensions as ITER to operate in steady-state L-mode very comfortably with a fusion power gain of 6.7 and a thermal fusion power of 2.1 GW. Taking this further can determine the approximate minimum scale for a steady-state L-mode tokamak hybrid reactor, which is a major radius of 5.2 m and an aspect ratio of 2.8. This minimum scale device operates barely within the steady-state L-mode realm with a thermal fusion power of 1.7 GW. Basic thermal hydraulic analysis demonstrates that pressurized helium could cool the pebble bed fission blanket with a flow rate below 10 m/s. The Brayton cycle thermal efficiency is 41%. This reactor, dubbed the Steady-state L-mode non-Enriched Uranium Tokamak Hybrid (SLEUTH), with its very fast neutron spectrum, could be superior to pure fission reactors in terms of breeding fissile fuel and transmuting deleterious fission products. It would likely function best as a prolific plutonium breeder, and the plutonium it produces could actually be more

A fission-fusion hybrid reactor in steady-state L-mode tokamak configuration with natural uranium

NASA Astrophysics Data System (ADS)

Reed, Mark; Parker, Ronald R.; Forget, Benoit

2012-06-01

This work develops a conceptual design for a fusion-fission hybrid reactor operating in steady-state L-mode tokamak configuration with a subcritical natural or depleted uranium pebble bed blanket. A liquid lithium-lead alloy breeds enough tritium to replenish that consumed by the D-T fusion reaction. The fission blanket augments the fusion power such that the fusion core itself need not have a high power gain, thus allowing for fully non-inductive (steady-state) low confinement mode (L-mode) operation at relatively small physical dimensions. A neutron transport Monte Carlo code models the natural uranium fission blanket. Maximizing the fission power gain while breeding sufficient tritium allows for the selection of an optimal set of blanket parameters, which yields a maximum prudent fission power gain of approximately 7. A 0-D tokamak model suffices to analyze approximate tokamak operating conditions. This fission blanket would allow the fusion component of a hybrid reactor with the same dimensions as ITER to operate in steady-state L-mode very comfortably with a fusion power gain of 6.7 and a thermal fusion power of 2.1 GW. Taking this further can determine the approximate minimum scale for a steady-state L-mode tokamak hybrid reactor, which is a major radius of 5.2 m and an aspect ratio of 2.8. This minimum scale device operates barely within the steady-state L-mode realm with a thermal fusion power of 1.7 GW. Basic thermal hydraulic analysis demonstrates that pressurized helium could cool the pebble bed fission blanket with a flow rate below 10 m/s. The Brayton cycle thermal efficiency is 41%. This reactor, dubbed the Steady-state L-mode non-Enriched Uranium Tokamak Hybrid (SLEUTH), with its very fast neutron spectrum, could be superior to pure fission reactors in terms of breeding fissile fuel and transmuting deleterious fission products. It would likely function best as a prolific plutonium breeder, and the plutonium it produces could actually be more

Dynamic diagnostics of the error fields in tokamaks

NASA Astrophysics Data System (ADS)

Pustovitov, V. D.

2007-07-01

The error field diagnostics based on magnetic measurements outside the plasma is discussed. The analysed methods rely on measuring the plasma dynamic response to the finite-amplitude external magnetic perturbations, which are the error fields and the pre-programmed probing pulses. Such pulses can be created by the coils designed for static error field correction and for stabilization of the resistive wall modes, the technique developed and applied in several tokamaks, including DIII-D and JET. Here analysis is based on the theory predictions for the resonant field amplification (RFA). To achieve the desired level of the error field correction in tokamaks, the diagnostics must be sensitive to signals of several Gauss. Therefore, part of the measurements should be performed near the plasma stability boundary, where the RFA effect is stronger. While the proximity to the marginal stability is important, the absolute values of plasma parameters are not. This means that the necessary measurements can be done in the diagnostic discharges with parameters below the nominal operating regimes, with the stability boundary intentionally lowered. The estimates for ITER are presented. The discussed diagnostics can be tested in dedicated experiments in existing tokamaks. The diagnostics can be considered as an extension of the 'active MHD spectroscopy' used recently in the DIII-D tokamak and the EXTRAP T2R reversed field pinch.

Advanced Tokamak Stability Theory

NASA Astrophysics Data System (ADS)

Zheng, Linjin

2015-03-01

The intention of this book is to introduce advanced tokamak stability theory. We start with the derivation of the Grad-Shafranov equation and the construction of various toroidal flux coordinates. An analytical tokamak equilibrium theory is presented to demonstrate the Shafranov shift and how the toroidal hoop force can be balanced by the application of a vertical magnetic field in tokamaks. In addition to advanced theories, this book also discusses the intuitive physics pictures for various experimentally observed phenomena.

Integrated tokamak modeling: when physics informs engineering and research planning

NASA Astrophysics Data System (ADS)

Poli, Francesca

2017-10-01

Simulations that integrate virtually all the relevant engineering and physics aspects of a real tokamak experiment are a power tool for experimental interpretation, model validation and planning for both present and future devices. This tutorial will guide through the building blocks of an ``integrated'' tokamak simulation, such as magnetic flux diffusion, thermal, momentum and particle transport, external heating and current drive sources, wall particle sources and sinks. Emphasis is given to the connection and interplay between external actuators and plasma response, between the slow time scales of the current diffusion and the fast time scales of transport, and how reduced and high-fidelity models can contribute to simulate a whole device. To illustrate the potential and limitations of integrated tokamak modeling for discharge prediction, a helium plasma scenario for the ITER pre-nuclear phase is taken as an example. This scenario presents challenges because it requires core-edge integration and advanced models for interaction between waves and fast-ions, which are subject to a limited experimental database for validation and guidance. Starting from a scenario obtained by re-scaling parameters from the demonstration inductive ``ITER baseline'', it is shown how self-consistent simulations that encompass both core and edge plasma regions, as well as high-fidelity heating and current drive source models are needed to set constraints on the density, magnetic field and heating scheme. This tutorial aims at demonstrating how integrated modeling, when used with adequate level of criticism, can not only support design of operational scenarios, but also help to asses the limitations and gaps in the available models, thus indicating where improved modeling tools are required and how present experiments can help their validation and inform research planning. Work supported by DOE under DE-AC02-09CH1146.

Phase locking of multi-helicity neoclassical tearing modes in tokamak plasmas

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

Fitzpatrick, Richard

2015-04-15

The attractive “hybrid” tokamak scenario combines comparatively high q{sub 95} operation with improved confinement compared with the conventional H{sub 98,y2} scaling law. Somewhat unusually, hybrid discharges often exhibit multiple neoclassical tearing modes (NTMs) possessing different mode numbers. The various NTMs are eventually observed to phase lock to one another, giving rise to a significant flattening, or even an inversion, of the core toroidal plasma rotation profile. This behavior is highly undesirable because the loss of core plasma rotation is known to have a deleterious effect on plasma stability. This paper presents a simple, single-fluid, cylindrical model of the phase lockingmore » of two NTMs with different poloidal and toroidal mode numbers in a tokamak plasma. Such locking takes place via a combination of nonlinear three-wave coupling and conventional toroidal coupling. In accordance with experimental observations, the model predicts that there is a bifurcation to a phase-locked state when the frequency mismatch between the modes is reduced to one half of its original value. In further accordance, the phase-locked state is characterized by the permanent alignment of one of the X-points of NTM island chains on the outboard mid-plane of the plasma, and a modified toroidal angular velocity profile, interior to the outermost coupled rational surface, which is such that the core rotation is flattened, or even inverted.« less

T-L Plane Abstraction-Based Energy-Efficient Real-Time Scheduling for Multi-Core Wireless Sensors

PubMed Central

Kim, Youngmin; Lee, Ki-Seong; Pham, Ngoc-Son; Lee, Sun-Ro; Lee, Chan-Gun

2016-01-01

Energy efficiency is considered as a critical requirement for wireless sensor networks. As more wireless sensor nodes are equipped with multi-cores, there are emerging needs for energy-efficient real-time scheduling algorithms. The T-L plane-based scheme is known to be an optimal global scheduling technique for periodic real-time tasks on multi-cores. Unfortunately, there has been a scarcity of studies on extending T-L plane-based scheduling algorithms to exploit energy-saving techniques. In this paper, we propose a new T-L plane-based algorithm enabling energy-efficient real-time scheduling on multi-core sensor nodes with dynamic power management (DPM). Our approach addresses the overhead of processor mode transitions and reduces fragmentations of the idle time, which are inherent in T-L plane-based algorithms. Our experimental results show the effectiveness of the proposed algorithm compared to other energy-aware scheduling methods on T-L plane abstraction. PMID:27399722

T-L Plane Abstraction-Based Energy-Efficient Real-Time Scheduling for Multi-Core Wireless Sensors.

PubMed

Kim, Youngmin; Lee, Ki-Seong; Pham, Ngoc-Son; Lee, Sun-Ro; Lee, Chan-Gun

2016-07-08

Energy efficiency is considered as a critical requirement for wireless sensor networks. As more wireless sensor nodes are equipped with multi-cores, there are emerging needs for energy-efficient real-time scheduling algorithms. The T-L plane-based scheme is known to be an optimal global scheduling technique for periodic real-time tasks on multi-cores. Unfortunately, there has been a scarcity of studies on extending T-L plane-based scheduling algorithms to exploit energy-saving techniques. In this paper, we propose a new T-L plane-based algorithm enabling energy-efficient real-time scheduling on multi-core sensor nodes with dynamic power management (DPM). Our approach addresses the overhead of processor mode transitions and reduces fragmentations of the idle time, which are inherent in T-L plane-based algorithms. Our experimental results show the effectiveness of the proposed algorithm compared to other energy-aware scheduling methods on T-L plane abstraction.

Edge turbulence effect on ultra-fast swept reflectometry core measurements in tokamak plasmas

NASA Astrophysics Data System (ADS)

Zadvitskiy, G. V.; Heuraux, S.; Lechte, C.; Hacquin, S.; Sabot, R.

2018-02-01

Ultra-fast frequency-swept reflectometry (UFSR) enables one to provide information about the turbulence radial wave-number spectrum and perturbation amplitude with good spatial and temporal resolutions. However, a data interpretation of USFR is quiet tricky. An iterative algorithm to solve this inverse problem was used in past works, Gerbaud (2006 Rev. Sci. Instrum. 77 10E928). For a direct solution, a fast 1D Helmholtz solver was used. Two-dimensional effects are strong and should be taken into account during data interpretation. As 2D full-wave codes are still too time consuming for systematic application, fast 2D approaches based on the Born approximation are of prime interest. Such methods gives good results in the case of small turbulence levels. However in tokamak plasmas, edge turbulence is usually very strong and can distort and broaden the probing beam Sysoeva et al (2015 Nucl. Fusion 55 033016). It was shown that this can change reflectometer phase response from the plasma core. Comparison between 2D full wave computation and the simplified Born approximation was done. The approximated method can provide a right spectral shape, but it is unable to describe a change of the spectral amplitude with an edge turbulence level. Computation for the O-mode wave with the linear density profile in the slab geometry and for realistic Tore-Supra density profile, based on the experimental data turbulence amplitude and spectrum, were performed to investigate the role of strong edge turbulence. It is shown that the spectral peak in the signal amplitude variation spectrum which rises with edge turbulence can be a signature of strong edge turbulence. Moreover, computations for misaligned receiving and emitting antennas were performed. It was found that the signal amplitude variation peak changes its position with a receiving antenna poloidal displacement.

Plasma detachment in divertor tokamaks

DOE PAGES

Leonard, A. W.

2018-02-07

In this study, observations of divertor plasma detachment in tokamaks are reviewed. Plasma detachment is characterized in terms of transport and dissipation of power, momentum and particle flux along the open field lines from the midplane to the divertor. Asymmetries in detachment onset and other characteristics between the inboard and outboard divertor plasmas is found to be primarily driven by plasmamore » $$\\vec{E}$$ x $$\\vec{B}$$ drifts. The effect of divertor plate geometry and magnetic configuration on divertor detachment is summarized. Control of divertor detachment has progressed with a development of a number of diagnostics to characterize the detached state in real-time. Finally the compatibility of detached divertor operation with high performance core plasmas is examined.« less

Plasma detachment in divertor tokamaks

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

Leonard, A. W.

In this study, observations of divertor plasma detachment in tokamaks are reviewed. Plasma detachment is characterized in terms of transport and dissipation of power, momentum and particle flux along the open field lines from the midplane to the divertor. Asymmetries in detachment onset and other characteristics between the inboard and outboard divertor plasmas is found to be primarily driven by plasmamore » $$\\vec{E}$$ x $$\\vec{B}$$ drifts. The effect of divertor plate geometry and magnetic configuration on divertor detachment is summarized. Control of divertor detachment has progressed with a development of a number of diagnostics to characterize the detached state in real-time. Finally the compatibility of detached divertor operation with high performance core plasmas is examined.« less

Repacking the Core of T4 lysozyme by automated design.

PubMed

Mooers, Blaine H M; Datta, Deepshikha; Baase, Walter A; Zollars, Eric S; Mayo, Stephen L; Matthews, Brian W

2003-09-19

Automated protein redesign, as implemented in the program ORBIT, was used to redesign the core of phage T4 lysozyme. A total of 26 buried or partially buried sites in the C-terminal domain were allowed to vary both their sequence and side-chain conformation while the backbone and non-selected side-chains remained fixed. A variant with seven substitutions ("Core-7") was identified as having the most favorable energy. The redesign experiment was repeated with a penalty for the presence of methionine residues. In this case the redesigned protein ("Core-10") had ten amino acid changes. The two designed proteins, as well as the constituent single mutants, and several single-site revertants were over-expressed in Escherichia coli, purified, and subjected to crystallographic and thermal analyses. The thermodynamic and structural data show that some repacking was achieved although neither redesigned protein was more stable than the wild-type protein. The use of the methionine penalty was shown to be effective. Several of the side-chain rotamers in the predicted structure of Core-10 differ from those observed. Rather than changing to new rotamers predicted by the design process, side-chains tend to maintain conformations similar to those seen in the native molecule. In contrast, parts of the backbone change by up to 2.8A relative to both the designed structure and wild-type. Water molecules that are present within the lysozyme molecule were removed during the design process. In the redesigned protein the resultant cavities were, to some degree, re-occupied by side-chain atoms. In the observed structure, however, water molecules were still bound at or near their original sites. This suggests that it may be preferable to leave such water molecules in place during the design procedure. The results emphasize the specificity of the packing that occurs within the core of a typical protein. While point substitutions within the core are tolerated they almost always result in a loss

On the breakdown modes and parameter space of Ohmic Tokamak startup

NASA Astrophysics Data System (ADS)

Peng, Yanli; Jiang, Wei; Zhang, Ya; Hu, Xiwei; Zhuang, Ge; Innocenti, Maria; Lapenta, Giovanni

2017-10-01

Tokamak plasma has to be hot. The process of turning the initial dilute neutral hydrogen gas at room temperature into fully ionized plasma is called tokamak startup. Even with over 40 years of research, the parameter ranges for the successful startup still aren't determined by numerical simulations but by trial and errors. However, in recent years it has drawn much attention due to one of the challenges faced by ITER: the maximum electric field for startup can't exceed 0.3 V/m, which makes the parameter range for successful startup narrower. Besides, this physical mechanism is far from being understood either theoretically or numerically. In this work, we have simulated the plasma breakdown phase driven by pure Ohmic heating using a particle-in-cell/Monte Carlo code, with the aim of giving a predictive parameter range for most tokamaks, even for ITER. We have found three situations during the discharge, as a function of the initial parameters: no breakdown, breakdown and runaway. Moreover, breakdown delay and volt-second consumption under different initial conditions are evaluated. In addition, we have simulated breakdown on ITER and confirmed that when the electric field is 0.3 V/m, the optimal pre-filling pressure is 0.001 Pa, which is in good agreement with ITER's design.

Energetic particles in spherical tokamak plasmas

NASA Astrophysics Data System (ADS)

McClements, K. G.; Fredrickson, E. D.

2017-05-01

Spherical tokamaks (STs) typically have lower magnetic fields than conventional tokamaks, but similar mass densities. Suprathermal ions with relatively modest energies, in particular beam-injected ions, consequently have speeds close to or exceeding the Alfvén velocity, and can therefore excite a range of Alfvénic instabilities which could be driven by (and affect the behaviour of) fusion α-particles in a burning plasma. STs heated with neutral beams, including the small tight aspect ratio tokamak (START), the mega amp spherical tokamak (MAST), the national spherical torus experiment (NSTX) and Globus-M, have thus provided an opportunity to study toroidal Alfvén eigenmodes (TAEs), together with higher frequency global Alfvén eigenmodes (GAEs) and compressional Alfvén eigenmodes (CAEs), which could affect beam current drive and channel fast ion energy into bulk ions in future devices. In NSTX GAEs were correlated with a degradation of core electron energy confinement. In MAST pulses with reduced magnetic field, CAEs were excited across a wide range of frequencies, extending to the ion cyclotron range, but were suppressed when hydrogen was introduced to the deuterium plasma, apparently due to mode conversion at ion-ion hybrid resonances. At lower frequencies fishbone instabilities caused fast particle redistribution in some MAST and NSTX pulses, but this could be avoided by moving the neutral beam line away from the magnetic axis or by operating the plasma at either high density or elevated safety factor. Fast ion redistribution has been observed during GAE avalanches on NSTX, while in both NSTX and MAST fast ions were transported by saturated kink modes, sawtooth crashes, resonant magnetic perturbations and TAEs. The energy dependence of fast ion redistribution due to both sawteeth and TAEs has been studied in Globus-M. High energy charged fusion products are unconfined in present-day STs, but have been shown in MAST to provide a useful diagnostic of beam ion

Energetic particles in spherical tokamak plasmas

DOE PAGES

McClements, K. G.; Fredrickson, E. D.

2017-03-21

Spherical tokamaks (STs) typically have lower magnetic fields than conventional tokamaks, but similar mass densities. Suprathermal ions with relatively modest energies, in particular beam-injected ions, consequently have speeds close to or exceeding the Alfvén velocity, and can therefore excite a range of Alfvénic instabilities which could be driven by (and affect the behaviour of) fusion α-particles in a burning plasma. STs heated with neutral beams, including the small tight aspect ratio tokamak (START), the mega amp spherical tokamak (MAST), the national spherical torus experiment (NSTX) and Globus-M, have thus provided an opportunity to study toroidal Alfvén eigenmodes (TAEs), together withmore » higher frequency global Alfvén eigenmodes (GAEs) and compressional Alfvén eigenmodes (CAEs), which could affect beam current drive and channel fast ion energy into bulk ions in future devices. In NSTX GAEs were correlated with a degradation of core electron energy confinement. In MAST pulses with reduced magnetic field, CAEs were excited across a wide range of frequencies, extending to the ion cyclotron range, but were suppressed when hydrogen was introduced to the deuterium plasma, apparently due to mode conversion at ion–ion hybrid resonances. At lower frequencies fishbone instabilities caused fast particle redistribution in some MAST and NSTX pulses, but this could be avoided by moving the neutral beam line away from the magnetic axis or by operating the plasma at either high density or elevated safety factor. Fast ion redistribution has been observed during GAE avalanches on NSTX, while in both NSTX and MAST fast ions were transported by saturated kink modes, sawtooth crashes, resonant magnetic perturbations and TAEs. The energy dependence of fast ion redistribution due to both sawteeth and TAEs has been studied in Globus-M. High energy charged fusion products are unconfined in present-day STs, but have been shown in MAST to provide a useful diagnostic of

Modelling the power deposition into a spherical tokamak fusion power plant

NASA Astrophysics Data System (ADS)

Windsor, C. G.; Morgan, J. G.; Buxton, P. F.; Costley, A. E.; Smith, G. D. W.; Sykes, A.

2017-03-01

Numerical studies have been made to improve the performance of the central column of a superconducting spherical tokamak fusion pilot plant. The assumed neutron shield includes concentric layers of tungsten carbide and water. The relative thickness of the water layers was varied and a minimum power deposition was found at about 17% of water. It was found advantageous to have an approximately 1.7 times thicker water layer next to the core and a similarly thinner layer next to the plasma. The use of tungsten boride instead of tungsten carbide was shown to make an improvement especially if placed close to the central superconducting core, the inner layer alone reducing the power deposition by 29%. Engineering features such as a central steel tie-bar, an insulating thermal vacuum gap, a wall gap next to the plasma and knowledge of the vertical energy distribution are essential to a successful design and their effects on the power deposition are shown in an appendix. The results have been fitted to model distributions and incorporated into the Tokamak Energy System Code, which can then give predictions of the power deposition as a function of other parameters such as the plasma major radius and the maximum magnetic field permitted on the superconductors.

CD4+ Primary T Cells Expressing HCV-Core Protein Upregulate Foxp3 and IL-10, Suppressing CD4 and CD8 T Cells

PubMed Central

Aguado, Enrique; Garcia-Cozar, Francisco

2014-01-01

Adaptive T cell responses are critical for controlling HCV infection. While there is clinical evidence of a relevant role for regulatory T cells in chronic HCV-infected patients, based on their increased number and function; mechanisms underlying such a phenomena are still poorly understood. Accumulating evidence suggests that proteins from Hepatitis C virus can suppress host immune responses. We and others have shown that HCV is present in CD4+ lymphocytes from chronically infected patients and that HCV-core protein induces a state of unresponsiveness in the CD4+ tumor cell line Jurkat. Here we show that CD4+ primary T cells lentivirally transduced with HCV-core, not only acquire an anergic phenotype but also inhibit IL-2 production and proliferation of bystander CD4+ or CD8+ T cells in response to anti-CD3 plus anti-CD28 stimulation. Core-transduced CD4+ T cells show a phenotype characterized by an increased basal secretion of the regulatory cytokine IL-10, a decreased IFN-γ production upon stimulation, as well as expression of regulatory T cell markers, CTLA-4, and Foxp3. A significant induction of CD4+CD25+CD127lowPD-1highTIM-3high regulatory T cells with an exhausted phenotype was also observed. Moreover, CCR7 expression decreased in HCV-core expressing CD4+ T cells explaining their sequestration in inflamed tissues such as the infected liver. This work provides a new perspective on de novo generation of regulatory CD4+ T cells in the periphery, induced by the expression of a single viral protein. PMID:24465502

Parasitic momentum flux in the tokamak core

NASA Astrophysics Data System (ADS)

Stoltzfus-Dueck, T.

2017-10-01

Tokamak plasmas rotate spontaneously without applied torque. This intrinsic rotation is important for future low-torque devices such as ITER, since rotation stabilizes certain instabilities. In the mid-radius `gradient region,' which reaches from the sawtooth inversion radius out to the pedestal top, intrinsic rotation profiles may be either flat or hollow, and can transition suddenly between these two states, an unexplained phenomenon referred to as rotation reversal. Theoretical efforts to explain the mid-radius rotation shear have largely focused on quasilinear models, in which the phase relationships of some selected instability result in a nondiffusive momentum flux (``residual stress''). In contrast, the present work demonstrates the existence of a robust, fully nonlinear symmetry-breaking momentum flux that follows from the free-energy flow in phase space and does not depend on any assumed linear eigenmode structure. The physical origin is an often-neglected portion of the radial ExB drift, which is shown to drive a symmetry-breaking outward flux of co-current momentum whenever free energy is transferred from the electrostatic potential to ion parallel flows. The fully nonlinear derivation relies only on conservation properties and symmetry, thus retaining the important contribution of damped modes. The resulting rotation peaking is counter-current and scales as temperature over plasma current. As first demonstrated by Landau, this free-energy transfer (thus also the corresponding residual stress) becomes inactive when frequencies are much higher than the ion transit frequency, which allows sudden transitions between hollow and flat profiles. Simple estimates suggest that this mechanism may be consistent with experimental observations. This work was funded in part by the Max-Planck/Princeton Center for Plasma Physics and in part by the U.S. Dept. of Energy, Office of Science, Contract No. DE-AC02-09CH11466.

Design of tangential multi-energy SXR cameras for tokamak plasmas

NASA Astrophysics Data System (ADS)

Yamazaki, H.; Delgado-Aparicio, L. F.; Pablant, N.; Hill, K.; Bitter, M.; Takase, Y.; Ono, M.; Stratton, B.

2017-10-01

A new synthetic diagnostic capability has been built to study the response of tangential multi-energy soft x-ray pin-hole cameras for arbitrary plasma densities (ne , D), temperature (Te) and ion concentrations (nZ). For tokamaks and future facilities to operate safely in a high-pressure long-pulse discharge, it is imperative to address key issues associated with impurity sources, core transport and high-Z impurity accumulation. Multi-energy soft xray imaging provides a unique opportunity for measuring, simultaneously, a variety of important plasma properties (e.g. Te, nZ and ΔZeff). These systems are designed to sample the continuum- and line-emission from low- to high-Z impurities (e.g. C, O, Al, Si, Ar, Ca, Fe, Ni and Mo) in multiple energy-ranges. These x-ray cameras will be installed in the MST-RFP, as well as NSTX-U and DIII-D tokamaks, measuring the radial structure of the photon emissivity with a radial resolution below 1 cm at a 500 Hz frame rate and a photon-energy resolution of 500 eV. The layout and response expected for the new systems will be shown for different plasma conditions and impurity concentrations. The effect of toroidal rotation driving poloidal asymmetries in the core radiation is also addressed for the case of NSTX-U.

Georgia Tech Studies of Sub-Critical Advanced Burner Reactors with a D-T Fusion Tokamak Neutron Source for the Transmutation of Spent Nuclear Fuel

NASA Astrophysics Data System (ADS)

Stacey, W. M.

2009-09-01

The possibility that a tokamak D-T fusion neutron source, based on ITER physics and technology, could be used to drive sub-critical, fast-spectrum nuclear reactors fueled with the transuranics (TRU) in spent nuclear fuel discharged from conventional nuclear reactors has been investigated at Georgia Tech in a series of studies which are summarized in this paper. It is found that sub-critical operation of such fast transmutation reactors is advantageous in allowing longer fuel residence time, hence greater TRU burnup between fuel reprocessing stages, and in allowing higher TRU loading without compromising safety, relative to what could be achieved in a similar critical transmutation reactor. The required plasma and fusion technology operating parameter range of the fusion neutron source is generally within the anticipated operational range of ITER. The implications of these results for fusion development policy, if they hold up under more extensive and detailed analysis, is that a D-T fusion tokamak neutron source for a sub-critical transmutation reactor, built on the basis of the ITER operating experience, could possibly be a logical next step after ITER on the path to fusion electrical power reactors. At the same time, such an application would allow fusion to contribute to meeting the nation's energy needs at an earlier stage by helping to close the fission reactor nuclear fuel cycle.

A novel flexible field-aligned coordinate system for tokamak edge plasma simulation

NASA Astrophysics Data System (ADS)

Leddy, J.; Dudson, B.; Romanelli, M.; Shanahan, B.; Walkden, N.

2017-03-01

Tokamak plasmas are confined by a magnetic field that limits the particle and heat transport perpendicular to the field. Parallel to the field the ionised particles can move freely, so to obtain confinement the field lines are "closed" (i.e. form closed surfaces of constant poloidal flux) in the core of a tokamak. Towards, the edge, however, the field lines intersect physical surfaces, leading to interaction between neutral and ionised particles, and the potential melting of the material surface. Simulation of this interaction is important for predicting the performance and lifetime of future tokamak devices such as ITER. Field-aligned coordinates are commonly used in the simulation of tokamak plasmas due to the geometry and magnetic topology of the system. However, these coordinates are limited in the geometry they allow in the poloidal plane due to orthogonality requirements. A novel 3D coordinate system is proposed herein that relaxes this constraint so that any arbitrary, smoothly varying geometry can be matched in the poloidal plane while maintaining a field-aligned coordinate. This system is implemented in BOUT++ and tested for accuracy using the method of manufactured solutions. A MAST edge cross-section is simulated using a fluid plasma model and the results show expected behaviour for density, temperature, and velocity. Finally, simulations of an isolated divertor leg are conducted with and without neutrals to demonstrate the ion-neutral interaction near the divertor plate and the corresponding beneficial decrease in plasma temperature.

Design, simulation and construction of the Taban tokamak

NASA Astrophysics Data System (ADS)

H, R. MIRZAEI; R, AMROLLAHI

2018-04-01

This paper describes the design and construction of the Taban tokamak, which is located in Amirkabir University of Technology, Tehran, Iran. The Taban tokamak was designed for plasma investigation. The design, simulation and construction of essential parts of the Taban tokamak such as the toroidal field (TF) system, ohmic heating (OH) system and equilibrium field system and their power supplies are presented. For the Taban tokamak, the toroidal magnetic coil was designed to produce a maximum field of 0.7 T at R = 0.45 m. The power supply of the TF was a 130 kJ, 0–10 kV capacitor bank. Ripples of toroidal magnetic field at the plasma edge and plasma center are 0.2% and 0.014%, respectively. For the OH system with 3 kA current, the stray field in the plasma region is less than 40 G over 80% of the plasma volume. The power supply of the OH system consists of two stages, as follows. The fast bank stage is a 120 μF, 0–5 kV capacitor that produces 2.5 kA in 400 μs and the slow bank stage is 93 mF, 600 V that can produce a maximum of 3 kA. The equilibrium system can produce uniform magnetic field at plasma volume. This system’s power supply, like the OH system, consists of two stages, so that the fast bank stage is 500 μF, 800 V and the slow bank stage is 110 mF, 200 V.

Can tokamaks PFC survive a single event of any plasma instabilities?

NASA Astrophysics Data System (ADS)

Hassanein, A.; Sizyuk, V.; Miloshevsky, G.; Sizyuk, T.

2013-07-01

Plasma instability events such as disruptions, edge-localized modes (ELMs), runaway electrons (REs), and vertical displacement events (VDEs) are continued to be serious events and most limiting factors for successful tokamak reactor concept. The plasma-facing components (PFCs), e.g., wall, divertor, and limited surfaces of a tokamak as well as coolant structure materials are subjected to intense particle and heat loads and must maintain a clean and stable surface environment among them and the core/edge plasma. Typical ITER transient events parameters are used for assessing the damage from these four different instability events. HEIGHTS simulation showed that a single event of a disruption, giant ELM, VDE, or RE can cause significant surface erosion (melting and vaporization) damage to PFC, nearby components, and/or structural materials (VDE, RE) melting and possible burnout of coolant tubes that could result in shut down of reactor for extended repair time.

The circuit of polychromator for Experimental Advanced Superconducting Tokamak edge Thomson scattering diagnostic.

PubMed

Zang, Qing; Hsieh, C L; Zhao, Junyu; Chen, Hui; Li, Fengjuan

2013-09-01

The detector circuit is the core component of filter polychromator which is used for scattering light analysis in Thomson scattering diagnostic, and is responsible for the precision and stability of a system. High signal-to-noise and stability are primary requirements for the diagnostic. Recently, an upgraded detector circuit for weak light detecting in Experimental Advanced Superconducting Tokamak (EAST) edge Thomson scattering system has been designed, which can be used for the measurement of large electron temperature (T(e)) gradient and low electron density (n(e)). In this new circuit, a thermoelectric-cooled avalanche photodiode with the aid circuit is involved for increasing stability and enhancing signal-to-noise ratio (SNR), especially the circuit will never be influenced by ambient temperature. These features are expected to improve the accuracy of EAST Thomson diagnostic dramatically. Related mechanical construction of the circuit is redesigned as well for heat-sinking and installation. All parameters are optimized, and SNR is dramatically improved. The number of minimum detectable photons is only 10.

Tokamak reactor for treating fertile material or waste nuclear by-products

DOEpatents

Kotschenreuther, Michael T.; Mahajan, Swadesh M.; Valanju, Prashant M.

2012-10-02

Disclosed is a tokamak reactor. The reactor includes a first toroidal chamber, current carrying conductors, at least one divertor plate within the first toroidal chamber and a second chamber adjacent to the first toroidal chamber surrounded by a section that insulates the reactor from neutrons. The current carrying conductors are configured to confine a core plasma within enclosed walls of the first toroidal chamber such that the core plasma has an elongation of 1.5 to 4 and produce within the first toroidal chamber at least one stagnation point at a perpendicular distance from an equatorial plane through the core plasma that is greater than the plasma minor radius. The at least one divertor plate and current carrying conductors are configured relative to one another such that the current carrying conductors expand the open magnetic field lines at the divertor plate.

Largescale Long-term particle Simulations of Runaway electrons in Tokamaks

NASA Astrophysics Data System (ADS)

Liu, Jian; Qin, Hong; Wang, Yulei

2016-10-01

To understand runaway dynamical behavior is crucial to assess the safety of tokamaks. Though many important analytical and numerical results have been achieved, the overall dynamic behaviors of runaway electrons in a realistic tokamak configuration is still rather vague. In this work, the secular full-orbit simulations of runaway electrons are carried out based on a relativistic volume-preserving algorithm. Detailed phase-space behaviors of runaway electrons are investigated in different timescales spanning 11 orders. A detailed analysis of the collisionless neoclassical scattering is provided when considering the coupling between the rotation of momentum vector and the background field. In large timescale, the initial condition of runaway electrons in phase space globally influences the runaway distribution. It is discovered that parameters and field configuration of tokamaks can modify the runaway electron dynamics significantly. Simulations on 10 million cores of supercomputer using the APT code have been completed. A resolution of 107 in phase space is used, and simulations are performed for 1011 time steps. Largescale simulations show that in a realistic fusion reactor, the concern of runaway electrons is not as serious as previously thought. This research was supported by National Magnetic Connement Fusion Energy Research Project (2015GB111003, 2014GB124005), the National Natural Science Foundation of China (NSFC-11575185, 11575186) and the GeoAlgorithmic Plasma Simulator (GAPS) Project.

Physics evaluation of compact tokamak ignition experiments

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

Uckan, N.A.; Houlberg, W.A.; Sheffield, J.

1985-01-01

At present, several approaches for compact, high-field tokamak ignition experiments are being considered. A comprehensive method for analyzing the potential physics operating regimes and plasma performance characteristics of such ignition experiments with O-D (analytic) and 1-1/2-D (WHIST) transport models is presented. The results from both calculations are in agreement and show that there are regimes in parameter space in which a class of small (R/sub o/ approx. 1-2 m), high-field (B/sub o/ approx. 8-13 T) tokamaks with aB/sub o/S/q/sub */ approx. 25 +- 5 and kappa = b/a approx. 1.6-2.0 appears ignitable for a reasonable range of transport assumptions. Consideringmore » both the density and beta limits, an evaluation of the performance is presented for various forms of chi/sub e/ and chi/sub i/, including degradation at high power and sawtooth activity. The prospects of ohmic ignition are also examined. 16 refs., 13 figs.« less

Measurements of impurity concentrations and transport in the Lithium Tokamak Experiment

NASA Astrophysics Data System (ADS)

Boyle, Dennis Patrick

This thesis presents new measurements of core impurity concentrations and transport in plasmas with lithium coatings on all-metal plasma facing components (PFCs) in the Lithium Tokamak Experiment (LTX). LTX is a modest-sized spherical tokamak uniquely capable of operating with large area solid and/or liquid lithium coatings essentially surrounding the entire plasma (as opposed to just the divertor or limiter region in other devices). Lithium (Li) wall-coatings have improved plasma performance and confinement in several tokamaks with carbon (C) PFCs, including the National Spherical Torus Experiment (NSTX). In NSTX, contamination of the core plasma with Li impurities was very low (<0.1%) despite extensive divertor coatings. Low Li levels in NSTX were found to be largely due to neoclassical forces from the high level of C impurities. Studying impurity levels and transport with Li coatings on stainless steel surfaces in LTX is relevant to future devices (including future enhancements to NSTX-Upgrade) with all-metal PFCs. The new measurements in this thesis were enabled by a refurbished Thomson scattering system and improved impurity spectroscopy, primarily using a novel visible spectrometer monitoring several Li, C, and oxygen (O) emission lines. A simple model was used to account for impurities in unmeasured charge states, assuming constant density in the plasma core and constant concentration in the edge. In discharges with solid Li coatings, volume averaged impurity concentrations were low but non-negligible, with 2-4% Li, 0.6-2% C, 0.4-0.7% O, and Z eff<1.2. Transport was assessed using the TRANSP, NCLASS, and MIST codes. Collisions with the main H ions dominated the neoclassical impurity transport, unlike in NSTX, where collisions with C dominated. Furthermore, neoclassical transport coefficients calculated with NCLASS were similar across all impurity species and differed no more than a factor of two, in contrast to NSTX where they differed by an order of

Measurements of impurity concentrations and transport in the Lithium Tokamak Experiment

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

Boyle, Dennis Patrick

This thesis presents new measurements of core impurity concentrations and transport in plasmas with lithium coatings on all-metal plasma facing components (PFCs) in the Lithium Tokamak Experiment (LTX). LTX is a modest-sized spherical tokamak uniquely capable of operating with large area solid and/or liquid lithium coatings essentially surrounding the entire plasma (as opposed to just the divertor or limiter region in other devices). Lithium (Li) wall-coatings have improved plasma performance and confinement in several tokamaks with carbon (C) PFCs, including the National Spherical Torus Experiment (NSTX). In NSTX, contamination of the core plasma with Li impurities was very low (<0.1%)more » despite extensive divertor coatings. Low Li levels in NSTX were found to be largely due to neoclassical forces from the high level of C impurities. Studying impurity levels and transport with Li coatings on stainless steel surfaces in LTX is relevant to future devices (including future enhancements to NSTX-Upgrade) with all-metal PFCs. The new measurements in this thesis were enabled by a refurbished Thomson scattering system and improved impurity spectroscopy, primarily using a novel visible spectrometer monitoring several Li, C, and oxygen (O) emission lines. A simple model was used to account for impurities in unmeasured charge states, assuming constant density in the plasma core and constant concentration in the edge. In discharges with solid Li coatings, volume averaged impurity concentrations were low but non-negligible, with~2-4% Li, ~0.6-2% C, ~0.4-0.7% O, and Z_eff<1.2. Transport was assessed using the TRANSP, NCLASS, and MIST codes. Collisions with the main H ions dominated the neoclassical impurity transport, unlike in NSTX, where collisions with C dominated. Furthermore, neoclassical transport coefficients calculated with NCLASS were similar across all impurity species and differed no more than a factor of two, in contrast to NSTX where they differed by an order

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

Numerical optimization of perturbative coils for tokamaks

NASA Astrophysics Data System (ADS)

Lazerson, Samuel; Park, Jong-Kyu; Logan, Nikolas; Boozer, Allen; NSTX-U Research Team

2014-10-01

Numerical optimization of coils which apply three dimensional (3D) perturbative fields to tokamaks is presented. The application of perturbative 3D magnetic fields in tokamaks is now commonplace for control of error fields, resistive wall modes, resonant field drive, and neoclassical toroidal viscosity (NTV) torques. The design of such systems has focused on control of toroidal mode number, with coil shapes based on simple window-pane designs. In this work, a numerical optimization suite based on the STELLOPT 3D equilibrium optimization code is presented. The new code, IPECOPT, replaces the VMEC equilibrium code with the IPEC perturbed equilibrium code, and targets NTV torque by coupling to the PENT code. Fixed boundary optimizations of the 3D fields for the NSTX-U experiment are underway. Initial results suggest NTV torques can be driven by normal field spectrums which are not pitch-resonant with the magnetic field lines. Work has focused on driving core torque with n = 1 and edge torques with n = 3 fields. Optimizations of the coil currents for the planned NSTX-U NCC coils highlight the code's free boundary capability. This manuscript has been authored by Princeton University under Contract Number DE-AC02-09CH11466 with the U.S. Department of Energy.

78 FR 31592 - T-Mobile Usa, Inc., Core Fault Isolation Team, Engineering Division, Bethlehem, Pennsylvania...

Federal Register 2010, 2011, 2012, 2013, 2014

2013-05-24

... DEPARTMENT OF LABOR Employment and Training Administration [TA-W-82,371] T-Mobile Usa, Inc., Core Fault Isolation Team, Engineering Division, Bethlehem, Pennsylvania; Notice of Affirmative Determination...., Core Fault Isolation Team, Engineering Division, Bethlehem, Pennsylvania (subject firm). The...

Experimental studies of toroidal correlations of plasma density fluctuations along the magnetic field lines in the T-10 tokamak and first results of numerical modeling

NASA Astrophysics Data System (ADS)

Buldakov, M. A.; Vershkov, V. A.; Isaev, M. Yu; Shelukhin, D. A.

2017-10-01

The antenna system of reflectometry diagnostics at the T-10 tokamak allows to study long-range toroidal correlations of plasma density fluctuations along the magnetic field lines. The antenna systems are installed in two poloidal cross-sections of the vacuum chamber separated by a 90° angle in the toroidal direction. The experiments, which were conducted at the low field side, showed that the high level of toroidal correlations is observed only for quasi-coherent fluctuations. However, broadband and stochastic low frequency fluctuations are not correlated. Numerical modeling of the plasma turbulence structure in the T-10 tokamak was conducted to interpret the experimental results and take into account non-locality of reflectometry measurements. In the model used, it was assumed that the magnitudes of density fluctuations are constant along the magnetic field lines. The 2D full-wave Tamic-RTH code was used to model the reflectometry signals. High level of correlations for quasi-coherent fluctuations was obtained during the modeling, which agrees with the experimental observations. However, the performed modeling also predicts high level of correlations for broadband fluctuations, which contradicts the experimental data. The modeling showed that the effective reflection radius, from which the information on quasi-coherent plasma turbulence is obtained, is shifted outwards from the reflection radius by approximately 7 mm.

Multi-scale gyrokinetic simulation of Alcator C-Mod tokamak discharges

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

Howard, N. T., E-mail: nthoward@psfc.mit.edu; White, A. E.; Greenwald, M.

2014-03-15

Alcator C-Mod tokamak discharges have been studied with nonlinear gyrokinetic simulation simultaneously spanning both ion and electron spatiotemporal scales. These multi-scale simulations utilized the gyrokinetic model implemented by GYRO code [J. Candy and R. E. Waltz, J. Comput. Phys. 186, 545 (2003)] and the approximation of reduced electron mass (μ = (m{sub D}/m{sub e}){sup .5} = 20.0) to qualitatively study a pair of Alcator C-Mod discharges: a low-power discharge, previously demonstrated (using realistic mass, ion-scale simulation) to display an under-prediction of the electron heat flux and a high-power discharge displaying agreement with both ion and electron heat flux channels [N. T. Howard et al.,more » Nucl. Fusion 53, 123011 (2013)]. These multi-scale simulations demonstrate the importance of electron-scale turbulence in the core of conventional tokamak discharges and suggest it is a viable candidate for explaining the observed under-prediction of electron heat flux. In this paper, we investigate the coupling of turbulence at the ion (k{sub θ}ρ{sub s}∼O(1.0)) and electron (k{sub θ}ρ{sub e}∼O(1.0)) scales for experimental plasma conditions both exhibiting strong (high-power) and marginally stable (low-power) low-k (k{sub θ}ρ{sub s} < 1.0) turbulence. It is found that reduced mass simulation of the plasma exhibiting marginally stable low-k turbulence fails to provide even qualitative insight into the turbulence present in the realistic plasma conditions. In contrast, multi-scale simulation of the plasma condition exhibiting strong turbulence provides valuable insight into the coupling of the ion and electron scales.« 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

42GHz ECRH assisted Plasma Breakdown in tokamak SST-1

NASA Astrophysics Data System (ADS)

Shukla, B. K.; Pradhan, S.; Patel, Paresh; Babu, Rajan; Patel, Jatin; Patel, Harshida; Dhorajia, Pragnesh; Tanna, V.; Atrey, P. K.; Manchanda, R.; Gupta, Manoj; Joisa, Shankar; Gupta, C. N.; Danial, Raju; Singh, Prashant; Jha, R.; Bora, D.

2015-03-01

In SST-1, 42GHz ECRH system has been commissioned to carry out breakdown and heating experiments at 0.75T and 1.5T operating toroidal magnetic fields. The 42GHz ECRH system consists of high power microwave source Gyrotron capable to deliver 500kW microwave power for 500ms duration, approximately 20 meter long transmission line and a mirror based launcher. The ECRH power in fundamental O-mode & second harmonic X-mode is launched from low field side (radial port) of the tokamak. At 0.75T operation, approximately 300 kW ECH power is launched in second harmonic X-mode and successful ECRH assisted breakdown is achieved at low loop_voltage ~ 3V. The ECRH power is launched around 45ms prior to loop voltage. The hydrogen pressure in tokamak is maintained ~ 1×10-5mbar and the pre-ionized density is ~ 4×1012/cc. At 1.5T operating toroidal magnetic field, the ECH power is launched in fundamental O-mode. The ECH power at fundamental harmonic is varied from 100 kW to 250 kW and successful breakdown is achieved in all ECRH shots. In fundamental harmonic there is no delay in breakdown while at second harmonic ~ 40ms delay is observed, which is normal in case of second harmonic ECRH assisted breakdown.

Doppler-broadening measurements of x-ray lines for determination of the ion temperature in tokamak plasmas

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

Bitter, M.; von Goeler, S.; Horton, R.

1979-01-29

Ion-temperature results are deduced from Doppler-broadening measurements of the K..cap alpha.. (1s-2p) resonance line emitted from heliumlike iron impurity ions in the hot central core of PLT (Princeton Large Torus) tokamak discharges. The measurements were performed using a high-resolution Bragg-crystal spectrometer with a multiwire proportional counter.

Noninductively Driven Tokamak Plasmas at Near-Unity Toroidal Beta

DOE PAGES

Schlossberg, David J.; Bodner, Grant M.; Bongard, Michael W.; ...

2017-07-01

Access to and characterization of sustained, toroidally confined plasmas with a very high plasma-to-magnetic pressure ratio (β t), low internal inductance, high elongation, and nonsolenoidal current drive is a central goal of present tokamak plasma research. Stable access to this desirable parameter space is demonstrated in plasmas with ultralow aspect ratio and high elongation. Local helicity injection provides nonsolenoidal sustainment, low internal inductance, and ion heating. Equilibrium analyses indicate β t up to ~100% with a minimum |B| well spanning up to ~50% of the plasma volume.

Noninductively Driven Tokamak Plasmas at Near-Unity Toroidal Beta.

PubMed

Schlossberg, D J; Bodner, G M; Bongard, M W; Burke, M G; Fonck, R J; Perry, J M; Reusch, J A

2017-07-21

Access to and characterization of sustained, toroidally confined plasmas with a very high plasma-to-magnetic pressure ratio (β_{t}), low internal inductance, high elongation, and nonsolenoidal current drive is a central goal of present tokamak plasma research. Stable access to this desirable parameter space is demonstrated in plasmas with ultralow aspect ratio and high elongation. Local helicity injection provides nonsolenoidal sustainment, low internal inductance, and ion heating. Equilibrium analyses indicate β_{t} up to ∼100% with a minimum |B| well spanning up to ∼50% of the plasma volume.

U.S. EPA, Pesticide Product Label, TNT GARBAGE CAN SPRAY AND DEODORIZER, 05/14/1970

EPA Pesticide Factsheets

2011-04-21

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Current drive by spheromak injection into a tokamak

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

Brown, M.R.; Bellan, P.M.

1990-04-30

We report the first observation of current drive by injection of a spheromak plasma into a tokamak (Caltech ENCORE small reasearch tokamak) due to the process of helicity injection. After an abrupt 30% increase, the tokamak current decays by a factor of 3 due to plasma cooling caused by the merging of the relatively cold spheromak with the tokamak. The tokamak density profile peaks sharply due to the injected spheromak plasma ({ital {bar n}}{sub 3} increases by a factor of 6) then becomes hollow, suggestive of an interchange instability.

Continuous, edge localized ion heating during non-solenoidal plasma startup and sustainment in a low aspect ratio tokamak

NASA Astrophysics Data System (ADS)

Burke, M. G.; Barr, J. L.; Bongard, M. W.; Fonck, R. J.; Hinson, E. T.; Perry, J. M.; Reusch, J. A.; Schlossberg, D. J.

2017-07-01

Plasmas in the Pegasus spherical tokamak are initiated and grown by the non-solenoidal local helicity injection (LHI) current drive technique. The LHI system consists of three adjacent electron current sources that inject multiple helical current filaments that can reconnect with each other. Anomalously high impurity ion temperatures are observed during LHI with T i,OV  ⩽  650 eV, which is in contrast to T i,OV  ⩽  70 eV from Ohmic heating alone. Spatial profiles of T i,OV indicate an edge localized heating source, with T i,OV ~ 650 eV near the outboard major radius of the injectors and dropping to ~150 eV near the plasma magnetic axis. Experiments without a background tokamak plasma indicate the ion heating results from magnetic reconnection between adjacent injected current filaments. In these experiments, the HeII T i perpendicular to the magnetic field is found to scale with the reconnecting field strength, local density, and guide field, while {{T}\\text{i,\\parallel}} experiences little change, in agreement with two-fluid reconnection theory. This ion heating is not expected to significantly impact the LHI plasma performance in Pegasus, as it does not contribute significantly to the electron heating. However, estimates of the power transfer to the bulk ion are quite large, and thus LHI current drive provides an auxiliary ion heating mechanism to the tokamak plasma.

Continuous, edge localized ion heating during non-solenoidal plasma startup and sustainment in a low aspect ratio tokamak

DOE PAGES

Burke, Marcus G.; Barr, Jayson L.; Bongard, Michael W.; ...

2017-05-16

Plasmas in the Pegasus spherical tokamak are initiated and grown by the non-solenoidal local helicity injection (LHI) current drive technique. The LHI system consists of three adjacent electron current sources that inject multiple helical current filaments that can reconnect with each other. Anomalously high impurity ion temperatures are observed during LHI with T i,OV ≤ 650 eV, which is in contrast to T i,OV ≤ 70 eV from Ohmic heating alone. Spatial profiles of T i,OV indicate an edge localized heating source, with T i,OV ~ 650 eV near the outboard major radius of the injectors and dropping to ~150 eV near the plasma magnetic axis. Experiments without a background tokamak plasma indicate the ion heating results from magnetic reconnection between adjacent injected current filaments. In these experiments, the HeII T i perpendicular to the magnetic field is found to scale with the reconnecting field strength, local density, and guide field, whilemore » $${{T}_{\\text{i},\\parallel}}$$ experiences little change, in agreement with two-fluid reconnection theory. In conclusion, this ion heating is not expected to significantly impact the LHI plasma performance in Pegasus, as it does not contribute significantly to the electron heating. However, estimates of the power transfer to the bulk ion are quite large, and thus LHI current drive provides an auxiliary ion heating mechanism to the tokamak plasma.« less

Cytotoxic T lymphocytes and CD4 epitope mutations in the pre-core/core region of hepatitis B virus in chronic hepatitis B carriers in Northeast Iran.

PubMed

Zhand, Sareh; Tabarraei, Alijan; Nazari, Amineh; Moradi, Abdolvahab

2017-07-01

Hepatitis B virus (HBV) is vulnerable to many various mutations. Those within epitopes recognized by sensitized T cells may influence the re-emergence of the virus. This study was designed to investigate the mutation in immune epitope regions of HBV pre-core/core among chronic HBV patients of Golestan province, Northeast Iran. In 120 chronic HBV carriers, HBV DNA was extracted from blood plasma samples and PCR was done using specific primers. Direct sequencing and alignment of the pre-core/core region were applied using reference sequence from Gene Bank database (Accession Number AB033559). The study showed 27 inferred amino acid substitutions, 9 of which (33.3%) were in CD4 and 2 (7.4%) in cytotoxic T lymphocytes' (CTL) epitopes and 16 other mutations (59.2%) were observed in other regions. CTL escape mutations were not commonly observed in pre-core/core sequences of chronic HBV carriers in the locale of study. It can be concluded that most of the inferred amino acid substitutions occur in different immune epitopes other than CTL and CD4.

The Aneutronic Rodless Ultra Low Aspect Ratio Tokamak

NASA Astrophysics Data System (ADS)

Ribeiro, Celso

2016-10-01

The replacement of the metal centre-post in spherical tokamaks (STs) by a plasma centre-post (PCP, the TF current carrier) is the ideal scenario for a ST reactor. A simple rodless ultra low aspect-ratio tokamak (RULART) using a screw-pinch PCP ECR-assisted with an external solenoid has been proposed in the most compact RULART [Ribeiro C, SOFE-15]. There the solenoid provided the stabilizing field for the PCP and the toroidal electrical field for the tokamak start-up, which will stabilize further the PCP, acting as stabilizing closed conducting surface. Relative low TF will be required. The compactness (high ratio of plasma-spherical vessel volume) may provide passive stabilization and easier access to L-H mode transition. It is presented here: 1) stability analysis of the PCP (initially MHD stable due to the hollow J profile); 2) tokamak equilibrium simulations, and 3) potential use for aneutronic reactions studies via pairs of proton p and boron 11B ion beams in He plasmas. The beams' line-of-sights sufficiently miss the sources of each other, thus allowing a near maximum relative velocities and reactivity. The reactions should occur close to the PCP mid-plane. Some born alphas should cross the PCP and be dragged by the ion flow (higher momentum exchange) towards the anode but escape directly to a direct electricity converter. Others will reach evenly the vessel directly or via thermal diffusion (favourable heating by the large excursion 2a), leading to the lowest power wall load possible. This might be a potential hybrid direct-steam cycle conversion reactor scheme, nearly aneutronic, and with no ash or particle retention problems, as opposed to the D-T thermal reaction proposals.

CAMAC throughput of a new RISC-based data acquisition computer at the DIII-D tokamak

NASA Astrophysics Data System (ADS)

Vanderlaan, J. F.; Cummings, J. W.

1993-10-01

The amount of experimental data acquired per plasma discharge at DIII-D has continued to grow. The largest shot size in May 1991 was 49 Mbyte; in May 1992, 66 Mbyte; and in April 1993, 80 Mbyte. The increasing load has prompted the installation of a new Motorola 88100-based MODCOMP computer to supplement the existing core of three older MODCOMP data acquisition CPU's. New Kinetic Systems CAMAC serial highway driver hardware runs on the 88100 VME bus. The new operating system is MODCOMP REAL/IX version of AT&T System V UNIX with real-time extensions and networking capabilities; future plans call for installation of additional computers of this type for tokamak and neutral beam control functions. Experiences with the CAMAC hardware and software will be chronicled, including observation of data throughput. The Enhanced Serial Highway crate controller is advertised as twice as fast as the previous crate controller, and computer I/O speeds are expected to also increase data rates.

T1 relaxivity of core-encapsulated gadolinium liposomal contrast agents--effect of liposome size and internal gadolinium concentration.

PubMed

Ghaghada, Ketan; Hawley, Catherine; Kawaji, Keigo; Annapragada, Ananth; Mukundan, Srinivasan

2008-10-01

Long circulating core-encapsulated gadolinium (CE-Gd) liposomal nanoparticles that have surface conjugated polyethylene glycol are a promising platform technology for use as blood pool T1-based magnetic resonance (MR) contrast agents. The objective of this study was to investigate the effect of liposome size and internal (core) Gd concentration on the T1 relaxivity of CE-Gd liposomes. Twelve different liposomal formulations were synthesized and characterized, resulting in a size (50, 100, 200, and 400 nm) and core Gd-concentration (200, 350, and 500 mM) "matrix" of test samples. Subsequently, CE-Gd liposomes were diluted in deionized water (four diluted samples) and molar T1 relaxivity (r1) measurements were performed at 2- and 7-T MR field strengths. The r1 of CE-Gd liposomes was inversely related to the liposome size. The largest change in r1 was observed between liposomes that were extruded through 50- and 100-nm filter membranes. At both field strengths, the variation in internal gadolinium concentration did not show any significant correlation (alpha < or = 0.05) with r1. The size of CE-Gd liposomal nanoparticles significantly affects the T1 relaxivity. An inverse relation was observed between liposome size and T1 relaxivity. The T1 relaxivity did not change significantly with core Gd concentration over the measured concentration range.

Natural Divertor Spherical Tokamak Plasmas with bean shape and ergodic limiter

NASA Astrophysics Data System (ADS)

Ribeiro, Celso; Herrera, Julio; Chavez, Esteban; Tritz, Kevin

2013-10-01

The former spherical tokamak (ST) MEDUSA (Madison EDUcation Small Aspect.ratio tokamak, R < 0.14 m, a < 0.10 m, BT < 0.5T, Ip < 40 kA, 3 ms pulse) is being recommissioned in Costa Rica Institute of Technology. The main objectives of the MEDUSA-CR project are training and to clarify several issues in relevant physics for conventional and mainly STs, including beta studies in bean-shaped ST plasmas, transport, heating and current drive via Alfvén wave, and natural divertor STs with ergodic magnetic limiter. We report here improvements in the self-consistency of these equilibrium comparisons and a preliminary study of their MHD stability beta limits. VIE-ITCR, IAEA-CRP contract 17592, National Instruments of Costa Rica.

Evidence of coupling to Global Alfv{acute e}ne Eigenmodes during Alfv{acute e}n wave current drive experiments on the Phaedrus-T tokamak

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

Vukovic, M.; Wukitch, S.; Harper, M.

1996-02-01

A series of experiments designed to explore mechanisms of power deposition during Alfv{acute e}n wave current drive experiments on the Phaedrus-T tokamak has shown evidence of power deposition via mode conversion of Global Alfv{acute e}n Eigenmodes at the Alfv{acute e}n resonance. Observation of radially localized RF induced density fluctuations in the plasma and their location vs. {ital B}{sub {ital T}} is in agreement with the predictions of behaviour of GAE damping on the AR by the toroidal code LION. Furthermore, the change in the time evolution of the loop voltage, is consistent with the change of effective power deposition radius,more » {ital r}{sub PD}, and is in agreement with the density fluctuations radius. {copyright} {ital 1996 American Institute of Physics.}« less

Core Fucosylation on T Cells, Required for Activation of T-Cell Receptor Signaling and Induction of Colitis in Mice, Is Increased in Patients With Inflammatory Bowel Disease.

PubMed

Fujii, Hironobu; Shinzaki, Shinichiro; Iijima, Hideki; Wakamatsu, Kana; Iwamoto, Chizuru; Sobajima, Tomoaki; Kuwahara, Ryusuke; Hiyama, Satoshi; Hayashi, Yoshito; Takamatsu, Shinji; Uozumi, Naofumi; Kamada, Yoshihiro; Tsujii, Masahiko; Taniguchi, Naoyuki; Takehara, Tetsuo; Miyoshi, Eiji

2016-06-01

Attachment of a fucose molecule to the innermost N-glycan in a glycoprotein (core fucosylation) regulates the activity of many growth factor receptors and adhesion molecules. The process is catalyzed by α1-6 fucosyltransferase (FUT8) and required for immune regulation, but it is not clear whether this process is dysregulated during disease pathogenesis. We investigated whether core fucosylation regulates T-cell activation and induction of colitis in mice, and is altered in patients with inflammatory bowel disease (IBD). Biopsy samples were collected from inflamed and noninflamed regions of intestine from patients (8 with Crohn's disease, 4 with ulcerative colitis, and 4 without IBD [controls]) at Osaka University Hospital. Colitis was induced in FUT8-deficient (Fut8(-/-)) mice and Fut8(+/+) littermates by administration of trinitrobenzene sulfonic acid. Intestinal tissues were collected and analyzed histologically. Immune cells were collected and analyzed by lectin flow cytometry, immunofluorescence, and reverse-transcription polymerase chain reaction, as well as for production of cytokines and levels of T-cell receptor (TCR) in lipid raft fractions. T-cell function was analyzed by intraperitoneal injection of CD4(+)CD62L(+) naïve T cells into RAG2-deficient mice. Levels of core fucosylation were increased on T cells from mice with colitis, compared with mice without colitis, as well as on inflamed mucosa from patients with IBD, compared with their noninflamed tissues or tissues from control patients. Fut8(-/-) mice developed less-severe colitis than Fut8(+/+) mice, and T cells from Fut8(-/-) mice produced lower levels of T-helper 1 and 2 cytokines. Adoptive transfer of Fut8(-/-) T cells to RAG2-deficient mice reduced the severity of colitis. Compared with CD4(+) T cells from Fut8(+/+) mice, those from Fut8(-/-) mice expressed similar levels of TCR and CD28, but these proteins did not contain core fucosylation. TCR complexes formed on CD4(+) T cells from Fut8

Plasma Confinement in the UCLA Electric Tokamak.

NASA Astrophysics Data System (ADS)

Taylor, Robert J.

2001-10-01

The main goal of the newly constructed large Electric Tokamak (R = 5 m, a = 1 m, BT < 0.25 T) is to access an omnigeneous, unity beta(S.C. Cowley, P.K. Kaw, R.S. Kelly, R.M. Kulsrud, Phys. fluids B 3 (1991) 2066.) plasma regime. The design goal was to achieve good confinement at low magnetic fields, consistent with the high beta goal. To keep the program cost down, we adopted the use of ICRF as the primary heating source. Consequently, antenna surfaces covering 1/2 of the surface of the tokamak has been prepared for heating and current drive. Very clean hydrogenic plasmas have been achieved with loop voltage below 0.7 volt and densities 3 times above the Murakami limit, n(0) > 8 x 10^12 cm-3 when there is no MHD activity. The electron temperature, derived from the plasma conductivity is > 250 eV with a central electron energy confinement time > 350 msec in ohmic conditions. The sawteeth period is 50 msec. Edge plasma rotation is induced by plasma biasing via electron injection in an analogous manner to that seen in CCT(R.J. Taylor, M.L. Brown, B.D. Fried, H. Grote, J.R. Liberati, G.J. Morales, P. Pribyl, D. Darrow, and M. Ono. Phys. Rev Lett. 63 2365 1989.) and the neoclassical bifurcation is close to that described by Shaing et al(K.C. Shaing and E.C. Crume, Phys. Rev. Lett. 63 2369 (1989).). In the ohmic phase the confinement tends to be MHD limited. The ICRF heating eliminates the MHD disturbances. Under second harmonic heating conditions, we observe an internal confinement peaking characterized by doubling of the core density and a corresponding increase in the central electron temperature. Charge exchange data, Doppler data in visible H-alpha light, and EC radiation all indicate that ICRF heating works much better than expected. The major effort is focused on increasing the power input and controlling the resulting equilibrium. This task appears to be easy since our current pulses are approaching the 3 second mark without RF heating or current drive. Our

The circuit of polychromator for Experimental Advanced Superconducting Tokamak edge Thomson scattering diagnostic

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

Zang, Qing; Zhao, Junyu; Chen, Hui

2013-09-15

The detector circuit is the core component of filter polychromator which is used for scattering light analysis in Thomson scattering diagnostic, and is responsible for the precision and stability of a system. High signal-to-noise and stability are primary requirements for the diagnostic. Recently, an upgraded detector circuit for weak light detecting in Experimental Advanced Superconducting Tokamak (EAST) edge Thomson scattering system has been designed, which can be used for the measurement of large electron temperature (T{sub e}) gradient and low electron density (n{sub e}). In this new circuit, a thermoelectric-cooled avalanche photodiode with the aid circuit is involved for increasingmore » stability and enhancing signal-to-noise ratio (SNR), especially the circuit will never be influenced by ambient temperature. These features are expected to improve the accuracy of EAST Thomson diagnostic dramatically. Related mechanical construction of the circuit is redesigned as well for heat-sinking and installation. All parameters are optimized, and SNR is dramatically improved. The number of minimum detectable photons is only 10.« less

Fishbone activity in experimental advanced superconducting tokamak neutral beam injection plasma

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

Xu, Liqing; Zhang, Jizong; Chen, Kaiyun, E-mail: Kychen@ipp.cas.cn, E-mail: lqhu@ipp.cas.cn

2015-12-15

Repetitive fishbones near the trapped ion procession frequency were observed for the first time in the neutral beam injection high confinement plasmas in Experimental Advanced Superconducting Tokamak (EAST) tokamak, and diagnosed using a solid-state neutral particle analyzer based on a compact silicon photodiode together with an upgraded high spatial-temporal-resolution multi-arrays soft X-ray (SX) system. This 1/1 typical internal kink mode propagates in the ion-diamagnetism direction with a rotation speed faster than the bulk plasma in the plasma frame. From the SX measurements, this mode frequency is typical of chirping down and the energetic particle effect related to the twisting modemore » structure. This ion fishbone was found able to trigger a multiple core sawtooth crashes with edge-2/1 sideband modes, as well as to lead to a transition from fishbone to long lived saturated kink mode to fishbone. Furthermore, using SX tomography, a correlation between mode amplitude and mode frequency was found. Finally, a phenomenological prey–predator model was found to reproduce the fishbone nonlinear process well.« less

Formation and stability of impurity "snakes" in tokamak plasmas.

PubMed

Delgado-Aparicio, L; Sugiyama, L; Granetz, R; Gates, D A; Rice, J E; Reinke, M L; Bitter, M; Fredrickson, E; Gao, C; Greenwald, M; Hill, K; Hubbard, A; Hughes, J W; Marmar, E; Pablant, N; Podpaly, Y; Scott, S; Wilson, R; Wolfe, S; Wukitch, S

2013-02-08

New observations of the formation and dynamics of long-lived impurity-induced helical "snake" modes in tokamak plasmas have recently been carried out on Alcator C-Mod. The snakes form as an asymmetry in the impurity ion density that undergoes a seamless transition from a small helically displaced density to a large crescent-shaped helical structure inside q<1, with a regularly sawtoothing core. The observations show that the conditions for the formation and persistence of a snake cannot be explained by plasma pressure alone. Instead, many features arise naturally from nonlinear interactions in a 3D MHD model that separately evolves the plasma density and temperature.

SOCS1 and SOCS3 Are Targeted by Hepatitis C Virus Core/gC1qR Ligation To Inhibit T-Cell Function

PubMed Central

Yao, Zhi Qiang; Waggoner, Stephen N.; Cruise, Michael W.; Hall, Caroline; Xie, Xuefang; Oldach, David W.; Hahn, Young S.

2005-01-01

T cells play an important role in the control of hepatitis C virus (HCV) infection. We have previously demonstrated that the HCV core inhibits T-cell responses through interaction with gC1qR. We show here that core proteins from chronic and resolved HCV patients differ in sequence, gC1qR-binding ability, and T-cell inhibition. Specifically, chronic core isolates bind to gC1qR more efficiently and inhibit T-cell proliferation as well as gamma interferon (IFN-γ) production more profoundly than resolved core isolates. This inhibition is mediated by the disruption of STAT phosphorylation through the induction of SOCS molecules. Silencing either SOCS1 or SOCS3 by small interfering RNA dramatically augments the production of IFN-γ in T cells, thereby abrogating the inhibitory effect of core. Additionally, the ability of core proteins from patients with chronic infections to induce SOCS proteins and suppress STAT activation greatly exceeds that of core proteins from patients with resolved infections. These results suggest that the HCV core/gC1qR-induced T-cell dysfunction involves the induction of SOCS, a powerful inhibitor of cytokine signaling, which represents a novel mechanism by which a virus usurps the host machinery for persistence. PMID:16306613

Effects of Hot Limiter Biasing on Tokamak Runaway Discharges

NASA Astrophysics Data System (ADS)

Salar Elahi, A.; Ghoranneviss, M.; Ghanbari, M. R.

2013-10-01

In this research hot limiter biasing effects on the Runaway discharges were investigated. First wall of the tokamak reactors can affects serious damage due to the high energy runaway electrons during a major disruption and therefore its life time can be reduced. Therefore, it is important to find methods to decrease runaway electron generation and their energy. Tokamak limiter biasing is one of the methods for controlling the radial electric field and can induce a transition to an improved confinement state. In this article generation of runaway electrons and the energy they can obtain will be investigated theoretically. Moreover, in order to apply radial biasing an emissive limiter biasing is utilized. The biased limiter can apply +380 V in the status of cold and hot to the plasma and result in the increase of negative bias current in hot status. In fact, in this experiment we try to decrease the generation of runaway electrons and their energy by using emissive limiter biasing inserted on the IR-T1 tokamak. The mean energy of these electrons was obtained by spectroscopy of hard X-ray. Also, the plasma current center shift was measured from the vertical field coil characteristics in presence of limiter biasing. The calculation is made focusing on the vertical field coil current and voltage changes due to a horizontal displacement of plasma column.

Three-dimensional magnetohydrodynamic equilibrium of quiescent H-modes in tokamak systems

NASA Astrophysics Data System (ADS)

Cooper, W. A.; Graves, J. P.; Duval, B. P.; Sauter, O.; Faustin, J. M.; Kleiner, A.; Lanthaler, S.; Patten, H.; Raghunathan, M.; Tran, T.-M.; Chapman, I. T.; Ham, C. J.

2016-06-01

Three dimensional free boundary magnetohydrodynamic equilibria that recover saturated ideal kink/peeling structures are obtained numerically. Simulations that model the JET tokamak at fixed < β > =1.7% with a large edge bootstrap current that flattens the q-profile near the plasma boundary demonstrate that a radial parallel current density ribbon with a dominant m /n  =  5/1 Fourier component at {{I}\\text{t}}=2.2 MA develops into a broadband spectrum when the toroidal current I t is increased to 2.5 MA.

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.

Public Data Set: Impedance of an Intense Plasma-Cathode Electron Source for Tokamak Plasma Startup

DOE Data Explorer

Hinson, Edward T. [University of Wisconsin-Madison] (ORCID:000000019713140X); Barr, Jayson L. [University of Wisconsin-Madison] (ORCID:0000000177685931); Bongard, Michael W. [University of Wisconsin-Madison] (ORCID:0000000231609746); Burke, Marcus G. [University of Wisconsin-Madison] (ORCID:0000000176193724); Fonck, Raymond J. [University of Wisconsin-Madison] (ORCID:0000000294386762); Perry, Justin M. [University of Wisconsin-Madison] (ORCID:0000000171228609)

2016-05-31

This data set contains openly-documented, machine readable digital research data corresponding to figures published in E.T. Hinson et al., 'Impedance of an Intense Plasma-Cathode Electron Source for Tokamak Plasma Startup,' Physics of Plasmas 23, 052515 (2016).

A Research Program of Spherical Tokamak in China

NASA Astrophysics Data System (ADS)

He, Ye-xi

2002-08-01

The mission of this program is to explore the spherical torus plasma with a SUNIST spherical tokamak. Main experiments in the start phase will be involved with breakdown and plasma current set-up with a mode of saving volt-second and without ohmic heating system, equilibrium and instability, current driving, heating and profile modification. The SUNIST is a university-scale conceptual spherical tokamak, with R = 0.3 m, A 1.3, Ip ~ 50 kA, BT < 0.15 T, and PRF = 100 kW. The only peculiarity of SUNIST is that there is a toroidal insulating break along the outer wall of vacuum vessel. The expected that advantages of this arrangement are helpful not only for saving flux swing, but also for having a deep understanding of what will influence the discharge startup and globe performances of plasma under different conditions of strong vessel eddy and ECR power assistance. Of course, the vessel structure of cross seal will be at a great risk of controlling vacuum quality, although we have achieved positive results on simulation test and vacuum vessel test.

The Research Progress of the J-TEXT Tokamak

NASA Astrophysics Data System (ADS)

Zhuang, Ge; Wang, Zhijiang; Ding, Yonghua; Zhang, Ming; Yang, Zhoujun; Gao, Li; Zhang, Xiaoqing; Hu, Xiwei; Pan, Yuan

2010-11-01

In 2004, the TEXT-U tokamak was disassembled and shipped to China, and later on settle down in Huazhong University of Science and Technology. The machine was renamed as the Joint TEXT (J-TEXT) tokamak and obtained its first plasma in 2007. The typical J-TEXT Ohmic discharge was performed in the limiter configuration with the main parameters as follows: major radius R=1.05 m, minor radius a=0.27m, toroidal magnetic field BT=2.2T, plasma current Ip>200kA, line-averaged density ne˜ 2-3 . 1019/m^3, and electron temperature Te0˜ 700eV. Up till now, a few diagnostic systems used to facilitate routine operation and experimental studies were designed and developed. Benefiting from these diagnostic tools, the observation of MHD activities and the statistical analysis of disruption events were done. And measurements of the electrostatic fluctuations in the edge region and conditional analysis of the intermittent burst events near the LCFS were also made as well. The preliminary results will be presented in detail in the meeting.

The Spherical Tokamak MEDUSA for Costa Rica

NASA Astrophysics Data System (ADS)

Ribeiro, Celso; Vargas, Ivan; Guadamuz, Saul; Mora, Jaime; Ansejo, Jose; Zamora, Esteban; Herrera, Julio; Chaves, Esteban; Romero, Carlos

2012-10-01

The former spherical tokamak (ST) MEDUSA (Madison EDUcation Small Aspect.ratio tokamak, R<0.14m, a<0.10m, BT<0.5T, Ip<40kA, 3ms pulse)[1] is in a process of donation to Costa Rica Institute of Technology. The main objective of MEDUSA is to train students in plasma physics /technical related issues which will help all tasks of the very low aspect ratio stellarator SCR-1(A≡R/>=3.6, under design[2]) and also the ongoing activities in low temperature plasmas. Courses in plasma physics at undergraduate and post-graduate joint programme levels are regularly conducted. The scientific programme is intend to clarify several issues in relevant physics for conventional and mainly STs, including transport, heating and current drive via Alfv'en wave, and natural divertor STs with ergodic magnetic limiter[3,4]. [1] G.D.Garstka, PhD thesis, University of Wisconsin at Madison, 1997 [2] L.Barillas et al., Proc. 19^th Int. Conf. Nucl. Eng., Japan, 2011 [3] C.Ribeiro et al., IEEJ Trans. Electrical and Electronic Eng., 2012(accepted) [4] C.Ribeiro et al., Proc. 39^th EPS Conf. Contr. Fusion and Plasma Phys., Sweden, 2012

Control and Data Acquisition for the Spherical Tokamak MEDUSA-CR

NASA Astrophysics Data System (ADS)

Soto, Christian; Gonzalez, Jeferson; Carvajal, Johan; Ribeiro, Celso

2013-10-01

The former spherical tokamak (ST) MEDUSA (Madison EDUcation Small Aspect.ratio tokamak, R < 0.14 m, a < 0.10 m, BT < 0.5 T, Ip < 40 kA, 3 ms pulse) is being recommissioned in Costa Rica Institute of Technology. The main objectives of the MEDUSA-CR project are training and to clarify several issues in relevant physics for conventional and mainly STs, including beta studies in bean-shaped ST plasmas, transport, heating and current drive via Alfvén wave, and natural divertor STs with ergodic magnetic limiter. We present here the control and data acquisition systems for MEDUSA-CR device which are based on National Instruments (NI) software (LabView) and hardware on loan to our laboratory via NI-Costa Rica. The interface with the energy, gas fueling, and security systems are also presented. VIE-ITCR, IAEA-CRP contract 17592, National Instruments of Costa Rica.

Energy balance in TM-1-MH Tokamak (ohmical heating)

NASA Astrophysics Data System (ADS)

Stoeckel, J.; Koerbel, S.; Kryska, L.; Kopecky, V.; Dadalec, V.; Datlov, J.; Jakubka, K.; Magula, P.; Zacek, F.; Pereverzev, G. V.

1981-10-01

Plasma in the TM-1-MH Tokamak was experimentally studied in the parameter range: tor. mg. field B = 1,3 T, plasma current I sub p = 14 kA, electron density N sub E 3.10 to the 19th power cubic meters. The two numerical codes are available for the comparison with experimental data. TOKATA-code solves simplified energy balance equations for electron and ion components. TOKSAS-code solves the detailed energy balance of the ion component.

Bootstrap current in a tokamak

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

Kessel, C.E.

1994-03-01

The bootstrap current in a tokamak is examined by implementing the Hirshman-Sigmar model and comparing the predicted current profiles with those from two popular approximations. The dependences of the bootstrap current profile on the plasma properties are illustrated. The implications for steady state tokamaks are presented through two constraints; the pressure profile must be peaked and {beta}{sub p} must be kept below a critical value.

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.

Simulation of magnetic island dynamics under resonant magnetic perturbation with the TEAR code and validation of the results on T-10 tokamak data

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

Ivanov, N. V.; Kakurin, A. M.

2014-10-15

Simulation of the magnetic island evolution under Resonant Magnetic Perturbation (RMP) in rotating T-10 tokamak plasma is presented with intent of TEAR code experimental validation. In the T-10 experiment chosen for simulation, the RMP consists of a stationary error field, a magnetic field of the eddy current in the resistive vacuum vessel and magnetic field of the externally applied controlled halo current in the plasma scrape-off layer (SOL). The halo-current loop consists of a rail limiter, plasma SOL, vacuum vessel, and external part of the circuit. Effects of plasma resistivity, viscosity, and RMP are taken into account in the TEARmore » code based on the two-fluid MHD approximation. Radial distribution of the magnetic flux perturbation is calculated with account of the externally applied RMP. A good agreement is obtained between the simulation results and experimental data for the cases of preprogrammed and feedback-controlled halo current in the plasma SOL.« less

Measurements of Dielectronic Recombination Rate Coefficients for He-like and H-like Metal Ions from Tokamak Plasmas

NASA Astrophysics Data System (ADS)

Bitter, M.; Hsuan, H.; von Goeler, S.; Hill, K. W.; Grek, B.; Johnson, D. W.; Decaux, V.

1991-01-01

The dielectronic recombination rate coefficients associated with the Δn = 1 radiative core transitions have been determined for Ti XXI, Fe XXV, Ni XXVII, and Ti XXII as a function of the electron temperature from the analysis of x-ray satellite spectra emitted from tokamak plasmas. The results obtained are in good agreement with the recent theoretical predictions of Bely-Dubau et al., and Karim and Bhalla.

The reconstruction and research progress of the TEXT-U tokamak in China

NASA Astrophysics Data System (ADS)

Zhuang, G.; Pan, Y.; Hu, X. W.; Wang, Z. J.; Ding, Y. H.; Zhang, M.; Gao, L.; Zhang, X. Q.; Yang, Z. J.; Yu, K. X.; Gentle, K. W.; Huang, H.; J-TEXT Team

2011-09-01

The TEXT/(TEXT-U) tokamak, formerly built and operated by the University of Texas at Austin in USA, was dismantled and shipped to China in 2004, and renamed as the Joint TEXT (J-TEXT) tokamak. The reconstruction work, which included reassembly of the machine and development of peripheral devices, was completed in the spring of 2007. Consequently, the first plasma was obtained at the end of 2007. At present, a typical J-TEXT ohmic discharge can produce a plasma with flattop current up to 220 kA and lasting for 300 ms, line-averaged density above 2 × 1019 m-3, and an electron temperature of about 800 eV, with a toroidal magnetic field of 2.2 T. A number of diagnostic devices used to facilitate the routine operation and experimental scenarios were developed on the J-TEXT tokamak. Hence, the measurements of the electrostatic fluctuations in the edge region and conditional analysis of the intermittent burst events near the last closed flux surface were undertaken. The observation and simple analysis of MHD activity and disruption events were also performed. The preliminary experimental results and the future research plan for the J-TEXT are described in detail.

User's manual for COAST 4: a code for costing and sizing tokamaks

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

Sink, D. A.; Iwinski, E. M.

1979-09-01

The purpose of this report is to document the computer program COAST 4 for the user/analyst. COAST, COst And Size Tokamak reactors, provides complete and self-consistent size models for the engineering features of D-T burning tokamak reactors and associated facilities involving a continuum of performance including highly beam driven through ignited plasma devices. TNS (The Next Step) devices with no tritium breeding or electrical power production are handled as well as power producing and fissile producing fusion-fission hybrid reactors. The code has been normalized with a TFTR calculation which is consistent with cost, size, and performance data published in themore » conceptual design report for that device. Information on code development, computer implementation and detailed user instructions are included in the text.« less

Numerical studies of edge localized instabilities in tokamaks

NASA Astrophysics Data System (ADS)

Wilson, H. R.; Snyder, P. B.; Huysmans, G. T. A.; Miller, R. L.

2002-04-01

A new computational tool, edge localized instabilities in tokamaks equilibria (ELITE), has been developed to help our understanding of short wavelength instabilities close to the edge of tokamak plasmas. Such instabilities may be responsible for the edge localized modes observed in high confinement H-mode regimes, which are a serious concern for next step tokamaks because of the high transient power loads which they can impose on divertor target plates. ELITE uses physical insight gained from analytic studies of peeling and ballooning modes to provide an efficient way of calculating the edge ideal magnetohydrodynamic stability properties of tokamaks. This paper describes the theoretical formalism which forms the basis for the code.

Core Outcomes in Ventilation Trials (COVenT): protocol for a core outcome set using a Delphi survey with a nested randomised trial and observational cohort study.

PubMed

Blackwood, Bronagh; Ringrow, Suzanne; Clarke, Mike; Marshall, John; Rose, Louise; Williamson, Paula; McAuley, Danny

2015-08-20

Among clinical trials of interventions that aim to modify time spent on mechanical ventilation for critically ill patients there is considerable inconsistency in chosen outcomes and how they are measured. The Core Outcomes in Ventilation Trials (COVenT) study aims to develop a set of core outcomes for use in future ventilation trials in mechanically ventilated adults and children. We will use a mixed methods approach that incorporates a randomised trial nested within a Delphi study and a consensus meeting. Additionally, we will conduct an observational cohort study to evaluate uptake of the core outcome set in published studies at 5 and 10 years following core outcome set publication. The three-round online Delphi study will use a list of outcomes that have been reported previously in a review of ventilation trials. The Delphi panel will include a range of stakeholder groups including patient support groups. The panel will be randomised to one of three feedback methods to assess the impact of the feedback mechanism on subsequent ranking of outcomes. A final consensus meeting will be held with stakeholder representatives to review outcomes. The COVenT study aims to develop a core outcome set for ventilation trials in critical care, explore the best Delphi feedback mechanism for achieving consensus and determine if participation increases use of the core outcome set in the long term.

Use of soft x-ray diagnostic on the COMPASS tokamak for investigations of sawteeth crash neighborhood and of plasma position using fast inversion methods

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

Imrisek, M.; Faculty of Mathematics and Physics, Charles University in Prague, Prague; Weinzettl, V.

2014-11-15

The soft x-ray diagnostic is suitable for monitoring plasma activity in the tokamak core, e.g., sawtooth instability. Moreover, spatially resolved measurements can provide information about plasma position and shape, which can supplement magnetic measurements. In this contribution, fast algorithms with the potential for a real-time use are tested on the data from the COMPASS tokamak. In addition, the soft x-ray data are compared with data from other diagnostics in order to discuss possible connection between sawtooth instability on one side and the transition to higher confinement mode, edge localized modes and productions of runaway electrons on the other side.

An enhanced tokamak startup model

NASA Astrophysics Data System (ADS)

Goswami, Rajiv; Artaud, Jean-François

2017-01-01

The startup of tokamaks has been examined in the past in varying degree of detail. This phase typically involves the burnthrough of impurities and the subsequent rampup of plasma current. A zero-dimensional (0D) model is most widely used where the time evolution of volume averaged quantities determines the detailed balance between the input and loss of particle and power. But, being a 0D setup, these studies do not take into consideration the co-evolution of plasma size and shape, and instead assume an unchanging minor and major radius. However, it is known that the plasma position and its minor radius can change appreciably as the plasma evolves in time to fill in the entire available volume. In this paper, an enhanced model for the tokamak startup is introduced, which for the first time takes into account the evolution of plasma geometry during this brief but highly dynamic period by including realistic one-dimensional (1D) effects within the broad 0D framework. In addition the effect of runaway electrons (REs) has also been incorporated. The paper demonstrates that the inclusion of plasma cross section evolution in conjunction with REs plays an important role in the formation and development of tokamak startup. The model is benchmarked against experimental results from ADITYA tokamak.

Blockade of PD-1/B7-H1 Interaction Restores Effector CD8+ T Cell Responses in a Hepatitis C Virus Core Murine Model1

PubMed Central

Lukens, John R.; Cruise, Michael W.; Lassen, Matthew G.; Hahn, Young S.

2010-01-01

The impaired function of CD8+ T cells is characteristic of hepatitis C virus (HCV) persistent infection. HCV core protein has been reported to inhibit CD8+ T cell responses. To determine the mechanism of the HCV core in suppressing Ag-specific CD8+ T cell responses, we generated a transgenic mouse, core(+) mice, where the expression of core protein is directed to the liver using the albumin promoter. Using a recombinant adenovirus to deliver Ag, we demonstrated that core(+) mice failed to clear adenovirus-LacZ (Ad-LacZ) infection in the liver. The effector function of LacZ-specific CD8+ T cells was particularly impaired in the livers of core(+) mice, with suppression of IFN-γ, TNF-α, and granzyme B production by CD8+ T cells. In addition, the impaired CD8+ T cell responses in core(+) mice were accompanied by the enhanced expression of the inhibitory receptor programmed death-1 (PD-1) by LacZ-specific CD8+ T cells and its ligand B7-H1 on liver dendritic cells following Ad-LacZ infection. Importantly, blockade of the PD-1/B7-H1 inhibitory pathway (using a B7-H1 blocking antibody) in core(+) mice enhanced effector function of CD8+ T cells and cleared Ad-LacZ-infection as compared with that in mice treated with control Ab. This suggests that the regulation of the PD-1/B7-H1 inhibitory pathway is crucial for HCV core-mediated impaired T cell responses and viral persistence in the liver. This also suggests that manipulation of the PD-1/B7-H1 pathway may be a potential immunotherapy to enhance effector T cell responses during persistent HCV infection. PMID:18354211

Radioactivity evaluation for the KSTAR tokamak.

PubMed

Kim, Hyunduk; Lee, Hee-Seock; Hong, Sukmo; Kim, Minho; Chung, Chinwha; Kim, Changsuk

2005-01-01

The deuterium-deuterium (D-D) reaction in the KSTAR (Korea Superconducting Tokamak Advanced Research) tokamak generates neutrons with a peak yield of 2.5 x 10(16) s(-1) through a pulse operation of 300 s. Since the structure material of the tokamak is irradiated with neutrons, this environment will restrict work around and inside the tokamak from a radiation protection physics point of view after shutdown. Identification of neutron-produced radionuclides and evaluation of absorbed dose in the structure material are needed to develop a guiding principle for radiation protection. The activation level was evaluated by MCNP4C2 and an inventory code, FISPACT. The absorbed dose in the working area decreased by 4.26 x 10(-4) mrem h(-1) in the inner vessel 1.5 d after shutdown. Furthermore, tritium strongly contributes to the contamination in the graphite tile. The amount of tritium produced by neutrons was 3.03 x 10(6) Bq kg(-1) in the carbon graphite of a plasma-facing wall.

Bulk ion heating with ICRF waves in tokamaks

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

Mantsinen, M. J., E-mail: mervi.mantsinen@bsc.es; Barcelona Supercomputing Center, Barcelona; Bilato, R.

2015-12-10

Heating with ICRF waves is a well-established method on present-day tokamaks and one of the heating systems foreseen for ITER. However, further work is still needed to test and optimize its performance in fusion devices with metallic high-Z plasma facing components (PFCs) in preparation of ITER and DEMO operation. This is of particular importance for the bulk ion heating capabilities of ICRF waves. Efficient bulk ion heating with the standard ITER ICRF scheme, i.e. the second harmonic heating of tritium with or without {sup 3}He minority, was demonstrated in experiments carried out in deuterium-tritium plasmas on JET and TFTR andmore » is confirmed by ICRF modelling. This paper focuses on recent experiments with {sup 3}He minority heating for bulk ion heating on the ASDEX Upgrade (AUG) tokamak with ITER-relevant all-tungsten PFCs. An increase of 80% in the central ion temperature T{sub i} from 3 to 5.5 keV was achieved when 3 MW of ICRF power tuned to the central {sup 3}He ion cyclotron resonance was added to 4.5 MW of deuterium NBI. The radial gradient of the T{sub i} profile reached locally values up to about 50 keV/m and the normalized logarithmic ion temperature gradients R/LT{sub i} of about 20, which are unusually large for AUG plasmas. The large changes in the T{sub i} profiles were accompanied by significant changes in measured plasma toroidal rotation, plasma impurity profiles and MHD activity, which indicate concomitant changes in plasma properties with the application of ICRF waves. When the {sup 3}He concentration was increased above the optimum range for bulk ion heating, a weaker peaking of the ion temperature profile was observed, in line with theoretical expectations.« less

Controlling fusion yield in tokamaks with spin polarized fuel, and feasibility studies on the DIII-D tokamak

DOE PAGES

Pace, D. C.; Lanctot, M. J.; Jackson, G. L.; ...

2015-09-21

The march towards electricity production through tokamaks requires the construction of new facilities and the inevitable replacement of the previous generation. There are, however, research topics that are better suited to the existing tokamaks, areas of great potential that are not sufficiently mature for implementation in high power machines, and these provide strong support for a balanced policy that includes the redirection of existing programs. Spin polarized fusion, in which the nuclei of tokamak fuel particles are spin-aligned and favorably change both the fusion cross-section and the distribution of initial velocity vectors of charged fusion products, is described here asmore » an example of a technological and physics topic that is ripe for development in a machine such as the DIII-D tokamak. In this study, such research and development experiments may not be efficient at the ITER-scale, while the plasma performance, diagnostic access, and collaborative personnel available within the United States’ magnetic fusion research program, and at the DIII-D facility in particular, provide a unique opportunity to further fusion progress.« less

Contributed Review: Nuclear magnetic resonance core analysis at 0.3 T

NASA Astrophysics Data System (ADS)

Mitchell, Jonathan; Fordham, Edmund J.

2014-11-01

Nuclear magnetic resonance (NMR) provides a powerful toolbox for petrophysical characterization of reservoir core plugs and fluids in the laboratory. Previously, there has been considerable focus on low field magnet technology for well log calibration. Now there is renewed interest in the study of reservoir samples using stronger magnets to complement these standard NMR measurements. Here, the capabilities of an imaging magnet with a field strength of 0.3 T (corresponding to 12.9 MHz for proton) are reviewed in the context of reservoir core analysis. Quantitative estimates of porosity (saturation) and pore size distributions are obtained under favorable conditions (e.g., in carbonates), with the added advantage of multidimensional imaging, detection of lower gyromagnetic ratio nuclei, and short probe recovery times that make the system suitable for shale studies. Intermediate field instruments provide quantitative porosity maps of rock plugs that cannot be obtained using high field medical scanners due to the field-dependent susceptibility contrast in the porous medium. Example data are presented that highlight the potential applications of an intermediate field imaging instrument as a complement to low field instruments in core analysis and for materials science studies in general.

Merging-compression formation of high temperature tokamak plasma

NASA Astrophysics Data System (ADS)

Gryaznevich, M. P.; Sykes, A.

2017-07-01

Merging-compression is a solenoid-free plasma formation method used in spherical tokamaks (STs). Two plasma rings are formed and merged via magnetic reconnection into one plasma ring that then is radially compressed to form the ST configuration. Plasma currents of several hundred kA and plasma temperatures in the keV-range have been produced using this method, however until recently there was no full understanding of the merging-compression formation physics. In this paper we explain in detail, for the first time, all stages of the merging-compression plasma formation. This method will be used to create ST plasmas in the compact (R ~ 0.4-0.6 m) high field, high current (3 T/2 MA) ST40 tokamak. Moderate extrapolation from the available experimental data suggests the possibility of achieving plasma current ~2 MA, and 10 keV range temperatures at densities ~1-5  ×  1020 m-3, bringing ST40 plasmas into a burning plasma (alpha particle heating) relevant conditions directly from the plasma formation. Issues connected with this approach for ST40 and future ST reactors are discussed

Memory T-Cell-Mediated Immune Responses Specific to an Alternative Core Protein in Hepatitis C Virus Infection

PubMed Central

Bain, Christine; Parroche, Peggy; Lavergne, Jean Pierre; Duverger, Blandine; Vieux, Claude; Dubois, Valérie; Komurian-Pradel, Florence; Trépo, Christian; Gebuhrer, Lucette; Paranhos-Baccala, Glaucia; Penin, François; Inchauspé, Geneviève

2004-01-01

In vitro studies have described the synthesis of an alternative reading frame form of the hepatitis C virus (HCV) core protein that was named F protein or ARFP (alternative reading frame protein) and includes a domain coded by the +1 open reading frame of the RNA core coding region. The expression of this protein in HCV-infected patients remains controversial. We have analyzed peripheral blood from 47 chronically or previously HCV-infected patients for the presence of T lymphocytes and antibodies specific to the ARFP. Anti-ARFP antibodies were detected in 41.6% of the patients infected with various HCV genotypes. Using a specific ARFP 99-amino-acid polypeptide as well as four ARFP predicted class I-restricted 9-mer peptides, we show that 20% of the patients display specific lymphocytes capable of producing gamma interferon, interleukin-10, or both cytokines. Patients harboring three different viral genotypes (1a, 1b, and 3) carried T lymphocytes reactive to genotype 1b-derived peptides. In longitudinal analysis of patients receiving therapy, both core and ARFP-specific T-cell- and B-cell-mediated responses were documented. The magnitude and kinetics of the HCV antigen-specific responses differed and were not linked with viremia or therapy outcome. These observations provide strong and new arguments in favor of the synthesis, during natural HCV infection, of an ARFP derived from the core sequence. Moreover, the present data provide the first demonstration of the presence of T-cell-mediated immune responses directed to this novel HCV antigen. PMID:15367612

Two-dimensional T2 distribution mapping in rock core plugs with optimal k-space sampling.

PubMed

Xiao, Dan; Balcom, Bruce J

2012-07-01

Spin-echo single point imaging has been employed for 1D T(2) distribution mapping, but a simple extension to 2D is challenging since the time increase is n fold, where n is the number of pixels in the second dimension. Nevertheless 2D T(2) mapping in fluid saturated rock core plugs is highly desirable because the bedding plane structure in rocks often results in different pore properties within the sample. The acquisition time can be improved by undersampling k-space. The cylindrical shape of rock core plugs yields well defined intensity distributions in k-space that may be efficiently determined by new k-space sampling patterns that are developed in this work. These patterns acquire 22.2% and 11.7% of the k-space data points. Companion density images may be employed, in a keyhole imaging sense, to improve image quality. T(2) weighted images are fit to extract T(2) distributions, pixel by pixel, employing an inverse Laplace transform. Images reconstructed with compressed sensing, with similar acceleration factors, are also presented. The results show that restricted k-space sampling, in this application, provides high quality results. Copyright © 2012 Elsevier Inc. All rights reserved.

Disruptions generated runaways in the FTU high field tokamak

NASA Astrophysics Data System (ADS)

Poli, F. M.; Esposito, B.; Maddaluno, G.; Martin-Solis, J. R.

2001-10-01

Disruptions in FTU are usually accompanied by the generation of a strong pulse of photoneutrons (YN 10^12n/s), resulting from photonuclear reactions induced by the bremsstrahlung radiation emitted when runaway electrons (REs) strike the plasma facing components. Measurements of YN during major disruptions on TS [1] showed variations of three orders of magnitude when the toroidal field Bt increases from 1.8T to 3.9T. Similar results were found on JT-60 [2], where no REs are produced for low Bt (<2.2T) and a large YN was measured for higher fields (up to 4T). The range of Bt available in FTU (4T-8T) allows to extend such analysis so that useful predictions can be obtained for operation in next-step high field tokamaks (IGNITOR, ITER). The dependence of YN on Bt is investigated in several FTU disruptions. YN increases with Bt for B_t=4T-6T, while no variation is found for B_t=6T-8T: the role played by ne and Ip on such trend is discussed. [1]P.Joyer,G.Martin,Contr.Fusion and Plasma Heating,Proc.17^thEPS Conf.Amsterdam(1990) [2]R.Yoshino et al.,Nucl.Fus.39 151 (1999)

Kinetic equilibrium reconstruction for the NBI- and ICRH-heated H-mode plasma on EAST tokamak

NASA Astrophysics Data System (ADS)

Zhen, ZHENG; Nong, XIANG; Jiale, CHEN; Siye, DING; Hongfei, DU; Guoqiang, LI; Yifeng, WANG; Haiqing, LIU; Yingying, LI; Bo, LYU; Qing, ZANG

2018-04-01

The equilibrium reconstruction is important to study the tokamak plasma physical processes. To analyze the contribution of fast ions to the equilibrium, the kinetic equilibria at two time-slices in a typical H-mode discharge with different auxiliary heatings are reconstructed by using magnetic diagnostics, kinetic diagnostics and TRANSP code. It is found that the fast-ion pressure might be up to one-third of the plasma pressure and the contribution is mainly in the core plasma due to the neutral beam injection power is primarily deposited in the core region. The fast-ion current contributes mainly in the core region while contributes little to the pedestal current. A steep pressure gradient in the pedestal is observed which gives rise to a strong edge current. It is proved that the fast ion effects cannot be ignored and should be considered in the future study of EAST.

Experiments with lithium limiter on T-11M tokamak and applications of the lithium capillary-pore system in future fusion reactor devices

NASA Astrophysics Data System (ADS)

Mirnov, S. V.; Azizov, E. A.; Evtikhin, V. A.; Lazarev, V. B.; Lyublinski, I. E.; Vertkov, A. V.; Prokhorov, D. Yu

2006-06-01

The paper is an overview of recent results of Li limiter testing in T-11M tokamak. The lithium limiter is based on the capillary-pore system (CPS) concept. The Li erosion process and deuterium (D2) and helium (He) sorption by Li first wall were investigated. The ability of capillary forces to confine the liquid Li in the CPS limiter during disruption was demonstrated. The idea of combined lithium limiter with thin (0.6 mm) CPS coating as a solution of the heat removal problem was realized. As a result the quasi steady-state tokamak regime with duration up to 0.3 s and clean (Zeff = 1) deuterium plasma has been achieved. The temporal evolution of the lithium surface temperature during discharge was measured by a IR radiometer and then was recalculated to the surface power load. For the estimation of the Li limiter erosion the Li neutral and ions spectral line emission were observed. The increase in lithium erosion as a result of limiter heating was discovered. The radial distribution of plasma column radiation measurements showed up to 90% of the total radiation losses in a relatively thin (5 cm) boundary layer and only 10% in a plasma centre during discharges with high Li influx. Oscillations of Li emission and saw-tooth-like oscillations of the limiter surface temperature have been detected in discharge regimes with highest Li limiter temperature (>600 °C). A version of Li CPS first wall of DEMO reactor and Li CPS limiter experiment in the International Thermonuclear Energy Reactor are suggested.

Study on the Characteristics of Plasma Profiles in Improved Confinement Plasmas in HT-7 Superconducting Tokamak

NASA Astrophysics Data System (ADS)

Zhang, Shouyin; Gao, Xiang; Li, Jiangang; Wan, Baonian; Kuang, Guangli; Mao, Jianshan; Zhang, Xiaodong; Xie, Jikang; Wan, Yuanxi; Team HT-7

2000-10-01

In HT-7 superconducting tokamak of circular limiter configuration (R0=122cm, a=30cm, Bt:1 ~2.2T), plasma profiles were modified and controlled by means of gas puffing, supersonic molecule injection, pellet injection, ICRF and IBW heating as well as LHW heating and current drive; improved plasma confinements were achieved either by application of one of the above measures or by the combination of them, study of the effects of the characteristics of plasma profiles on plasma confinements were performed. The results show that in most of the improved confinement plasmas in HT-7, there are very steep and strong peeking electron temperature profiles in core plasma, and/or large decrease of local temperature in radius of 0.5 ~0.7a which makes temperature gradient steeper when improvements begin, as temperature profile evolves back to previous normal shape the improvements end. Electron density profile and soft X-ray profiles were studied as well. This research was supported under Natural Science Foundation of China contract No.19905010.

UCLA Tokamak Program Close Out Report.

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

Taylor, Robert John

2014-02-04

The results of UCLA experimental fusion program are summarized. Starting with smaller devices like Microtor, Macrotor, CCT and ending the research on the large (5 m) Electric Tokamak. CCT was the most diagnosed device for H-mode like physics and the effects of rotation induced radial fields. ICRF heating was also studied but plasma heating of University Type Tokamaks did not produce useful results due to plasma edge disturbances of the antennae. The Electric Tokamak produced better confinement in the seconds range. However, it presented very good particle confinement due to an "electric particle pinch". This effect prevented us from reachingmore » a quasi steady state. This particle accumulation effect was numerically explained by Shaing's enhanced neoclassical theory. The PI believes that ITER will have a good energy confinement time but deleteriously large particle confinement time and it will disrupt on particle pinching at nominal average densities. The US fusion research program did not study particle transport effects due to its undue focus on the physics of energy confinement time. Energy confinement time is not an issue for energy producing tokamaks. Controlling the ash flow will be very expensive.« less

Modular architecture of the T4 phage superfamily: A conserved core genome and a plastic periphery

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

Comeau, Andre M.; Bertrand, Claire; Letarov, Andrei

2007-06-05

Among the most numerous objects in the biosphere, phages show enormous diversity in morphology and genetic content. We have sequenced 7 T4-like phages and compared their genome architecture. All seven phages share a core genome with T4 that is interrupted by several hyperplastic regions (HPRs) where most of their divergence occurs. The core primarily includes homologues of essential T4 genes, such as the virion structure and DNA replication genes. In contrast, the HPRs contain mostly novel genes of unknown function and origin. A few of the HPR genes that can be assigned putative functions, such as a series of novelmore » Internal Proteins, are implicated in phage adaptation to the host. Thus, the T4-like genome appears to be partitioned into discrete segments that fulfil different functions and behave differently in evolution. Such partitioning may be critical for these large and complex phages to maintain their flexibility, while simultaneously allowing them to conserve their highly successful virion design and mode of replication.« less

NIMROD modeling of poloidal flow damping in tokamaks using kinetic closures

NASA Astrophysics Data System (ADS)

Jepson, J. R.; Hegna, C. C.; Held, E. D.

2017-10-01

Calculations of poloidal flow damping in a tokamak are undertaken using two different implementations of the ion drift kinetic equation (DKE) in the extended MHD code NIMROD. The first approach is hybrid fluid/kinetic and uses a Chapman Enskog-like (CEL) Ansatz. Closure of the evolving lower-order fluid moment equations for n, V , and T is provided by solutions to the ion CEL-DKE written in the macroscopic flow reference frame. The second implementation solves the DKE using a delta-f approach. Here, the delta-f distribution describes all of the information beyond a static, lowest-order Maxwellian. We compare the efficiency and accuracy of these two approaches for a simple initial value problem that monitors the relaxation of the poloidal flow profile in high- and low-aspect-ratio tokamak geometry. The computation results are compared against analytic predictions of time dependent closures for the parallel viscous force. Supported by DoE Grants DE-FG02-86ER53218 and DE-FG02-04ER54746.

Effects of high Z probe on plasma behavior in HT-6M tokamak

NASA Astrophysics Data System (ADS)

Li, J.; Gong, X.; Luo, L.; Yin, F. X.; Noda, N.; Wan, B.; Xu, W.; Gao, X.; Yin, F.; Jiang, J. G.; Wu, Z.; Zhao, J. Y.; Wu, M.; Liu, S.; Han, Y.

1997-02-01

Molybdenum and tungsten probes have been tested in HT-6M tokamak under various discharge conditions aiming to find out the conditions in which high Z PFC can be used without serious degradation of core plasma performance. In normal OH discharges, the degradation of core plasma performance was found only when the probe was inserted beyond 3.0 cm inside the last closed flux surface (LCFS). The plasma performance did not change with positive biasing to the probe, whereas central Te degraded during negative biasing of -100 V. The insertion of the Mo probe to 1.5 cm inside the LCFS made a change in the threshold power of the L-H transition in EOH discharges. These results suggest a certain operation range of the H-mode in the EOH discharge with the Mo probe in HT-6M.

2D electron density profile measurement in tokamak by laser-accelerated ion-beam probe.

PubMed

Chen, Y H; Yang, X Y; Lin, C; Wang, L; Xu, M; Wang, X G; Xiao, C J

2014-11-01

A new concept of Heavy Ion Beam Probe (HIBP) diagnostic has been proposed, of which the key is to replace the electrostatic accelerator of traditional HIBP by a laser-driven ion accelerator. Due to the large energy spread of ions, the laser-accelerated HIBP can measure the two-dimensional (2D) electron density profile of tokamak plasma. In a preliminary simulation, a 2D density profile was reconstructed with a spatial resolution of about 2 cm, and with the error below 15% in the core region. Diagnostics of 2D density fluctuation is also discussed.

On the applicability of the standard approaches for evaluating a neoclassical radial electric field in a tokamak edge region

DOE PAGES

Dorf, M. A.; Cohen, R. H.; Simakov, A. N.; ...

2013-08-27

The use of the standard approaches for evaluating a neoclassical radial electric field E r, i.e., the Ampere (or gyro-Poisson) equation, requires accurate calculation of the difference between the gyroaveraged electron and ion particle fluxes (or densities). In the core of a tokamak, the nontrivial difference appears only in high-order corrections to a local Maxwellian distribution due to the intrinsic ambipolarity of particle transport. The evaluation of such high-order corrections may be inconsistent with the accuracy of the standard long wavelength gyrokinetic equation (GKE), thus imposing limitations on the applicability of the standard approaches. However, in the edge of amore » tokamak, charge-exchange collisions with neutrals and prompt ion orbit losses can drive non-intrinsically ambipolar particle fluxes for which a nontrivial (E r-dependent) difference between the electron and ion fluxes appears already in a low order and can be accurately predicted by the long wavelength GKE. As a result, the parameter regimes where the radial electric field dynamics in the tokamak edge region is dominated by the non-intrinsically ambipolar processes, thus allowing for the use of the standard approaches, are discussed.« less

A power-balance model for local helicity injection startup in a spherical tokamak

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

Barr, Jayson L.; Bongard, Michael W.; Burke, Marcus G.

A 0D circuit model for predicting I p( t) in Local Helicity Injection (LHI) discharges is developed. Analytic formulas for estimating the surface flux of finite-A plasmas developed are modified and expanded to treat highly shaped, ultralow-A tokamak geometry using a database of representative equilibria. Model predictions are compared to sample LHI discharges in the A ~ 1 Pegasus spherical tokamak, and are found to agree within 15% of experimental I p( t). High performance LHI discharges are found to follow the Taylor relaxation current limit for approximately the first half of the current ramp, or I p ≲ 75more » kA. The second half of the current ramp follows a limit imposed by power-balance as plasmas expand from high- A to ultralow- A. Here, this shape evolution generates a significant drop in external plasma inductance, effectively using the plasma’s initially high inductance to drive the current ramp and provide > 70% of the current drive V-s. Projections using this model indicate the relative influences of higher helicity input rate and injector current on the attainable total plasma current.« less

A power-balance model for local helicity injection startup in a spherical tokamak

DOE PAGES

Barr, Jayson L.; Bongard, Michael W.; Burke, Marcus G.; ...

2018-05-15

A 0D circuit model for predicting I p( t) in Local Helicity Injection (LHI) discharges is developed. Analytic formulas for estimating the surface flux of finite-A plasmas developed are modified and expanded to treat highly shaped, ultralow-A tokamak geometry using a database of representative equilibria. Model predictions are compared to sample LHI discharges in the A ~ 1 Pegasus spherical tokamak, and are found to agree within 15% of experimental I p( t). High performance LHI discharges are found to follow the Taylor relaxation current limit for approximately the first half of the current ramp, or I p ≲ 75more » kA. The second half of the current ramp follows a limit imposed by power-balance as plasmas expand from high- A to ultralow- A. Here, this shape evolution generates a significant drop in external plasma inductance, effectively using the plasma’s initially high inductance to drive the current ramp and provide > 70% of the current drive V-s. Projections using this model indicate the relative influences of higher helicity input rate and injector current on the attainable total plasma current.« less

Measurements of impurity concentrations and transport in the Lithium Tokamak Experiment

NASA Astrophysics Data System (ADS)

Boyle, D. P.; Bell, R. E.; Kaita, R.; Lucia, M.; Schmitt, J. C.; Scotti, F.; Kubota, S.; Hansen, C.; Biewer, T. M.; Gray, T. K.

2016-10-01

The Lithium Tokamak Experiment (LTX) is a modest-sized spherical tokamak with all-metal plasma facing components (PFCs), uniquely capable of operating with large area solid and/or liquid lithium coatings essentially surrounding the entire plasma. This work presents measurements of core plasma impurity concentrations and transport in LTX. In discharges with solid Li coatings, volume averaged impurity concentrations were low but non-negligible, with 2 - 4 % Li, 0.6 - 2 % C, 0.4 - 0.7 % O, and Zeff < 1.2 . Transport was assessed using the TRANSP, NCLASS, and MIST codes. Collisions with the main H ions dominated the neoclassical impurity transport, and neoclassical transport coefficients calculated with NCLASS were similar across all impurity species and differed no more than a factor of two. However, time-independent simulations with MIST indicated that neoclassical theory did not fully capture the impurity transport and anomalous transport likely played a significant role in determining impurity profiles. Progress on additional analysis, including time-dependent impurity transport simulations and impurity measurements with liquid lithium coatings, and plans for diagnostic upgrades and future experiments in LTX- β will also be presented. This work supported by US DOE contracts DE-AC02-09CH11466 and DE-AC05-00OR22725.

Hybrid neural network for density limit disruption prediction and avoidance on J-TEXT tokamak

NASA Astrophysics Data System (ADS)

Zheng, W.; Hu, F. R.; Zhang, M.; Chen, Z. Y.; Zhao, X. Q.; Wang, X. L.; Shi, P.; Zhang, X. L.; Zhang, X. Q.; Zhou, Y. N.; Wei, Y. N.; Pan, Y.; J-TEXT team

2018-05-01

Increasing the plasma density is one of the key methods in achieving an efficient fusion reaction. High-density operation is one of the hot topics in tokamak plasmas. Density limit disruptions remain an important issue for safe operation. An effective density limit disruption prediction and avoidance system is the key to avoid density limit disruptions for long pulse steady state operations. An artificial neural network has been developed for the prediction of density limit disruptions on the J-TEXT tokamak. The neural network has been improved from a simple multi-layer design to a hybrid two-stage structure. The first stage is a custom network which uses time series diagnostics as inputs to predict plasma density, and the second stage is a three-layer feedforward neural network to predict the probability of density limit disruptions. It is found that hybrid neural network structure, combined with radiation profile information as an input can significantly improve the prediction performance, especially the average warning time ({{T}warn} ). In particular, the {{T}warn} is eight times better than that in previous work (Wang et al 2016 Plasma Phys. Control. Fusion 58 055014) (from 5 ms to 40 ms). The success rate for density limit disruptive shots is above 90%, while, the false alarm rate for other shots is below 10%. Based on the density limit disruption prediction system and the real-time density feedback control system, the on-line density limit disruption avoidance system has been implemented on the J-TEXT tokamak.

Three-dimensional observation of an helical hot structure during a sawtooth crash in the WT-3 tokamak.

PubMed

Yamaguchi, S; Igami, H; Tanaka, H; Maekawa, T

2004-07-23

Sawtooth crashes in an Ohmically heated plasma in the WT-3 tokamak have been observed by using soft x-ray computer tomography at three different poloidal cross sections around the torus. Initially, collapsing proceeds slowly with keeping the helical structure of an m = 1/n = 1 hot core around the torus. It accelerates as the helical hot structure is strongly deformed and fades away in the manner that the hot core at the high field side becomes obscure and disappears, while that at the low field side is deformed into a thin crescent aligned along the inversion circle, which survives even at the completion of the crash. Copyright 2004 The American Physical Society

First wall design of aluminum alloy R-tokamak

NASA Astrophysics Data System (ADS)

Hamada, Y.; Matsuoka, K.; Ogawa, Y.; Kitagawa, S.; Toi, K.; Yamazaki, K.; Abe, Y.; Amano, T.; Fujita, J.; Kaneko, O.; Kawahata, K.; Kuroda, T.; Matsuura, K.; Midzuno, Y.; Naitou, H.; Noda, N.; Ohkubo, K.; Oka, Y.; Sakurai, K.; Tanahashi, S.; Watari, T.

1984-05-01

A design study of a low-activation D-T tokamak Reacting Plasma Project In Nagoya has been finished. The study emphasizes the vacuum vessel and the bumper limiter. Our choice of materials (aluminum vacuum vessel, copper conductors, aluminum TF coil case and lead shield) results in a radiation level of about 1 × 10 -3 times that of a TFTR type design, and 1 × 10 -4 times that of JET type design, at 2 weeks after one D-T shot. Thick graphite tiles will be fixed directly on the aluminum vacuum vessel using aluminum spring washers and bolts. With this simplified structure of the bumper limiter, the inner surface temperature of the thick aluminum vacuum vessel will be less than 120°C which is required to reduce the overaging effect of the aluminum alloy.

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.

Transport properties of interacting magnetic islands in tokamak plasmas

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

Gianakon, T.A.; Callen, J.D.; Hegna, C.C.

1993-10-01

This paper explores the equilibrium and transient transport properties of a mixed magnetic topology model for tokamak equilibria. The magnetic topology is composed of a discrete set of mostly non-overlapping magnetic islands centered on the low-order rational surfaces. Transport across the island regions is fast due to parallel transport along the stochastic magnetic field lines about the separatrix of each island. Transport between island regions is assumed to be slow due to a low residual cross-field transport. In equilibrium, such a model leads to: a nonlinear dependence of the heat flux on the pressure gradient; a power balance diffusion coefficientmore » which increases from core to edge; and profile resiliency. Transiently, such a model also exhibits a heat pulse diffusion coefficient larger than the power balance diffusion coefficient.« less

Effects of Resonant Helical Field on Toroidal Field Ripple in IR-T1 Tokamak

NASA Astrophysics Data System (ADS)

Mahdavipour, B.; Salar Elahi, A.; Ghoranneviss, M.

2018-02-01

The toroidal magnetic field which is created by toroidal coils has the ripple in torus space. This magnetic field ripple has an importance in plasma equilibrium and stability studies in tokamak. In this paper, we present the investigation of the interaction between the toroidal magnetic field ripple and resonant helical field (RHF). We have estimated the amplitude of toroidal field ripples without and with RHF (with different q = m/n) ( m = 2, m = 3, m = 4, m = 5, m = 2 & 3, n = 1) using “Comsol Multiphysics” software. The simulations show that RHF has effects on the toroidal ripples.

Physics of the Tokamak Pedestal, and Implications for Magnetic Fusion Energy

NASA Astrophysics Data System (ADS)

Snyder, Philip

2017-10-01

High performance in tokamaks is achieved via the spontaneous formation of a transport barrier in the outer few percent of the confined plasma. This narrow insulating layer, referred to as a ``pedestal,'' typically results in a >30x increase in pressure across a 0.4-5cm layer. Predicted fusion power scales with the square of the pedestal top pressure (or ``pedestal height''), hence a fusion reactor strongly benefits from a high pedestal, provided this can be attained without large Edge Localized Modes (ELMs), which may erode plasma facing materials. The overlap of drift orbit, turbulence, and equilibrium scales across this narrow layer leads to rich and complex physics, and challenges traditional analytic and computational approaches. We review studies employing gyrokinetic, neoclassical, MHD, and other methods, which have explored how a range of instabilities, influenced by complex geometry, and strong ExB flows and bootstrap current, drive transport across the pedestal and guide its structure and dynamics. Development of high resolution diagnostics, and coordinated experiments on several tokamaks, have validated understanding of important aspects of the physics, while highlighting open issues. A predictive model (EPED) has proven capable of predicting the pedestal height and width to 20-25% accuracy in large statistical studies. This model was used to predict a new, high pedestal ``Super H-Mode'' regime, which was subsequently discovered on DIII-D, and motivated experiments on Alcator C-Mod which achieved world record, reactor relevant pedestal pressure. We review open issues including improved formalism, particle and momentum transport, the role of neutrals and impurities, ELM control, and pedestal formation. Finally we discuss coupling pedestal and core predictive models to enable more comprehensive optimization of the tokamak fusion concept. Supported by the US DOE under DE-FG02-95ER54309, FC02-06ER54873, DE-FC02-04ER54698, DE-FC02-99ER54512.

First neutral beam injection experiments on KSTAR tokamak.

PubMed

Jeong, S H; Chang, D H; Kim, T S; In, S R; Lee, K W; Jin, J T; Chang, D S; Oh, B H; Bae, Y S; Kim, J S; Park, H T; Watanabe, K; Inoue, T; Kashiwagi, M; Dairaku, M; Tobari, H; Hanada, M

2012-02-01

The first neutral beam (NB) injection system of the Korea Superconducting Tokamak Advanced Research (KSTAR) tokamak was partially completed in 2010 with only 1∕3 of its full design capability, and NB heating experiments were carried out during the 2010 KSTAR operation campaign. The ion source is composed of a JAEA bucket plasma generator and a KAERI large multi-aperture accelerator assembly, which is designed to deliver a 1.5 MW, NB power of deuterium at 95 keV. Before the beam injection experiments, discharge, and beam extraction characteristics of the ion source were investigated. The ion source has good beam optics in a broad range of beam perveance. The optimum perveance is 1.1-1.3 μP, and the minimum beam divergence angle measured by the Doppler shift spectroscopy is 0.8°. The ion species ratio is D(+):D(2)(+):D(3)(+) = 75:20:5 at beam current density of 85 mA/cm(2). The arc efficiency is more than 1.0 A∕kW. In the 2010 KSTAR campaign, a deuterium NB power of 0.7-1.5 MW was successfully injected into the KSTAR plasma with a beam energy of 70-90 keV. L-H transitions were observed within a wide range of beam powers relative to a threshold value. The edge pedestal formation in the T(i) and T(e) profiles was verified through CES and electron cyclotron emission diagnostics. In every deuterium NB injection, a burst of D-D neutrons was recorded, and increases in the ion temperature and plasma stored energy were found.

Response of plasma rotation to resonant magnetic perturbations in J-TEXT tokamak

NASA Astrophysics Data System (ADS)

Yan, W.; Chen, Z. Y.; Huang, D. W.; Hu, Q. M.; Shi, Y. J.; Ding, Y. H.; Cheng, Z. F.; Yang, Z. J.; Pan, X. M.; Lee, S. G.; Tong, R. H.; Wei, Y. N.; Dong, Y. B.; J-TEXT Team

2018-03-01

The response of plasma toroidal rotation to the external resonant magnetic perturbations (RMP) has been investigated in Joint Texas Experimental Tokamak (J-TEXT) ohmic heating plasmas. For the J-TEXT’s plasmas without the application of RMP, the core toroidal rotation is in the counter-current direction while the edge rotation is near zero or slightly in the co-current direction. Both static RMP experiments and rotating RMP experiments have been applied to investigate the plasma toroidal rotation. The core toroidal rotation decreases to lower level with static RMP. At the same time, the edge rotation can spin to more than 20 km s-1 in co-current direction. On the other hand, the core plasma rotation can be slowed down or be accelerated with the rotating RMP. When the rotating RMP frequency is higher than mode frequency, the plasma rotation can be accelerated to the rotating RMP frequency. The plasma confinement is improved with high frequency rotating RMP. The plasma rotation is decelerated to the rotating RMP frequency when the rotating RMP frequency is lower than the mode frequency. The plasma confinement also degrades with low frequency rotating RMP.

Core/shell Fe3O4/Gd2O3 nanocubes as T1-T2 dual modal MRI contrast agents

NASA Astrophysics Data System (ADS)

Li, Fenfen; Zhi, Debo; Luo, Yufeng; Zhang, Jiqian; Nan, Xiang; Zhang, Yunjiao; Zhou, Wei; Qiu, Bensheng; Wen, Longping; Liang, Gaolin

2016-06-01

T1-T2 dual modal magnetic resonance imaging (MRI) has attracted considerable interest because it offers complementary diagnostic information, leading to more precise diagnosis. To date, a number of nanostructures have been reported as T1-T2 dual modal MR contrast agents (CAs). However, hybrids of nanocubes with both iron and gadolinium (Gd) elements as T1-T2 dual modal CAs have not been reported. Herein, we report the synthesis of novel core/shell Fe3O4/Gd2O3 nanocubes as T1-T2 dual-modal CAs and their application for enhanced T1-T2 MR imaging of rat livers. A relaxivity study at 1.5 T indicated that our Fe3O4/Gd2O3 nanocubes have an r1 value of 45.24 mM-1 s-1 and an r2 value of 186.51 mM-1 s-1, which were about two folds of those of Gd2O3 nanoparticles and Fe3O4 nanocubes, respectively. In vivo MR imaging of rats showed both T1-positive and T2-negative contrast enhancements in the livers. We envision that our Fe3O4/Gd2O3 nanocubes could be applied as T1-T2 dual modal MR CAs for a wide range of theranostic applications in the near future.T1-T2 dual modal magnetic resonance imaging (MRI) has attracted considerable interest because it offers complementary diagnostic information, leading to more precise diagnosis. To date, a number of nanostructures have been reported as T1-T2 dual modal MR contrast agents (CAs). However, hybrids of nanocubes with both iron and gadolinium (Gd) elements as T1-T2 dual modal CAs have not been reported. Herein, we report the synthesis of novel core/shell Fe3O4/Gd2O3 nanocubes as T1-T2 dual-modal CAs and their application for enhanced T1-T2 MR imaging of rat livers. A relaxivity study at 1.5 T indicated that our Fe3O4/Gd2O3 nanocubes have an r1 value of 45.24 mM-1 s-1 and an r2 value of 186.51 mM-1 s-1, which were about two folds of those of Gd2O3 nanoparticles and Fe3O4 nanocubes, respectively. In vivo MR imaging of rats showed both T1-positive and T2-negative contrast enhancements in the livers. We envision that our Fe3O4/Gd2O3 nanocubes

A spatially resolving x-ray crystal spectrometer for measurement of ion-temperature and rotation-velocity profiles on the Alcator C-Mod tokamak.

PubMed

Hill, K W; Bitter, M L; Scott, S D; Ince-Cushman, A; Reinke, M; Rice, J E; Beiersdorfer, P; Gu, M-F; Lee, S G; Broennimann, Ch; Eikenberry, E F

2008-10-01

A new spatially resolving x-ray crystal spectrometer capable of measuring continuous spatial profiles of high resolution spectra (lambda/d lambda>6000) of He-like and H-like Ar K alpha lines with good spatial (approximately 1 cm) and temporal (approximately 10 ms) resolutions has been installed on the Alcator C-Mod tokamak. Two spherically bent crystals image the spectra onto four two-dimensional Pilatus II pixel detectors. Tomographic inversion enables inference of local line emissivity, ion temperature (T(i)), and toroidal plasma rotation velocity (upsilon(phi)) from the line Doppler widths and shifts. The data analysis techniques, T(i) and upsilon(phi) profiles, analysis of fusion-neutron background, and predictions of performance on other tokamaks, including ITER, will be presented.

Internal transport barrier in tokamak and helical plasmas

NASA Astrophysics Data System (ADS)

Ida, K.; Fujita, T.

2018-03-01

The differences and similarities between the internal transport barriers (ITBs) of tokamak and helical plasmas are reviewed. By comparing the characteristics of the ITBs in tokamak and helical plasmas, the mechanisms of the physics for the formation and dynamics of the ITB are clarified. The ITB is defined as the appearance of discontinuity of temperature, flow velocity, or density gradient in the radius. From the radial profiles of temperature, flow velocity, and density the ITB is characterized by the three parameters of normalized temperature gradient, R/{L}T, the location, {ρ }{ITB}, and the width, W/a, and can be expressed by ‘weak’ ITB (small R/{L}T) or ‘strong’ (large R/{L}T), ‘small’ ITB (small {ρ }{ITB}) or ‘large’ ITB (large {ρ }{ITB}), and ‘narrow’ (small W/a) or ‘wide’ (large W/a). Three key physics elements for the ITB formation, radial electric field shear, magnetic shear, and rational surface (and/or magnetic island) are described. The characteristics of electron and ion heat transport and electron and impurity transport are reviewed. There are significant differences in ion heat transport and electron heat transport. The dynamics of ITB formation and termination is also discussed. The emergence of the location of the ITB is sometimes far inside the ITB foot in the steady-state phase and the ITB region shows radial propagation during the formation of the ITB. The non-diffusive terms in momentum transport and impurity transport become more dominant in the plasma with the ITB. The reversal of the sign of non-diffusive terms in momentum transport and impurity transport associated with the formation of the ITB reported in helical plasma is described. Non-local transport plays an important role in determining the radial profile of temperature and density. The spontaneous change in temperature curvature (second radial derivative of temperature) in the ITB region is described. In addition, the key parameters of the control of the

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.

Stability at high performance in the MAST spherical tokamak

NASA Astrophysics Data System (ADS)

Buttery, R. J.; Akers, R.; Arends, E.; Conway, N. J.; Counsell, G. F.; Cunningham, G.; Gimblett, C. G.; Gryaznevich, M.; Hastie, R. J.; Hole, M. J.; Lehane, I.; Martin, R.; Patel, A.; Pinfold, T.; Sauter, O.; Taylor, D.; Turri, G.; Valovic, M.; Walsh, M. J.; Wilson, H. R.; MAST Team

2004-09-01

The development of reliable H-modes on MAST, together with advances in heating power and a range of high spatial resolution diagnostics, has provided a platform to enable MAST to address some of the most important issues of tokamak stability. In particular the high bgr potential of the spherical tokamak is highlighted with stable operation at bgrN ~ 5-6, bgrT ~ 16% and bgrp up to ~2. Magnetic diagnostic evaluation of the global bgr parameters is independently confirmed by kinetic profile data. Calculations indicate that the bgrN values are in the vicinity of no-wall stability limits. Studies of neoclassical tearing modes (NTMs) have been extended to explore their effects and develop avoidance strategies. Experiments have demonstrated that sawteeth play a strong role in triggering NTMs—by avoiding large sawteeth a much higher bgrN value has been reached. The significance of NTMs is confirmed, with large islands observed using the 300 point Thomson scattering diagnostic, and locking of large n = 1 modes frequently leading to disruptions, which become more rapid at low q95. The role of error fields has been explored. H-mode plasmas are also limited by edge localized modes (ELMs), with confinement degraded as the ELM frequency rises. However, in contrast to the conventional tokamak, the ELMs in high performing regimes on MAST (HIPB98Y2 ~ 1) appear to be type III in nature. Modelling using the ELITE code, which incorporates finite n corrections, identifies instability to peeling modes, consistent with a type III interpretation. It also shows considerable scope to raise pressure gradients before ballooning type modes (perhaps associated with type I ELMs) occur. The calculations show that narrow pedestals can support much stronger pressure gradients than might be expected from simple n = infin ballooning calculations. Finally sawteeth are shown to degrade confinement by ~10-15% in particular cases examined. They are observed not to remove the q = 1 surface in the cases

Energy, Vacuum, Gas Fueling, and Security Systems for the Spherical Tokamak MEDUSA-CR

NASA Astrophysics Data System (ADS)

Gonzalez, Jeferson; Soto, Christian; Carvajal, Johan; Ribeiro, Celso

2013-10-01

The former spherical tokamak (ST) MEDUSA (Madison EDUcation Small Aspect.ratio tokamak, R < 0.14 m, a < 0.10 m, BT < 0.5 T, Ip < 40 kA, 3 ms pulse) is being recommissioned in Costa Rica Institute of Technology. The main objectives of the MEDUSA-CR project are training and to clarify several issues in relevant physics for conventional and mainly STs, including beta studies in bean-shaped ST plasmas, transport, heating and current drive via Alfvén wave, and natural divertor STs with ergodic magnetic limiter. We present here the energy, vacuum, gas fueling, and security systems for MEDUSA-CR device. The interface with the control and data acquisition systems based on National Instruments (NI) software (LabView) and hardware (on loan to our laboratory via NI-Costa Rica) are also presented. VIE-ITCR, IAEA-CRP contract 17592, National Instruments of Costa Rica.

Fishbone Oscillations in the Experimental Advanced Superconductivity Tokamak

NASA Astrophysics Data System (ADS)

Xu, Li-Qing; Hu, Li-Qun; Yuan, Yi; Li, Ying-Ying; Zhong, Guo-Qiang; Liu, Hai-Qing; Chen, Kai-Yun; Shi, Tong-Hui; Duan, Yan-Min

2018-03-01

A fishbone oscillation was observed in the neutral beam injection plasma at Experimental Advanced Superconductivity Tokamak (EAST). This m = 1/n = 1 ( m, n: poloidal, toroidal mode numbers, respectively) typical internal kink mode travels in the ion-diamagnetism direction in the poloidal section with a rotation speed close to the ion diamagnetic drift frequency. A high thermal plasma beta and high amounts of energetic ions are necessary for the mode to develop. Fishbone oscillations can expel heavy impurities in the core, which favors sustaining a high-performance plasma. The born frequency of the fishbone oscillation is the ion diamagnetic drift frequency and the chirping down of the frequency during the initial growth phase is the result of a drop in iondiamagnetic drift frequency. The excitation energy is thought to be due to the thermal plasma pressure gradient; however, the development of a fishbone oscillation is related to energetic ions.

Edge-localized-modes in tokamaks

DOE PAGES

Leonard, Anthony W.

2014-09-11

Edge-localized-modes (ELMs) are a ubiquitous feature of H-mode in tokamaks. When gradients in the H-mode transport barrier grow to exceed the MHD stability limit the ELM instability grows explosively rapidly transporting energy and particles onto open field lines and material surfaces. Though ELMs provide additional particle and impurity transport through the H-mode transport barrier, enabling steady operation, the resulting heat flux transients to plasma facing surfaces project to large amplitude in future low collisionality burning plasma tokamaks. Measurements of the ELM heat flux deposition onto material surfaces in the divertor and main chamber indicate significant broadening compared to inter-ELM heatmore » flux, with a timescale for energy deposition that is consistent with sonic ion flow and numerical simulation. Comprehensive ELM simulation is highlighting the important physics processes of ELM transport including parallel transport due to magnetic reconnection and turbulence resulting from collapse of the H-mode transport barrier. As a result, encouraging prospects for ELM control and/or suppression in future tokamaks include intrinsic modes of ELM free operation, ELM triggering with frequent small pellet injection and the application of 3D magnetic fields.« less

Fitting of the Thomson scattering density and temperature profiles on the COMPASS tokamak

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

Stefanikova, E.; Division of Fusion Plasma Physics, KTH Royal Institute of Technology, SE-10691 Stockholm; Peterka, M.

2016-11-15

A new technique for fitting the full radial profiles of electron density and temperature obtained by the Thomson scattering diagnostic in H-mode discharges on the COMPASS tokamak is described. The technique combines the conventionally used modified hyperbolic tangent function for the edge transport barrier (pedestal) fitting and a modification of a Gaussian function for fitting the core plasma. Low number of parameters of this combined function and their straightforward interpretability and controllability provide a robust method for obtaining physically reasonable profile fits. Deconvolution with the diagnostic instrument function is applied on the profile fit, taking into account the dependence onmore » the actual magnetic configuration.« less

Time dependent 14 MeV neutrons measurement using a polycrystalline chemical vapor deposited diamond detector at the JET tokamak

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

Angelone, M.; Pillon, M.; Bertalot, L.

A polycrystalline chemical vapor deposited (CVD) diamond detector was installed on a JET tokamak in order to monitor the time dependent 14 MeV neutron emission produced by D-T plasma pulses during the Trace Tritium Experiment (TTE) performed in October 2003. This was the first tentative ever attempted to use a CVD diamond detector as neutron monitor in a tokamak environment. Despite its small active volume, the detector was able to detect the 14 MeV neutron emission (>1.0x10{sup 15} n/shot) with good reliability and stability during the experimental campaign that lasted five weeks. The comparison with standard silicon detectors presently usedmore » at JET as 14 MeV neutron monitors is reported, showing excellent correlation between the measurements. The results prove that CVD diamond detectors can be reliably used in a tokamak environment and therefore confirm the potential of this technology for next step machines like ITER.« less

Department of Defense In-House RDT&E Activities

DTIC Science & Technology

1979-10-30

OIVAI .. I-:JNLI’IONS/EQJIPMtN’r/CAPAt3ILIrflt.4 tNC’i4PAb-Eý, APPLO RUA INCL WRC DEV AND) SPT IN INTI:RLO got: ~ISI6 ANALYSIS. JAM,IJEtPSi(7 iNrt7L...LU*. CAPT. R. 4. tsROWN TECH1.L1R. MR. THOMAS RAMIREZ F. ~PROGRAM LJATA BY FISCAL YEAR (MILLION S PiKOtRAM 1979 IQ80fk~(ACTUJAL) (ACT + EsT) TUTAL

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

Muscle spindles exhibit core lesions and extensive degeneration of intrafusal fibers in the Ryr1{sup I4895T/wt} mouse model of core myopathy

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

Zvaritch, Elena; MacLennan, David H., E-mail: david.maclennan@utoronto.ca

Muscle spindles from the hind limb muscles of adult Ryr1{sup I4895T/wt} (IT/+) mice exhibit severe structural abnormalities. Up to 85% of the spindles are separated from skeletal muscle fascicles by a thick layer of connective tissue. Many intrafusal fibers exhibit degeneration, with Z-line streaming, compaction and collapse of myofibrillar bundles, mitochondrial clumping, nuclear shrinkage and pyknosis. The lesions resemble cores observed in the extrafusal myofibers of this animal model and of core myopathy patients. Spindle abnormalities precede those in extrafusal fibers, indicating that they are a primary pathological feature in this murine Ryr1-related core myopathy. Muscle spindle involvement, if confirmedmore » for human core myopathy patients, would provide an explanation for an array of devastating clinical features characteristic of these diseases and provide novel insights into the pathology of RYR1-related myopathies. - Highlights: • Muscle spindles exhibit structural abnormalities in a mouse model of core myopathy. • Myofibrillar collapse and mitochondrial clumping is observed in intrafusal fibers. • Myofibrillar degeneration follows a pattern similar to core formation in extrafusal myofibers. • Muscle spindle abnormalities are a part of the pathological phenotype in the mouse model of core myopathy. • Direct involvement of muscle spindles in the pathology of human RYR1-related myopathies is proposed.« less

Lower Hybrid Wave Induced SOL Emissivity Variation at High Density on the Alcator C-Mod Tokamak

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

Faust, I.; Terry, J. L.; Reinke, M. L.

Lower Hybrid Current Drive (LHCD) in the Alcator C-Mod tokamak provides current profile control for the generation of Advanced Tokamak (AT) plasmas. Non-thermal electron bremsstrahlung emission decreases dramatically at n-bar{sub e}>1{center_dot}10{sup 20}[m{sup -3}] for diverted discharges, indicating low current drive efficiency. It is suggested that Scrape-Off-Layer (SOL) collisional absorption of LH waves is the cause for the absence of non-thermal electrons at high density. VUV and visible spectroscopy in the SOL provide direct information on collision excitation processes. Deuterium Balmer-, Lyman- and He-I transition emission measurements were used for initial characterization of SOL electron-neutral collisional absorption. Data from Helium andmore » Deuterium LHCD discharges were characterized by an overall increase in the emissivity as well as an outward radial shift in the emissivity profile with increasing plasma density and applied LHCD power. High-temperature, high-field (T{sub e} = 5keV,B{sub t} = 8T) helium discharges at high density display increased non-thermal signatures as well as reduced SOL emissivity. Variations in emissivity due to LHCD were seen in SOL regions not magnetically connected to the LH Launcher, indicating global SOL effects due to LHCD.« less

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.

Gamma ray imager on the DIII-D tokamak

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

Pace, D. C., E-mail: pacedc@fusion.gat.com; Taussig, D.; Eidietis, N. W.

2016-04-15

A gamma ray camera is built for the DIII-D tokamak [J. Luxon, Nucl. Fusion 42, 614 (2002)] that provides spatial localization and energy resolution of gamma flux by combining a lead pinhole camera with custom-built detectors and optimized viewing geometry. This diagnostic system is installed on the outer midplane of the tokamak such that its 123 collimated sightlines extend across the tokamak radius while also covering most of the vertical extent of the plasma volume. A set of 30 bismuth germanate detectors can be secured in any of the available sightlines, allowing for customizable coverage in experiments with runaway electronsmore » in the energy range of 1–60 MeV. Commissioning of the gamma ray imager includes the quantification of electromagnetic noise sources in the tokamak machine hall and a measurement of the energy spectrum of background gamma radiation. First measurements of gamma rays coming from the plasma provide a suitable testbed for implementing pulse height analysis that provides the energy of detected gamma photons.« less

Gamma ray imager on the DIII-D tokamak

DOE PAGES

Pace, D. C.; Cooper, C. M.; Taussig, D.; ...

2016-04-13

A gamma ray camera is built for the DIII-D tokamak [J. Luxon, Nucl. Fusion 42, 614 (2002)] that provides spatial localization and energy resolution of gamma flux by combining a lead pinhole camera with custom-built detectors and optimized viewing geometry. This diagnostic system is installed on the outer midplane of the tokamak such that its 123 collimated sightlines extend across the tokamak radius while also covering most of the vertical extent of the plasma volume. A set of 30 bismuth germanate detectors can be secured in any of the available sightlines, allowing for customizable coverage in experiments with runaway electronsmore » in the energy range of 1- 60 MeV. Commissioning of the gamma ray imager includes the quantification of electromagnetic noise sources in the tokamak machine hall and a measurement of the energy spectrum of background gamma radiation. In conclusion, first measurements of gamma rays coming from the plasma provide a suitable testbed for implementing pulse height analysis that provides the energy of detected gamma photons.« less

Tokamak und Stellarator - zwei Wege zur Fusionsenergie: Fusionsforschung

NASA Astrophysics Data System (ADS)

Milch, Isabella

2006-07-01

Im Laufe der Fusionsforschung haben sich zwei Bautypen für ein zukünftiges Kraftwerk als besonders aussichtsreich erwiesen: Tokamak und Stellarator. Mit dem geplanten Tokamak-Experimentalreaktor ITER steht die internationale Fusionsforschung vor der Demonstration eines Energie liefernden Plasmas. Parallel soll die in Greifswald entstehende Forschungsanlage Wendelstein 7-X die Kraftwerkstauglichkeit des alternativen Bauprinzips der Stellaratoren zeigen.

P- T- t constraints on the development of the Doi Inthanon metamorphic core complex domain and implications for the evolution of the western gneiss belt, northern Thailand

NASA Astrophysics Data System (ADS)

Macdonald, A. S.; Barr, S. M.; Miller, B. V.; Reynolds, P. H.; Rhodes, B. P.; Yokart, B.

2010-01-01

The western gneiss belt in northern Thailand is exposed within two overlapping Cenozoic structural domains: the extensional Doi Inthanon metamorphic core complex domain located west of the Chiang Mai basin, and the Mae Ping strike-slip fault domain located west of the Tak batholith. New P- T estimates and U-Pb and 40Ar/ 39Ar age determinations from the Doi Inthanon domain show that the gneiss there records a complex multi-stage history that can be represented by a clockwise P- T- t path. U-Pb zircon and titanite dating of mylonitic calc-silicate gneiss from the Mae Wang area of the complex indicates that the paragneissic sequence experienced high-grade, medium-pressure metamorphism (M1) in the Late Triassic - Early Jurassic (ca. 210 Ma), in good agreement with previously determined zircon ages from the underlying core orthogneiss exposed on Doi Inthanon. Late Cretaceous monazite ages of 84 and 72 Ma reported previously from the core orthogneiss are attributed to a thermal overprint (M2) to upper-amphibolite facies in the sillimanite field. U-Pb zircon and monazite dating of granitic mylonite from the Doi Suthep area of the complex provides an upper age limit of 40 Ma (Late Eocene) for the early stage(s) of development of the actual core complex, by initially ductile, low-angle extensional shearing under lower amphibolite-facies conditions (M3), accompanied by near-isothermal diapiric rise and decompression melting. 40Ar/ 39Ar laserprobe dating of muscovite from both Doi Suthep and Doi Inthanon provided Miocene ages of ca. 26-15 Ma, representing cooling through the ca. 350 °C isotherm and marking late-stage development of the core complex by detachment faulting of the cover rocks and isostatic uplift of the sheared core zone and mantling gneisses in the footwall. Similarities in the thermochronology of high-grade gneisses exposed in the core complex and shear zone domains in the western gneiss belt of northern Thailand (and also in northern Vietnam, Laos, Yunnan

Resistive edge mode instability in stellarator and tokamak geometries

NASA Astrophysics Data System (ADS)

Mahmood, M. Ansar; Rafiq, T.; Persson, M.; Weiland, J.

2008-09-01

Geometrical effects on linear stability of electrostatic resistive edge modes are investigated in the three-dimensional Wendelstein 7-X stellarator [G. Grieger et al., Plasma Physics and Controlled Nuclear Fusion Research 1990 (International Atomic Energy Agency, Vienna, 1991), Vol. 3, p. 525] and the International Thermonuclear Experimental Reactor [Progress in the ITER Physics Basis, Nucl. Fusion 7, S1, S285 (2007)]-like equilibria. An advanced fluid model is used for the ions together with the reduced Braghinskii equations for the electrons. Using the ballooning mode representation, the drift wave problem is set as an eigenvalue equation along a field line and is solved numerically using a standard shooting technique. A significantly larger magnetic shear and a less unfavorable normal curvature in the tokamak equilibrium are found to give a stronger finite-Larmor radius stabilization and a more narrow mode spectrum than in the stellarator. The effect of negative global magnetic shear in the tokamak is found to be stabilizing. The growth rate on a tokamak magnetic flux surface is found to be comparable to that on a stellarator surface with the same global magnetic shear but the eigenfunction in the tokamak is broader than in the stellarator due to the presence of large negative local magnetic shear (LMS) on the tokamak surface. A large absolute value of the LMS in a region of unfavorable normal curvature is found to be stabilizing in the stellarator, while in the tokamak case, negative LMS is found to be stabilizing and positive LMS destabilizing.

Observations of compound sawteeth in ion cyclotron resonant heating plasma using ECE imaging on experimental advanced superconducting tokamak

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

Hussain, Azam; Zhao, Zhenling; Xie, Jinlin, E-mail: jlxie@ustc.edu.cn

The spatial and temporal evolutions of compound sawteeth were directly observed using 2D electron cyclotron emission imaging on experimental advanced superconducting tokamak. The compound sawtooth consists of partial and full collapses. After partial collapse, the hot core survives as only a small amount of heat disperses outwards, whereas in the following full collapse a large amount of heat is released and the hot core dissipates. The presence of two q = 1 surfaces was not observed. Instead, the compound sawtooth occurs mainly at the beginning of an ion cyclotron resonant frequency heating pulse and during the L-H transition phase, which may bemore » related to heat transport suppression caused by a decrease in electron heat diffusivity.« less

H-Mode Behavior Induced by Modulated Toroidal Current on HT-7 and HT-6M Tokamak

NASA Astrophysics Data System (ADS)

Mao, J. S.; Luo, J. R.; Xu, Y. H.; Zhao, J. Y.; Zhang, X. M.; Li, J. G.; Zhang, X. M.; Gao, X.; Li, Y. D.; Jie, Y. X.; Wu, Z. W.; Hu, L. Q.; Liu, S. X.; Zhang, X. D.; Bao, Y.; Yang, K.; Wang, G. X.; Chen, L.; Shi, Y. J.; Qin, P. J.; Gu, X. M.; Cui, N. Z.; Fan, H. Y.; Chen, Y. F.; Xia, C. Y.; Ruan, H. L.; Tong, X. D.; Phillips, P. E.

2001-10-01

An improved Ohmic confinement phase (similar to H-mode) has been observed during Modulating Toroidal Current on the Hefei Tokamak-6M (HT-6M) and Hefei super-conducting Tokamak-7 (HT-7). This improved plasma confinement phase is characterized by: (a) an increase in ne and T_e(0); (b) reduced H_α radiation from the edge; (c) steeper density and temperature profiles at the edge; (d) a more negative radial electric field inside the limiter; (e) a deeper electrostatic potential well at the edge; (f) reduced magnetic fluctuations at the edge; (g) MHD suppressing; (h) and by an increase in global energy confinement time, τ _e, by 27%-45%. The well-like structure of the radial electric field E_r, appears at an L-H like transition.

Structured Cable for High-Current Coils of Tokamaks

NASA Astrophysics Data System (ADS)

Benson, Christopher; McIntyre, Peter; Sattarov, Akhdiyor; Mann, Thomas

2011-10-01

The 45 kA superconducting cable for the ITER central solenoid coil has yielded questionable results in two recent tests. In both cases the cable Tc increased after cycling only a fraction of the design life, indicating degradation due to fatigue and fracture among the superconducting strands. The Accelerator Research Lab at Texas A&M University is developing a design for a Nb3Sn structured cable suitable for such tokamak coils. The superconductor is configured in 6 sub-cables, and each subcable is supported within a channel of a central support structure within a high-strength armor sheath. The structured cable addresses two issues that are thought to compromise opposition at high current. The strands are supported without cross-overs (which produce stress concentration); and armor sheath and core structure bypass stress through the coil and among subcables so that the stress within each subcable is only what is produced directly upon it. Details of the design and plans for development will be presented.

k-t Acceleration in pure phase encode MRI to monitor dynamic flooding processes in rock core plugs

NASA Astrophysics Data System (ADS)

Xiao, Dan; Balcom, Bruce J.

2014-06-01

Monitoring the pore system in sedimentary rocks with MRI when fluids are introduced is very important in the study of petroleum reservoirs and enhanced oil recovery. However, the lengthy acquisition time of each image, with pure phase encode MRI, limits the temporal resolution. Spatiotemporal correlations can be exploited to undersample the k-t space data. The stacked frames/profiles can be well approximated by an image matrix with rank deficiency, which can be recovered by nonlinear nuclear norm minimization. Sparsity of the x-t image can also be exploited for nonlinear reconstruction. In this work the results of a low rank matrix completion technique were compared with k-t sparse compressed sensing. These methods are demonstrated with one dimensional SPRITE imaging of a Bentheimer rock core plug and SESPI imaging of a Berea rock core plug, but can be easily extended to higher dimensionality and/or other pure phase encode measurements. These ideas will enable higher dimensionality pure phase encode MRI studies of dynamic flooding processes in low magnetic field systems.

Multi-core and GPU accelerated simulation of a radial star target imaged with equivalent t-number circular and Gaussian pupils

NASA Astrophysics Data System (ADS)

Greynolds, Alan W.

2013-09-01

Results from the GelOE optical engineering software are presented for the through-focus, monochromatic coherent and polychromatic incoherent imaging of a radial "star" target for equivalent t-number circular and Gaussian pupils. The FFT-based simulations are carried out using OpenMP threading on a multi-core desktop computer, with and without the aid of a many-core NVIDIA GPU accessing its cuFFT library. It is found that a custom FFT optimized for the 12-core host has similar performance to a simply implemented 256-core GPU FFT. A more sophisticated version of the latter but tuned to reduce overhead on a 448-core GPU is 20 to 28 times faster than a basic FFT implementation running on one CPU core.

A review of radiative detachment studies in tokamak advanced magnetic divertor configurations

DOE PAGES

Soukhanovskii, V. A.

2017-04-28

The present vision for a plasma–material interface in the tokamak is an axisymmetric poloidal magnetic X-point divertor. Four tasks are accomplished by the standard poloidal X-point divertor: plasma power exhaust; particle control (D/T and He pumping); reduction of impurity production (source); and impurity screening by the divertor scrape-off layer. A low-temperature, low heat flux divertor operating regime called radiative detachment is viewed as the main option that addresses these tasks for present and future tokamaks. Advanced magnetic divertor configuration has the capability to modify divertor parallel and cross-field transport, radiative and dissipative losses, and detachment front stability. Advanced magnetic divertormore » configurations are divided into four categories based on their salient qualitative features: (1) multiple standard X-point divertors; (2) divertors with higher order nulls; (3) divertors with multiple X-points; and (4) long poloidal leg divertors (and also with multiple X-points). As a result, this paper reviews experiments and modeling in the area of radiative detachment in the advanced magnetic divertor configurations.« less

A review of radiative detachment studies in tokamak advanced magnetic divertor configurations

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

Soukhanovskii, V. A.

The present vision for a plasma–material interface in the tokamak is an axisymmetric poloidal magnetic X-point divertor. Four tasks are accomplished by the standard poloidal X-point divertor: plasma power exhaust; particle control (D/T and He pumping); reduction of impurity production (source); and impurity screening by the divertor scrape-off layer. A low-temperature, low heat flux divertor operating regime called radiative detachment is viewed as the main option that addresses these tasks for present and future tokamaks. Advanced magnetic divertor configuration has the capability to modify divertor parallel and cross-field transport, radiative and dissipative losses, and detachment front stability. Advanced magnetic divertormore » configurations are divided into four categories based on their salient qualitative features: (1) multiple standard X-point divertors; (2) divertors with higher order nulls; (3) divertors with multiple X-points; and (4) long poloidal leg divertors (and also with multiple X-points). As a result, this paper reviews experiments and modeling in the area of radiative detachment in the advanced magnetic divertor configurations.« less

In-home networks integrating high-capacity DMT data and DVB-T over large-core GI-POF.

PubMed

Beltrán, Marta; Shi, Yan; Okonkwo, Chigo; Llorente, Roberto; Tangdiongga, Eduward; Koonen, Ton

2012-12-31

The low-cost in-home distribution of full-standard digital TV jointly with high-bitrate data using 50 m long 1 mm core diameter graded-index plastic optical fiber (GI-POF) is proposed and experimentally demonstrated. Discrete multitone (DMT) modulation is demonstrated to provide an adaptive bitrate which can spectrally coexist with digital video broadcasting-terrestrial (DVB-T) signals in 470-862 MHz. A 3 Gb/s DMT signal and two DVB-T channels are generated, transmitted and received exhibiting excellent performance.

EBW H&CD Potential for Spherical Tokamaks

NASA Astrophysics Data System (ADS)

Urban, J.; Decker, J.; Peysson, Y.; Preinhaelter, J.; Shevchenko, V.; Taylor, G.; Vahala, L.; Vahala, G.

2011-12-01

Spherical tokamaks (STs), which feature relatively high neutron flux and good economy, operate generally in high-ß regimes, in which the usual EC O- and X- modes are cut-off. In this case, electron Bernstein waves (EBWs) seem to be the only option that can provide features similar to the EC waves—controllable localized heating and current drive (H&) that can be utilized for core plasma heating as well as for accurate plasma stabilization. We first derive an analytical expression for Gaussian beam OXB conversion efficiency. Then, an extensive numerical study of EBW H&CD performance in four typical ST plasmas (NSTX L- and H-mode, MAST Upgrade, NHTX) is performed. Coupled ray-tracing (AMR) and Fokker-Planck (LUKE) codes are employed to simulate EBWs of varying frequencies and launch conditions. Our results indicate that an efficient and universal EBW H&CD system is indeed viable. In particular, power can be deposited and current reasonably efficiently driven across the whole plasma radius. Such a system could be controlled by a suitably chosen launching antenna vertical position and would also be sufficiently robust.

Full wave simulations of helicon wave losses in the scrape-off-layer of the DIII-D tokamak

NASA Astrophysics Data System (ADS)

Lau, Cornwall; Jaeger, Fred; Berry, Lee; Bertelli, Nicola; Pinsker, Robert

2017-10-01

Helicon waves have been recently proposed as an off-axis current drive actuator for DIII-D. Previous modeling using the hot plasma, full wave code AORSA, has shown good agreement with the ray tracing code GENRAY for helicon wave propagation and absorption in the core plasma. AORSA, and a new, reduced finite-element-model show discrepancies between ray tracing and full wave occur in the scrape-off-layer (SOL), especially at high densities. The reduced model is much faster than AORSA, and reproduces most of the important features of the AORSA model. The reduced model also allows for larger parametric scans and for the easy use of arbitrary tokamak geometry. Results of the full wave codes, AORSA and COMSOL, will be shown for helicon wave losses in the SOL are shown for a large range of parameters, such as SOL density profiles, n||, radial and vertical locations of the antenna, and different tokamak vessel geometries. This work was supported by DE-AC05-00OR22725, DE-AC02-09CH11466, and DE-FC02-04ER54698.

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.

Charge Exchange Recombination Spectroscopy Based on Diagnostic Neutral Beam in HT-7 Tokamak

NASA Astrophysics Data System (ADS)

Shi, Yuejiang; Fu, Jia; Li, Yingying; William, Rowan; Huang, He; Wang, Fudi; Gao, Huixian; Huang, Juann; Zhou, Qian; Liu, Sheng; Zhang, Jian; Li, Jun; Xie, Yuanlai; Liu, Zhimin; Huang, Yiyun; Hu, Chundong; Wan, Baonian

2010-02-01

Charge exchange recombination spectroscopy (CXRS) based on a diagnostic neutral beam (DNB) installed in the HT-7 tokamak is introduced. DNB can provide a 6 A extracted current at 50 kV for 0.1 s in hydrogen. It can penetrate into the core plasma in HT-7. The CXRS system is designed to observe charge exchange (CX) transitions in the visible spectrum. CX light from the beam is focused onto 10 optical fibers, which view the plasma from -5 cm to 20 cm. The CXRS system can measure the ion temperature as low as 0.1 keV. With CXRS, the local ion temperature profile in HT-7 was obtained for the first time.

A Distributed Synchronization and Timing System on the EAST Tokamak

NASA Astrophysics Data System (ADS)

Luo, Jiarong; Wu, Yichun; Shu, Yantai

2008-08-01

A key requirement for the EAST distributed control system (EASTDCS) is time synchronization to an accuracy of <1 mus. In 2006 a Distributed Synchronization and Timing System (DSTS) was set up, which is based on the ATmega128 AVR microcontroller and the Nut/OS embedded Real Time Operating System (RTOS). The DSTS provides the control and the data acquisition systems with reference clocks (0.01 Hz 10 MHz) and delayed trigger times ( 1 mus 4294 s). These are produced by a Core Module Unit (CMU) connected by optical fibres to many Local Synchronized Node Units (LSNU). The fibres provide immunity from electrical noise and are of equal length to match clock and trigger delays between systems. This paper describes the architecture of the DSTS on the EAST tokamak and provides an overview of the characteristics of the main and local units.

Polarized fusion, its implications, and plans for a proof-of-principle experiment at the DIII-D tokamak

NASA Astrophysics Data System (ADS)

Sandorfi, A. M.; Deur, A.; Lowry, M. M.; Wei, X.; Pace, D.; Eidietis, N.; Hyatt, A.; Jackson, G. L.; Lanctot, M.; Smith, S.; St-John, H.; Miller, G. W.; Zheng, X.; Baylor, L. R.

2015-10-01

The cross section for the primary fusion reaction in a tokamak, D+t --> α +n, would increase by a factor of 1.5 if the fuels were spin polarized parallel to the local field, rather than randomly oriented. Simulations show further gains in reaction rate would accompany this increase in large-scale machines such as ITER, due to increased alpha heating. The potential realization of such benefits rests on the crucial question of the survival of spin polarization for periods comparable to the energy containment time. Despite encouraging calculations, technical challenges in preparing and handling polarized materials have prevented any direct tests. Advances in three areas - polarized material technologies developed for nuclear and particle physics as well as medical imaging, polymer pellets developed for Inertial Confinement, and cryogenic injection guns developed for fueling tokamaks - have matured to the point where a direct in situ measurement is possible using the mirror reaction, D+3He --> α +p. Designs and simulations of a proof-of-principle experiment at the DIII-D tokamak in San Diego will be discussed. Work carried out under US DOE Contract DE-AC05-06OR23177 supporting Jefferson Lab and General Atomics Internal R&D funding.

IODP Expedition 340T: Borehole Logging at Atlantis Massif Oceanic Core Complex

NASA Astrophysics Data System (ADS)

Blackman, D.; Slagle, A.; Harding, A.; Guerin, G.; McCaig, A.

2013-03-01

Integrated Ocean Drilling Program (IODP) Expedition 340T returned to the 1.4-km-deep Hole U1309D at Atlantis Massif to carry out borehole logging including vertical seismic profiling (VSP). Seismic, resistivity, and temperature logs were obtained throughout the geologic section in the footwall of this oceanic core complex. Reliable downhole temperature measurements throughout and the first seismic coverage of the 800-1400 meters below seafloor (mbsf) portion of the section were obtained. Distinct changes in velocity, resistivity, and magnetic susceptibility characterize the boundaries of altered, olivine-rich troctolite intervals within the otherwise dominantly gabbroic se-quence. Some narrow fault zones also are associated with downhole resistivity or velocity excursions. Small deviations in temperature were measured in borehole fluid adjacent to known faults at 750 mbsf and 1100 mbsf. This suggests that flow of seawater remains active along these zones of faulting and rock alteration. Vertical seismic profile station coverage at zero offset now extends the full length of the hole, including the uppermost 150 mbsf, where detachment processes are expected to have left their strongest imprint. Analysis of wallrock properties, together with alteration and structural characteristics of the cores from Site U1309, highlights the likely interplay between lithology, structure, lithospheric hydration, and core complex evolution. doi:10.2204/iodp.sd.15.04.2013

FLiT: a field line trace code for magnetic confinement devices

NASA Astrophysics Data System (ADS)

Innocente, P.; Lorenzini, R.; Terranova, D.; Zanca, P.

2017-04-01

This paper presents a field line tracing code (FLiT) developed to study particle and energy transport as well as other phenomena related to magnetic topology in reversed-field pinch (RFP) and tokamak experiments. The code computes magnetic field lines in toroidal geometry using curvilinear coordinates (r, ϑ, ϕ) and calculates the intersections of these field lines with specified planes. The code also computes the magnetic and thermal diffusivity due to stochastic magnetic field in the collisionless limit. Compared to Hamiltonian codes, there are no constraints on the magnetic field functional formulation, which allows the integration of whichever magnetic field is required. The code uses the magnetic field computed by solving the zeroth-order axisymmetric equilibrium and the Newcomb equation for the first-order helical perturbation matching the edge magnetic field measurements in toroidal geometry. Two algorithms are developed to integrate the field lines: one is a dedicated implementation of a first-order semi-implicit volume-preserving integration method, and the other is based on the Adams-Moulton predictor-corrector method. As expected, the volume-preserving algorithm is accurate in conserving divergence, but slow because the low integration order requires small amplitude steps. The second algorithm proves to be quite fast and it is able to integrate the field lines in many partially and fully stochastic configurations accurately. The code has already been used to study the core and edge magnetic topology of the RFX-mod device in both the reversed-field pinch and tokamak magnetic configurations.

iPSC-Derived MSCs that Are Genetically Engineered for Systemic Bone Augmentation

DTIC Science & Technology

2013-08-01

cloned into a pJET1.2 vector (Fermentas, Glen Burnie, MD) and sequenced by MCLAB (San Francisco, CA). Karyotyping and G-banding. GTG -banding chromosome...publication [25]. Karyotyping and G-banding Giemsa ( GTG )-banding chromosome analysis was carried out in the LLU Radiation Research Laboratories. Standard...banding GTG -banding chromosome analysis was carried out in the LLU Radiation Research Laboratories. Standard DNA spectral karyo- typing procedures

Multi-frequency ICRF diagnostic of Tokamak plasmas

NASA Astrophysics Data System (ADS)

Lafonteese, David James

This thesis explores the diagnostic possibilities of a fast wave-based method for measuring the ion density and temperature profiles of tokamak plasmas. In these studies fast waves are coupled to the plasma at frequencies at the second harmonic of the ion gyrofrequency, at which wave energy is absorbed by the finite-temperature ions. As the ion gyrofrequency is dependent upon the local magnetic field, which varies as l/R in a tokamak, this power absorption is radially localized. The simultaneous launching of multiple frequencies, all resonating at different plasma positions, allows local measurements of the ion density and temperature. To investigate the profile applications of wave damping measurements in a simulated tokamak, an inhouse slab-model ICRF code is developed. A variety of analysis methods are presented, and ion density and temperature profiles are reconstructed for hydrogen plasmas for the Electric Tokamak (ET) and ITER parameter spaces. These methods achieve promising results in simulated plasmas featuring bulk ion heating, off-axis RF heating, and density ramps. The experimental results of similar studies on the Electric Tokamak, a high aspect ratio (R/a = 5), low toroidal field (2.2 kG) device are then presented. In these studies, six fast wave frequencies were coupled using a single-strap, low-field-side antenna to ET plasmas. The frequencies were variable, and could be tuned to resonate at different radii for different experiments. Four magnetic pickup loops were used to measure of the toroidal component of the wave magnetic field. The expected greater eigenmode damping of center-resonant frequencies versus edge-resonant frequencies is consistently observed. Comparison of measured aspects of fast wave behavior in ET is made with the slab code predictions, which validate the code simulations under weakly-damped conditions. A density profile is measured for an ET discharge through analysis of the fast wave measurements, and is compared to an

Protection of tokamak plasma facing components by a capillary porous system with lithium

NASA Astrophysics Data System (ADS)

Lyublinski, I.; Vertkov, A.; Mirnov, S.; Lazarev, V.

2015-08-01

Development of plasma facing material (PFM) based on the Capillary-Porous System (CPS) with lithium and activity on realization of lithium application strategy are addressed to meet the challenges under the creation of steady-state tokamak fusion reactor and fusion neutron source. Presented overview of experimental study of lithium CPS in plasma devices demonstrates the progress in protection of tokamak plasma facing components (PFC) from damage, stabilization and self-renewal of liquid lithium surface, elimination of plasma pollution and lithium accumulation in tokamak chamber. The possibility of PFC protection from the high power load related to cooling of the tokamak boundary plasma by radiation of non-fully stripped lithium ions supported by experimental results. This approach demonstrated in scheme of closed loops of Li circulation in the tokamak vacuum chamber and realized in a series of design of tokamak in-vessel elements.

Deuterium-tritium experiments on the Tokamak Fusion Test reactor

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

Hosea, J.; Adler, J.H.; Alling, P.

The deuterium-tritium (D-T) experimental program on the Tokamak Fusion Test Reactor (TFTR) is underway and routine tritium operations have been established. The technology upgrades made to the TFTR facility have been demonstrated to be sufficient for supporting both operations and maintenance for an extended D-T campaign. To date fusion power has been increased to {approx}9 MW and several physics results of importance to the D-T reactor regime have been obtained: electron temperature, ion temperature, and plasma stored energy all increase substantially in the D-T regime relative to the D-D regime at the same neutral beam power and comparable limiter conditioning;more » possible alpha electron heating is indicated and energy confinement improvement with average ion mass is observed; and alpha particle losses appear to be classical with no evidence of TAE mode activity up to the PFUS {approx}6 MW level. Instability in the TAE mode frequency range has been observed at PFUS > 7 MW and its effect on performance in under investigation. Preparations are underway to enhance the alpha particle density further by increasing fusion power and by extending the neutral beam pulse length to permit alpha particle effects of relevance to the ITER regime to be more fully explored.« less

A symplectic map for trajectories of magnetic field lines in double-null divertor tokamaks

NASA Astrophysics Data System (ADS)

Crank, Willie; Ali, Halima; Punjabi, Alkesh

2009-11-01

The coordinates of the area-preserving map equations for integration of magnetic field line trajectories in tokamaks can be any coordinates for which a transformation to (ψ,θ,φ) coordinates exists [A. Punjabi, H. Ali, T. Evans, and A. Boozer, Phys. Lett. A 364, 140 (2007)]. ψ is toroidal magnetic flux, θ is poloidal angle, and φ is toroidal angle. This freedom is exploited to construct a map that represents the magnetic topology of double-null divertor tokamaks. For this purpose, the generating function of the simple map [A. Punjabi, A. Verma, and A. Boozer, Phys. Rev. Lett. 69, 3322 (1992)] is slightly modified. The resulting map equations for the double-null divertor tokamaks are: x1=x0-ky0(1-y0^2 ), y1=y0+kx1. k is the map parameter. It represents the generic topological effects of toroidal asymmetries. The O-point is at (0.0). The X-points are at (0,±1). The equilibrium magnetic surfaces are calculated. These surfaces are symmetric about the x- and y- axes. The widths of stochastic layer near the X-points in the principal plane, and the fractal dimensions of the magnetic footprints on the inboard and outboard side of upper and lower X-points are calculated from the map. This work is supported by US Department of Energy grants DE-FG02-07ER54937, DE-FG02-01ER54624 and DE-FG02-04ER54793.

The fast reciprocating magnetic probe system on the J-TEXT tokamak

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

Li, Fuming; Chen, Zhipeng, E-mail: zpchen@hust.edu.cn; Zhuang, Ge

The fast reciprocating magnetic probe (FRMP) system is newly developed on the Joint Texas Experimental Tokamak (J-TEXT) to measure the local magnetic fluctuations at the plasma edge. The magnetic probe array in the FRMP consists of four 2-dimensional magnetic probes arranged at different radial locations to detect local poloidal and radial magnetic fields. These probes are protected by a graphite and boron nitride casing to improve the frequency response of each probe; they are mounted on the head of a movable rod, which is oriented along radial direction at the top of the torus. In the experiments, multiple core diagnosticsmore » show that the insertion of the FRMP has little impact on the equilibrium of the plasma. Local magnetic fluctuations inside the last closed flux surface are successfully measured by the FRMP.« less

Engineering aspects of the HT-6M Tokamak

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

Not Available

1986-05-01

The HT-6M is a medium-sized tokamak being built in China. The principal aim of the project is to study high-power auxiliary heating (1-MW neutral beam injection, 1-MW ion cyclotron resonance heating, and 100-kW electron cyclotron resonance heating), high-..beta.. experiments, the transport process, and the formation and diffusion process of impurities. The main device parameters are: major plasma radius R = 65 cm, minor plasma radius a = 20 cm, plasma current I/subP/ = 150 kA, discharge time tau = 150 ms, toroidal field B/subT/ = 15 kG. Simplicity of construction, accessibility to the plasma, reliability in operation, and convenience formore » maintenance were particularly emphasized in the design. The important design features of the device and power supply system are described.« less

From pure fusion to fusion-fission Demo tokamaks

NASA Astrophysics Data System (ADS)

Mirnov, S. V.

2013-04-01

The major requirements for pure fusion tokamak reactors and tokamak-based fusion neutron sources (FNS) are analyzed together with possible paths from the present-day tokamak towards the FNS tokamak. The FNS are of interest for traditional fission reactors as a method of waste management by burning of long-lived transuranic radionuclides (minorities) and fission fuel breeding. The Russian fission community places several hard requirements on the quality of FNS suitable for the first step of the investigation program of minority burning and breeding. They are (a) a steady-state regime of neutron production (more than 80% of the operational time), (b) a neutron power flux density greater than >0.2 MW m-2, (c) a total surface integrated neutron power >10 MW. Among the different FNS projects, based on magnetically confined plasmas, only ‘classical tokamak’ is most likely to fulfill these requirements in the nearest future. Some of the most important improvements of the ‘classical tokamak’ needed for successful realization of the FNS are (1) decrease in Zeff (probably, by making use of lithium as a part of plasma-facing components), (2) He removal and closed loop DT fuel circulation, (3) increase in the energy of stationary injected neutral tritium beams up to 150-170 keV and (4) control of impurity contamination at the plasma center (probably, by local RF heating). These key issues are discussed.

The symmetric quartic map for trajectories of magnetic field lines in elongated divertor tokamak plasmas

NASA Astrophysics Data System (ADS)

Jones, Morgin; Wadi, Hasina; Ali, Halima; Punjabi, Alkesh

2009-04-01

The coordinates of the area-preserving map equations for integration of magnetic field line trajectories in divertor tokamaks can be any coordinates for which a transformation to (ψt,θ,φ) coordinates exists [A. Punjabi, H. Ali, T. Evans, and A. Boozer, Phys. Lett. A 364, 140 (2007)]. ψt is toroidal magnetic flux, θ is poloidal angle, and φ is toroidal angle. This freedom is exploited to construct the symmetric quartic map such that the only parameter that determines magnetic geometry is the elongation of the separatrix surface. The poloidal flux inside the separatrix, the safety factor as a function of normalized minor radius, and the magnetic perturbation from the symplectic discretization are all held constant, and only the elongation is κ varied. The width of stochastic layer, the area, and the fractal dimension of the magnetic footprint and the average radial diffusion coefficient of magnetic field lines from the stochastic layer; and how these quantities scale with κ is calculated. The symmetric quartic map gives the correct scalings which are consistent with the scalings of coordinates with κ. The effects of m =1, n =±1 internal perturbation with the amplitude that is expected to occur in tokamaks are calculated by adding a term [H. Ali, A. Punjabi, A. H. Boozer, and T. Evans, Phys. Plasmas 11, 1908 (2004)] to the symmetric quartic map. In this case, the width of stochastic layer scales as 0.35 power of κ. The area of the footprint is roughly constant. The average radial diffusion coefficient of field lines near the X-point scales linearly with κ. The low mn perturbation changes the quasisymmetric structure of the footprint, and reorganizes it into a single, large scale, asymmetric structure. The symmetric quartic map is combined with the dipole map [A. Punjabi, H. Ali, and A. H. Boozer, Phys. Plasmas 10, 3992 (2003)] to calculate the effects of magnetic perturbation from a current carrying coil. The coil position and coil current coil are

Glycosyltransferase Function in Core 2-Type Protein O Glycosylation▿

PubMed Central

Stone, Erica L.; Ismail, Mohd Nazri; Lee, Seung Ho; Luu, Ying; Ramirez, Kevin; Haslam, Stuart M.; Ho, Samuel B.; Dell, Anne; Fukuda, Minoru; Marth, Jamey D.

2009-01-01

Three glycosyltransferases have been identified in mammals that can initiate core 2 protein O glycosylation. Core 2 O-glycans are abundant among glycoproteins but, to date, few functions for these structures have been identified. To investigate the biological roles of core 2 O-glycans, we produced and characterized mice deficient in one or more of the three known glycosyltransferases that generate core 2 O-glycans (C2GnT1, C2GnT2, and C2GnT3). A role for C2GnT1 in selectin ligand formation has been described. We now report that C2GnT2 deficiency impaired the mucosal barrier and increased susceptibility to colitis. C2GnT2 deficiency also reduced immunoglobulin abundance and resulted in the loss of all core 4 O-glycan biosynthetic activity. In contrast, the absence of C2GnT3 altered behavior linked to reduced thyroxine levels in circulation. Remarkably, elimination of all three C2GnTs was permissive of viability and fertility. Core 2 O-glycan structures were reduced among tissues from individual C2GnT deficiencies and completely absent from triply deficient mice. C2GnT deficiency also induced alterations in I-branching, core 1 O-glycan formation, and O mannosylation. Although the absence of C2GnT and C4GnT activities is tolerable in vivo, core 2 O glycosylation exerts a significant influence on O-glycan biosynthesis and is important in multiple physiological processes. PMID:19349303

Non-Solenoidal Tokamak Startup via Inboard Local Helicity Injection on the Pegasus ST

NASA Astrophysics Data System (ADS)

Perry, J. M.; Barr, J. L.; Bodner, G. M.; Bongard, M. W.; Fonck, R. J.; Pachicano, J. L.; Reusch, J. A.; Rodriguez Sanchez, C.; Richner, N. J.; Schlossberg, D. J.

2016-10-01

Local helicity injection (LHI) is a non-solenoidal startup technique utilizing small injectors at the plasma edge to source current along helical magnetic field lines. Unstable injected current streams relax to a tokamak-like configuration with high toroidal current multiplication. Flexible placement of injectors permits tradeoffs between helicity injection rate, poloidal field induction, and magnetic geometry requirements for initial relaxation. Experiments using a new set of large-area injectors in the lower divertor explore the efficacy of high-field-side (HFS) injection. The increased area (4 cm2) current source is functional up to full Pegasus toroidal field (BT , inj = 0.23 T). However, relaxation to a tokamak state is increasingly frustrated for BT , inj > 0.15 T with uniform vacuum vertical field. Paths to relaxation at increased field include: manipulation of vacuum poloidal field geometry; increased injector current; and plasma initiation with outboard injectors, subsequently transitioning to divertor injector drive. During initial tests of HFS injectors, achieved Vinj was limited to 600 V by plasma-material interactions on the divertor plate, which may be mitigated by increasing injector elevation. In experiments with helicity injection as the dominant current drive Ip 0.13 MA has been attained, with T̲e > 100 eV and ne 1019 m-3. Extrapolation to full BT, longer pulse length, and Vinj 1 kV suggest Ip > 0.25 MA should be attainable in a plasma dominated by helicity drive. Work supported by US DOE Grant DE-FG02-96ER54375.

Source-to-incident-flux relation in a Tokamak blanket module

NASA Astrophysics Data System (ADS)

Imel, G. R.

The next-generation Tokamak experiments, including the Tokamak fusion test reactor (TFTR), will utilize small blanket modules to measure performance parameters such as tritium breeding profiles, power deposition profiles, and neutron flux profiles. Specifically, a neutron calorimeter (simply a neutron moderating blanket module) which permits inferring the incident 14 MeV flux based on measured temperature profiles was proposed for TFTR. The problem of how to relate this total scalar flux to the fusion neutron source is addressed. This relation is necessary since the calorimeter is proposed as a total fusion energy monitor. The methods and assumptions presented was valid for the TFTR Lithium Breeding Module (LBM), as well as other modules on larger Tokamak reactors.

Plasma production and preliminary results from the ADITYA Upgrade tokamak

NASA Astrophysics Data System (ADS)

R, L. TANNA; J, GHOSH; Harshita, RAJ; Rohit, KUMAR; Suman, AICH; Vaibhav, RANJAN; K, A. JADEJA; K, M. PATEL; S, B. BHATT; K, SATHYANARAYANA; P, K. CHATTOPADHYAY; M, N. MAKWANA; K, S. SHAH; C, N. GUPTA; V, K. PANCHAL; Praveenlal, EDAPPALA; Bharat, ARAMBHADIYA; Minsha, SHAH; Vismay, RAULJI; M, B. CHOWDHURI; S, BANERJEE; R, MANCHANDA; D, RAJU; P, K. ATREY; Umesh, NAGORA; J, RAVAL; Y, S. JOISA; K, TAHILIANI; S, K. JHA; M, V. GOPALKRISHANA

2018-07-01

The Ohmically heated circular limiter tokamak ADITYA (R 0 = 75 cm, a = 25 cm) has been upgraded to a tokamak named the ADITYA Upgrade (ADITYA-U) with an open divertor configuration with divertor plates. The main goal of ADITYA-U is to carry out dedicated experiments relevant for bigger fusion machines including ITER, such as the generation and control of runaway electrons, disruption prediction, and mitigation studies, along with an improvement in confinement with shaped plasma. The ADITYA tokamak was dismantled and the assembly of ADITYA-U was completed in March 2016. Integration of subsystems like data acquisition and remote operation along with plasma production and preliminary plasma characterization of ADITYA-U plasmas are presented in this paper.

Improving core surgical training in a major trauma centre.

PubMed

Morris, Daniel L J; Bryson, David J; Ollivere, Ben J; Forward, Daren P

2016-06-01

English Major Trauma Centres (MTCs) were established in April 2012. Increased case volume and complexity has influenced trauma and orthopaedic (T&O) core surgical training in these centres. To determine if T&O core surgical training in MTCs meets Joint Committee on Surgical Training (JCST) quality indicators including performance of T&O operative procedures and consultant supervised session attendance. An audit cycle assessing the impact of a weekly departmental core surgical trainee rota. The rota included allocated timetabled sessions that optimised clinical and surgical learning opportunities. Intercollegiate Surgical Curriculum Programme (ISCP) records for T&O core surgical trainees at a single MTC were analysed for 8 months pre and post rota introduction. Outcome measures were electronic surgical logbook evidence of leading T&O operative procedures and consultant validated work-based assessments (WBAs). Nine core surgical trainees completed a 4 month MTC placement pre and post introduction of the core surgical trainee rota. Introduction of core surgical trainee rota significantly increased the mean number of T&O operative procedures led by a core surgical trainee during a 4 month MTC placement from 20.2 to 34.0 (p<0.05). The mean number of hip hemiarthroplasty procedures led by a core surgical trainee during a 4 month MTC placement was significantly increased (0.3 vs 2.4 [p=0.04]). Those of dynamic hip screw fixation (2.3 vs 3.6) and ankle fracture fixation (0.7 vs 1.6) were not. Introduction of a core surgical trainee rota significantly increased the mean number of consultant validated WBAs completed by a core surgical trainee during a 4 month MTC placement from 1.7 to 6.6 (p<0.0001). Introduction of a departmental core surgical trainee rota utilising a 'problem-based' model can significantly improve T&O core surgical training in MTCs. Copyright © 2016 Elsevier Ltd. All rights reserved.

Impedance of an intense plasma-cathode electron source for tokamak startup

NASA Astrophysics Data System (ADS)

Hinson, E. T.; Barr, J. L.; Bongard, M. W.; Burke, M. G.; Fonck, R. J.; Perry, J. M.

2016-05-01

An impedance model is formulated and tested for the ˜1 kV , 1 kA/cm2 , arc-plasma cathode electron source used for local helicity injection tokamak startup. A double layer sheath is established between the high-density arc plasma ( narc≈1021 m-3 ) within the electron source, and the less dense external tokamak edge plasma ( nedge≈1018 m-3 ) into which current is injected at the applied injector voltage, Vinj . Experiments on the Pegasus spherical tokamak show that the injected current, Iinj , increases with Vinj according to the standard double layer scaling Iinj˜Vinj3 /2 at low current and transitions to Iinj˜Vinj1 /2 at high currents. In this high current regime, sheath expansion and/or space charge neutralization impose limits on the beam density nb˜Iinj/Vinj1 /2 . For low tokamak edge density nedge and high Iinj , the inferred beam density nb is consistent with the requirement nb≤nedge imposed by space-charge neutralization of the beam in the tokamak edge plasma. At sufficient edge density, nb˜narc is observed, consistent with a limit to nb imposed by expansion of the double layer sheath. These results suggest that narc is a viable control actuator for the source impedance.

Developing physics basis for the snowflake divertor in the DIII-D tokamak

NASA Astrophysics Data System (ADS)

Soukhanovskii, V. A.; Allen, S. L.; Fenstermacher, M. E.; Lasnier, C. J.; Makowski, M. A.; McLean, A. G.; Meyer, W. H.; Ryutov, D. D.; Kolemen, E.; Groebner, R. J.; Hyatt, A. W.; Leonard, A. W.; Osborne, T. H.; Petrie, T. W.; Watkins, J.

2018-03-01

Recent DIII-D results demonstrate that the snowflake (SF) divertor geometry (see standard divertor) enables significant manipulation of divertor heat transport for heat spreading and reduction in attached and radiative divertor regimes, between and during edge localized modes (ELMs), while maintaining good H-mode confinement. Snowflake divertor configurations have been realized in the DIII-D tokamak for several seconds in H-mode discharges with heating power P_NBI ≤slant 4 -5 MW and a range of plasma currents I_p=0.8-1.2 MA. In this work, inter-ELM transport and radiative SF divertor properties are studied. Significant impact of geometric properties on SOL and divertor plasma parameters, including increased poloidal magnetic flux expansion, divertor magnetic field line length and divertor volume, is confirmed. In the SF-minus configuration, heat deposition is affected by the geometry, and peak divertor heat fluxes are significantly reduced. In the SF-plus and near-exact SF configurations, divertor peak heat flux reduction and outer strike point heat flux profile broadening are observed. Inter-ELM sharing of power and particle fluxes between the main and additional snowflake divertor strike points has been demonstrated. The additional strike points typically receive up to 10-15% of total outer divertor power. Measurements of electron pressure and poloidal beta βp support the theoretically proposed churning mode that is driven by toroidal curvature and vertical pressure gradient in the weak poloidal field region. A comparison of the 4-4.5 MW NBI-heated H-mode plasmas with radiative SF divertor and the standard radiative divertor (both induced with additional gas puffing) shows a nearly complete power detachment and broader divertor radiated power distribution in the SF, as compared to a partial detachment and peaked localized radiation in the standard divertor. However, insignificant difference in the detachment onset w.r.t. density between the SF and the standard

The Dynamic Mutation Characteristics of Thermonuclear Reaction in Tokamak

PubMed Central

Li, Jing; Quan, Tingting; Zhang, Wei; Deng, Wei

2014-01-01

The stability and bifurcations of multiple limit cycles for the physical model of thermonuclear reaction in Tokamak are investigated in this paper. The one-dimensional Ginzburg-Landau type perturbed diffusion equations for the density of the plasma and the radial electric field near the plasma edge in Tokamak are established. First, the equations are transformed to the average equations with the method of multiple scales and the average equations turn to be a Z 2-symmetric perturbed polynomial Hamiltonian system of degree 5. Then, with the bifurcations theory and method of detection function, the qualitative behavior of the unperturbed system and the number of the limit cycles of the perturbed system for certain groups of parameter are analyzed. At last, the stability of the limit cycles is studied and the physical meaning of Tokamak equations under these parameter groups is given. PMID:24892099

Solenoid-free plasma start-up in spherical tokamaks

NASA Astrophysics Data System (ADS)

Raman, R.; Shevchenko, V. F.

2014-10-01

The central solenoid is an intrinsic part of all present-day tokamaks and most spherical tokamaks. The spherical torus (ST) confinement concept is projected to operate at high toroidal beta and at a high fraction of the non-inductive bootstrap current as required for an efficient reactor system. The use of a conventional solenoid in a ST-based fusion nuclear facility is generally believed to not be a possibility. Solenoid-free plasma start-up is therefore an area of extensive worldwide research activity. Solenoid-free plasma start-up is also relevant to steady-state tokamak operation, as the central transformer coil of a conventional aspect ratio tokamak reactor would be located in a high radiation environment but would be needed only during the initial discharge initiation and current ramp-up phases. Solenoid-free operation also provides greater flexibility in the selection of the aspect ratio and simplifies the reactor design. Plasma start-up methods based on induction from external poloidal field coils, helicity injection and radio frequency current drive have all made substantial progress towards meeting this important need for the ST. Some of these systems will now undergo the final stages of test in a new generation of large STs, which are scheduled to begin operations during the next two years. This paper reviews research to date on methods for inducing the initial start-up current in STs without reliance on the conventional central solenoid.

Observation of finite-. beta. MHD phenomena in tokamaks

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

McGuire, K.M.

1984-09-01

Stable high-beta plasmas are required for the tokamak to attain an economical fusion reactor. Recently, intense neutral beam heating experiments in tokamaks have shown new effects on plasma stability and confinement associated with high beta plasmas. The observed spectrum of MHD fluctuations at high beta is clearly dominated by the n = 1 mode when the q = 1 surface is in the plasma. The m/n = 1/1 mode drives other n = 1 modes through toroidal coupling and n > 1 modes through nonlinear coupling. On PDX, with near perpendicular injection, a resonant interaction between the n = 1more » internal kink and the trapped fast ions results in loss of beam particles and heating power. Key parameters in the theory are the value of q/sub 0/ and the injection angle. High frequency broadband magnetic fluctuations have been observed on ISX-B and D-III and a correlation with the deterioration of plasma confinement was reported. During enhanced confinement (H-mode) discharges in divertor plasmas, two new edge instabilities were observed, both localized radially near the separatrix. By assembling results from the different tokamak experiments, it is found that the simple theoretical ideal MHD beta limit has not been exceeded. Whether this represents an ultimate tokamak limit or if beta optimized configurations (Dee- or bean-shaped plasmas) can exceed this limit and perhaps enter a second regime of stability remains to be clarified.« less

OMFIT Tokamak Profile Data Fitting and Physics Analysis

DOE PAGES

Logan, N. C.; Grierson, B. A.; Haskey, S. R.; ...

2018-01-22

Here, One Modeling Framework for Integrated Tasks (OMFIT) has been used to develop a consistent tool for interfacing with, mapping, visualizing, and fitting tokamak profile measurements. OMFIT is used to integrate the many diverse diagnostics on multiple tokamak devices into a regular data structure, consistently applying spatial and temporal treatments to each channel of data. Tokamak data are fundamentally time dependent and are treated so from the start, with front-loaded and logic-based manipulations such as filtering based on the identification of edge-localized modes (ELMs) that commonly scatter data. Fitting is general in its approach, and tailorable in its application inmore » order to address physics constraints and handle the multiple spatial and temporal scales involved. Although community standard one-dimensional fitting is supported, including scale length–fitting and fitting polynomial-exponential blends to capture the H-mode pedestal, OMFITprofiles includes two-dimensional (2-D) fitting using bivariate splines or radial basis functions. These 2-D fits produce regular evolutions in time, removing jitter that has historically been smoothed ad hoc in transport applications. Profiles interface directly with a wide variety of models within the OMFIT framework, providing the inputs for TRANSP, kinetic-EFIT 2-D equilibrium, and GPEC three-dimensional equilibrium calculations. he OMFITprofiles tool’s rapid and comprehensive analysis of dynamic plasma profiles thus provides the critical link between raw tokamak data and simulations necessary for physics understanding.« less

OMFIT Tokamak Profile Data Fitting and Physics Analysis

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

Logan, N. C.; Grierson, B. A.; Haskey, S. R.

Here, One Modeling Framework for Integrated Tasks (OMFIT) has been used to develop a consistent tool for interfacing with, mapping, visualizing, and fitting tokamak profile measurements. OMFIT is used to integrate the many diverse diagnostics on multiple tokamak devices into a regular data structure, consistently applying spatial and temporal treatments to each channel of data. Tokamak data are fundamentally time dependent and are treated so from the start, with front-loaded and logic-based manipulations such as filtering based on the identification of edge-localized modes (ELMs) that commonly scatter data. Fitting is general in its approach, and tailorable in its application inmore » order to address physics constraints and handle the multiple spatial and temporal scales involved. Although community standard one-dimensional fitting is supported, including scale length–fitting and fitting polynomial-exponential blends to capture the H-mode pedestal, OMFITprofiles includes two-dimensional (2-D) fitting using bivariate splines or radial basis functions. These 2-D fits produce regular evolutions in time, removing jitter that has historically been smoothed ad hoc in transport applications. Profiles interface directly with a wide variety of models within the OMFIT framework, providing the inputs for TRANSP, kinetic-EFIT 2-D equilibrium, and GPEC three-dimensional equilibrium calculations. he OMFITprofiles tool’s rapid and comprehensive analysis of dynamic plasma profiles thus provides the critical link between raw tokamak data and simulations necessary for physics understanding.« less

Properties of ion temperature gradient and trapped electron modes in tokamak plasmas with inverted density profiles

NASA Astrophysics Data System (ADS)

Du, Huarong; Jhang, Hogun; Hahm, T. S.; Dong, J. Q.; Wang, Z. X.

2017-12-01

We perform a numerical study of linear stability of the ion temperature gradient (ITG) mode and the trapped electron mode (TEM) in tokamak plasmas with inverted density profiles. A local gyrokinetic integral equation is applied for this study. From comprehensive parametric scans, we obtain stability diagrams for ITG modes and TEMs in terms of density and temperature gradient scale lengths. The results show that, for the inverted density profile, there exists a normalized threshold temperature gradient above which the ITG mode and the TEM are either separately or simultaneously unstable. The instability threshold of the TEM for the inverted density profile is substantially different from that for normal and flat density profiles. In addition, deviations are found on the ITG threshold from an early analytic theory in sheared slab geometry with the adiabatic electron response [T. S. Hahm and W. M. Tang, Phys. Fluids B 1, 1185 (1989)]. A possible implication of this work on particle transport in pellet fueled tokamak plasmas is discussed.

A Higher Correlation of HCV Core Antigen with CD4+ T Cell Counts Compared with HCV RNA in HCV/HIV-1 Coinfected Patients

PubMed Central

Zhang, Weidong; Xi, Yuanlin; Cao, Guanghua; Zhi, Yuhong; Wang, Shuiwang; Xu, Chunhui; Wei, Lai; Lu, Fengmin; Zhuang, Hui

2011-01-01

Development of HCV infection is typically followed by chronic hepatitis C (CHC) in most patients, while spontaneous HCV viral clearance (SVC) occurs in only a minority of subjects. Compared with the widespread application of HCV RNA testing by quantitative RT-PCR technique, HCV core antigen detection may be an alternative indicator in the diagnosis of hepatitis C virus infections and in monitoring the status of infectious individuals. However, the correlation and differences between these two indicators in HCV infection need more investigation, especially in patients coinfected by HIV-1. In this study, a total of 354 anti-HCV and/or anti-HIV serum positive residents from a village of central China were enrolled. Besides HCV-related hepatopathic variables including clinical status, ALT, AST, anti-HCV Abs, as well as the altered CD4+/CD8+ T cell counts, HCV core antigen and HCV viral load were also measured. The concentration of serum HCV core antigen was highly correlated with level of HCV RNA in CHC patients with or without HIV-1 coinfection. Of note, HCV core antigen concentration was negatively correlated with CD4+ T cell count, while no correlation was found between HCV RNA level and CD4+ T cell count. Our findings suggested that quantitative detection of plasma HCV core antigen may be an alternative indicator of HCV RNA qPCR assay when evaluating the association between HCV replication and host immune status in HCV/HIV-1 coinfected patients. PMID:21858166

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.

Spherical tokamaks with plasma centre-post

NASA Astrophysics Data System (ADS)

Ribeiro, Celso

2013-10-01

The metal centre-post (MCP) in tokamaks is a structure which carries the total toroidal field current and also houses the Ohmic heating solenoid in conventional or low aspect ratio (Spherical)(ST) tokamaks. The MCP and solenoid are critical components for producing the toroidal field and for the limited Ohmic flux in STs. Constraints for a ST reactor related to these limitations lead to a minimum plasma aspect ratio of 1.4 which reduces the benefit of operation at higher betas in a more compact ST reactor. Replacing the MCP is of great interest for reactor-based ST studies since the device is simplified, compactness increased, and maintenance reduced. An experiment to show the feasibility of using a plasma centre-post (PCP) is being currently under construction and involves a high level of complexity. A preliminary study of a very simple PCP, which is ECR(Electron Cyclotron Resonance)-assisted and which includes an innovative fuelling system based on pellet injection, has recently been reported. This is highly suitable for an ultra-low aspect ratio tokamak (ULART) device. Advances on this PCP ECR-assisted concept within a ULART and the associated fuelling system are presented here, and will include the field topology for the PCP ECR-assisted scheme, pellet ablation modeling, and a possible global equilibrium simulation. VIE-ITCR, IAEA-CRP contr.17592, National Instruments-Costa Rica.

The effect of pressure anisotropy on ballooning modes in tokamak plasmas

NASA Astrophysics Data System (ADS)

Johnston, A.; Hole, M. J.; Qu, Z. S.; Hezaveh, H.

2018-06-01

Edge Localised Modes are thought to be caused by a spectrum of magnetohydrodynamic instabilities, including the ballooning mode. While ballooning modes have been studied extensively both theoretically and experimentally, the focus of the vast majority of this research has been on isotropic plasmas. The prevalence of pressure anisotropy in modern tokamaks thus motivates further study of these modes. This paper presents a numerical analysis of ballooning modes in anisotropic equilibria. The investigation was conducted using the newly-developed codes HELENA+ATF and MISHKA-A, which adds anisotropic physics to equilibria and stability analysis. We have examined the impact of anisotropy on the stability of an n = 30 ballooning mode, confirming results conform to previous calculations in the isotropic limit. Growth rates of ballooning modes in equilibria with different levels of anisotropy were then calculated using the stability code MISHKA-A. The key finding was that the level of anisotropy had a significant impact on ballooning mode growth rates. For {T}\\perp > {T}| | , typical of ICRH heating, the growth rate increases, while for {T}\\perp < {T}| | , typical of neutral beam heating, the growth rate decreases.

Geodesic acoustic modes in noncircular cross section tokamaks

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

Sorokina, E. A., E-mail: sorokina.ekaterina@gmail.com; Lakhin, V. P.; Konovaltseva, L. V.

2017-03-15

The influence of the shape of the plasma cross section on the continuous spectrum of geodesic acoustic modes (GAMs) in a tokamak is analyzed in the framework of the MHD model. An expression for the frequency of a local GAM for a model noncircular cross section plasma equilibrium is derived. Amendments to the oscillation frequency due to the plasma elongation and triangularity and finite tokamak aspect ratio are calculated. It is shown that the main factor affecting the GAM spectrum is the plasma elongation, resulting in a significant decrease in the mode frequency.

Spectral emission measurements of lithium on the lithium tokamak experiment.

PubMed

Gray, T K; Biewer, T M; Boyle, D P; Granstedt, E M; Kaita, R; Maingi, R; Majeski, R P

2012-10-01

There has been a long-standing collaboration between ORNL and PPPL on edge and boundary layer physics. As part of this collaboration, ORNL has a large role in the instrumentation and interpretation of edge physics in the lithium tokamak experiment (LTX). In particular, a charge exchange recombination spectroscopy (CHERS) diagnostic is being designed and undergoing staged testing on LTX. Here we present results of passively measured lithium emission at 5166.89 A in LTX in anticipation of active spectroscopy measurements, which will be enabled by the installation of a neutral beam in 2013. Preliminary measurements are made in transient LTX plasmas with plasma current, I(p) < 70 kA, ohmic heating power, P(oh) ∼ 0.3 MW and discharge lifetimes of 10-15 ms. Measurements are made with a short focal length spectrometer and optics similar to the CHERS diagnostics on NSTX [R. E. Bell, Rev. Sci. Instrum. 68(2), 1273-1280 (1997)]. These preliminary measurements suggest that even without the neutral beam for active spectroscopy, there is sufficient passive lithium emission to allow for line-of-sight profile measurements of ion temperature, T(i); toroidal velocity and v(t). Results show peak T(i) = 70 eV and peak v(t) = 45 km/s were reached 10 ms into the discharge.

Study of the L-mode tokamak plasma “shortfall” with local and global nonlinear gyrokinetic δf particle-in-cell simulation

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

Chowdhury, J.; Wan, Weigang; Chen, Yang

2014-11-15

The δ f particle-in-cell code GEM is used to study the transport “shortfall” problem of gyrokinetic simulations. In local simulations, the GEM results confirm the previously reported simulation results of DIII-D [Holland et al., Phys. Plasmas 16, 052301 (2009)] and Alcator C-Mod [Howard et al., Nucl. Fusion 53, 123011 (2013)] tokamaks with the continuum code GYRO. Namely, for DIII-D the simulations closely predict the ion heat flux at the core, while substantially underpredict transport towards the edge; while for Alcator C-Mod, the simulations show agreement with the experimental values of ion heat flux, at least within the range of experimental error.more » Global simulations are carried out for DIII-D L-mode plasmas to study the effect of edge turbulence on the outer core ion heat transport. The edge turbulence enhances the outer core ion heat transport through turbulence spreading. However, this edge turbulence spreading effect is not enough to explain the transport underprediction.« less

The ARIES Advanced and Conservative Tokamak Power Plant Study

DOE PAGES

Kessel, C. E; Tillak, M. S; Najmabadi, F.; ...

2015-12-22

Tokamak power plants are studied with advanced and conservative design philosophies to identify the impacts on the resulting designs and to provide guidance to critical research needs. Incorporating updated physics understanding and using more sophisticated engineering and physics analysis, the tokamak configurations have developed a more credible basis compared with older studies. The advanced configuration assumes a self-cooled lead lithium blanket concept with SiC composite structural material with 58% thermal conversion efficiency. This plasma has a major radius of 6.25 m, a toroidal field of 6.0 T, a q₉₅ of 4.5, aᵦ total N of 5.75, an H98 of 1.65,more » an n/n Gr of 1.0, and a peak divertor heat flux of 13.7 MW/m² . The conservative configuration assumes a dual-coolant lead lithium blanket concept with reduced activation ferritic martensitic steel structural material and helium coolant, achieving a thermal conversion efficiency of 45%. The plasma has a major radius of 9.75 m, a toroidal field of 8.75 T, a q₉₅ of 8.0, aᵦ total N of 2.5, an H₉₈ of 1.25, an n/n Gr of 1.3, and a peak divertor heat flux of 10 MW/m² . The divertor heat flux treatment with a narrow power scrape off width has driven the plasmas to larger major radius. Edge and divertor plasma simulations are targeting a basis for high radiated power fraction in the divertor, which is necessary for solutions to keep the peak heat flux in the range 10 to 15 MW/m² . Combinations of the advanced and conservative approaches show intermediate sizes. A new systems code using a database approach has been used and shows that the operating point is really an operating zone with some range of plasma and engineering parameters and very similar costs of electricity. Other papers in this issue provide more detailed discussion of the work summarized here.« less

A mechanism for large divertor plasma energy loss via lithium radiation in tokamaks

NASA Astrophysics Data System (ADS)

Rognlien, T. D.; Meier, E. T.; Soukhanovskii, V. A.

2012-10-01

Lithium has been used as a wall-conditioning element in a number of tokamaks over the years, including TFTR, FTU, and NSTX, where core plasma energy confinement and particle control are often found to improve following such conditioning. Here the possible role of Li in providing substantial energy loss for divertor plasmas via line radiation is reported. A multi-charge-state 2D UEDGE fluid model is used where the hydrogenic and Li ions and neutrals are each evolved as separate species and separate equations are solved for the electron and ion temperatures. It is shown that a sufficient level of Li neutrals evolving from the divertor surface via sputtering or evaporation can induce energy detachment of the divertor plasma, yielding a strongly radiating zone near the divertor where ionization and recombination from/to neutral Li can radiate most of the power flowing into the scrape-off layer while maintaining low core contamination. A local peaking of Li emissivity for electron temperatures near 1 eV appears to play an important role in the detachment of the mixed deuterium/Li plasma. Evidence of such behavior from NSTX discharges will be discussed.

ECRH Studies on Tokamak Plasmas.

DTIC Science & Technology

1980-10-10

r.I*cru.Dtrtibution uUnliited 300 Unicorn Pork Drive Woburn, Massachusetts 04801 ECRH STUDIES ON TOKAMAK PLASMAS JAYCOR Project No. 6183 Final Report...up techniques now in use or being suggested, include growing the plasma from a small minor radius or applying a negative voltage spike immediately

Inferring Core Tungsten Behavior Using SPRED During the DIII-D Metal Rings Campaign

NASA Astrophysics Data System (ADS)

Thomas, D. M.; Kaplan, D.; Groebner, R.; Grierson, B.; Unterberg, Z.; Victor, B.

2016-10-01

The GA SPRED EUV spectrometer was used to study core emission of highly charged tungsten ions (W40+-W45+) in the 120-135Å region during the recent Metal Rings Campaign. These experiments used two 5-cm wide toroidal rings of W-coated metal inserts exposed to a variety of DIII-D discharges to study effects of high-Z divertor erosion, migration, core uptake, and effects on advanced tokamak performance. For the proper core temperature range (2-4 keV), the measured multistate W emission forms a well defined spectral pattern that can be used to study the relative importance of strike point location, flux expansion, injected power, ELM characteristics and magnetic drift direction for high-Z core contamination in DIII-D. The spectra are fit using simple Gaussians to estimate concentrations using the historical SPRED intensity calibration. Calibration shots using known core dosages of pellet injected W are used to help infer the relative response of the instrument. Supported by US DOE under DE-FC02-04ER54698, DE-AC02-09CH11466, DE-AC05-00OR22725, DE-AC52-07NA27344.

Compatibility of lithium plasma-facing surfaces with high edge temperatures in the Lithium Tokamak Experiment

NASA Astrophysics Data System (ADS)

Majeski, R.; Bell, R. E.; Boyle, D. P.; Kaita, R.; Kozub, T.; LeBlanc, B. P.; Lucia, M.; Maingi, R.; Merino, E.; Raitses, Y.; Schmitt, J. C.; Allain, J. P.; Bedoya, F.; Bialek, J.; Biewer, T. M.; Canik, J. M.; Buzi, L.; Koel, B. E.; Patino, M. I.; Capece, A. M.; Hansen, C.; Jarboe, T.; Kubota, S.; Peebles, W. A.; Tritz, K.

2017-05-01

High edge electron temperatures (200 eV or greater) have been measured at the wall-limited plasma boundary in the Lithium Tokamak Experiment (LTX). Flat electron temperature profiles are a long-predicted consequence of low recycling boundary conditions. Plasma density in the outer scrape-off layer is very low, 2-3 × 1017 m-3, consistent with a low recycling metallic lithium boundary. Despite the high edge temperature, the core impurity content is low. Zeff is estimated to be ˜1.2, with a very modest contribution (<0.1) from lithium. Experiments are transient. Gas puffing is used to increase the plasma density. After gas injection stops, the discharge density is allowed to drop, and the edge is pumped by the low recycling lithium wall. An upgrade to LTX-LTX-β, which includes a 35A, 20 kV neutral beam injector (on loan to LTX from Tri-Alpha Energy) to provide core fueling to maintain constant density, as well as auxiliary heating, is underway. LTX-β is briefly described.

Compatibility of lithium plasma-facing surfaces with high edge temperatures in the Lithium Tokamak Experiment

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

Majeski, R.; Bell, R. E.; Boyle, D. P.

We measured high edge electron temperatures (200 eV or greater) at the wall-limited plasma boundary in the Lithium Tokamak Experiment (LTX). Flat electron temperature profiles are a long-predicted consequence of low recycling boundary conditions. Plasma density in the outer scrape-off layer is very low, 2-3 x 10(17) m(-3), consistent with a low recycling metallic lithium boundary. In spite of the high edge temperature, the core impurity content is low. Z(eff) is estimated to be similar to 1.2, with a very modest contribution (< 0.1) from lithium. Experiments are transient. Gas puffing is used to increase the plasma density. After gasmore » injection stops, the discharge density is allowed to drop, and the edge is pumped by the low recycling lithium wall. An upgrade to LTX-LTX-beta, which includes a 35A, 20 kV neutral beam injector (on loan to LTX from Tri-Alpha Energy) to provide core fueling to maintain constant density, as well as auxiliary heating, is underway. LTX-beta is briefly described.« less

Compatibility of lithium plasma-facing surfaces with high edge temperatures in the Lithium Tokamak Experiment

DOE PAGES

Majeski, R.; Bell, R. E.; Boyle, D. P.; ...

2017-03-20

We measured high edge electron temperatures (200 eV or greater) at the wall-limited plasma boundary in the Lithium Tokamak Experiment (LTX). Flat electron temperature profiles are a long-predicted consequence of low recycling boundary conditions. Plasma density in the outer scrape-off layer is very low, 2-3 x 10(17) m(-3), consistent with a low recycling metallic lithium boundary. In spite of the high edge temperature, the core impurity content is low. Z(eff) is estimated to be similar to 1.2, with a very modest contribution (< 0.1) from lithium. Experiments are transient. Gas puffing is used to increase the plasma density. After gasmore » injection stops, the discharge density is allowed to drop, and the edge is pumped by the low recycling lithium wall. An upgrade to LTX-LTX-beta, which includes a 35A, 20 kV neutral beam injector (on loan to LTX from Tri-Alpha Energy) to provide core fueling to maintain constant density, as well as auxiliary heating, is underway. LTX-beta is briefly described.« less

Characterization of a high-temperature superconducting conductor on round core cables in magnetic fields up to 20 T

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

van der Laan, D. C.; Noyes, P. D.; Miller, G. E.

2013-02-13

The next generation of high-ï¬eld magnets that will operate at magnetic ï¬elds substantially above 20 T, or at temperatures substantially above 4.2 K, requires high-temperature superconductors (HTS). Conductor on round core (CORC) cables, in which RE-Ba{sub 2}Cu{sub 3}O{sub 7-{delta}} (RE = rare earth) (REBCO) coated conductors are wound in a helical fashion on a flexible core, are a practical and versatile HTS cable option for low-inductance, high-field magnets. We performed the first tests of CORC magnet cables in liquid helium in magnetic fields of up to 20 T. A record critical current I{sub c} of 5021 A was measured atmore » 4.2 K and 19 T. In a cable with an outer diameter of 7.5 mm, this value corresponds to an engineering current density J{sub e} of 114 A mm{sup -2} , the highest J{sub e} ever reported for a superconducting cable at such high magnetic fields. Additionally, the first magnet wound from an HTS cable was constructed from a 6 m-long CORC cable. The 12-turn, double-layer magnet had an inner diameter of 9 cm and was tested in a magnetic field of 20 T, at which it had an I{sub c} of 1966 A. The cables were quenched repetitively without degradation during the measurements, demonstrating the feasibility of HTS CORC cables for use in high-field magnet applications.« less

The symmetric quartic map for trajectories of magnetic field lines in elongated divertor tokamak plasmas

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

Jones, Morgin; Wadi, Hasina; Ali, Halima

The coordinates of the area-preserving map equations for integration of magnetic field line trajectories in divertor tokamaks can be any coordinates for which a transformation to ({psi}{sub t},{theta},{phi}) coordinates exists [A. Punjabi, H. Ali, T. Evans, and A. Boozer, Phys. Lett. A 364, 140 (2007)]. {psi}{sub t} is toroidal magnetic flux, {theta} is poloidal angle, and {phi} is toroidal angle. This freedom is exploited to construct the symmetric quartic map such that the only parameter that determines magnetic geometry is the elongation of the separatrix surface. The poloidal flux inside the separatrix, the safety factor as a function of normalizedmore » minor radius, and the magnetic perturbation from the symplectic discretization are all held constant, and only the elongation is {kappa} varied. The width of stochastic layer, the area, and the fractal dimension of the magnetic footprint and the average radial diffusion coefficient of magnetic field lines from the stochastic layer; and how these quantities scale with {kappa} is calculated. The symmetric quartic map gives the correct scalings which are consistent with the scalings of coordinates with {kappa}. The effects of m=1, n={+-}1 internal perturbation with the amplitude that is expected to occur in tokamaks are calculated by adding a term [H. Ali, A. Punjabi, A. H. Boozer, and T. Evans, Phys. Plasmas 11, 1908 (2004)] to the symmetric quartic map. In this case, the width of stochastic layer scales as 0.35 power of {kappa}. The area of the footprint is roughly constant. The average radial diffusion coefficient of field lines near the X-point scales linearly with {kappa}. The low mn perturbation changes the quasisymmetric structure of the footprint, and reorganizes it into a single, large scale, asymmetric structure. The symmetric quartic map is combined with the dipole map [A. Punjabi, H. Ali, and A. H. Boozer, Phys. Plasmas 10, 3992 (2003)] to calculate the effects of magnetic perturbation from a current

Single Null Negative Triangularity Tokamak for Power Handling

NASA Astrophysics Data System (ADS)

Kikuchi, Mitsuru; Medvedev, S.; Takizuka, T.; Sauter, O.; Merle, A.; Coda, S.; Chen, D.; Li, J. X.

2017-10-01

Power and particle control in fusion reactor is challenge and we proposed the negative triangularity tokamak (NTT) to eliminate ELM by operating L-mode edge with improved core confinement. The SN configuration has more flexibility in shaping by adopting rectangular-shaped TF coils. The limiting normalized beta is 3.56 with wall stabilization and 3.14 without wall. The vertical stability is assured under a reasonable control system. The wetted area on the divertor plates becomes wider in proportion to the larger major radius at the divertor strike points due to the NT configuration. In addition to the major-radius effect, the ``Flux Tune Expansion (FTE)'' is adopted to further reduce the heat load on the divertor plate by factor of 2.6 with a coil current 3 MA. L-mode edge also allows further increase in wetted area. The fusion power of 3 GW is deliverable only at normalized beta 2.1. Therefore this reactor may be operable stably against the serious MHD activities. The CD power for SS operation is 175 MW at Q = 17. AC operation is also possible option. A required HH factor is relatively modest H = 1.12.

Impedance of an intense plasma-cathode electron source for tokamak startup

DOE PAGES

Hinson, Edward Thomas; Barr, Jayson L.; Bongard, Michael W.; ...

2016-05-31

In this study, an impedance model is formulated and tested for the ~1kV, ~1kA/cm 2, arc-plasma cathode electron source used for local helicity injection tokamak startup. A double layer sheath is established between the high-density arc plasma (n arc ≈ 10 21 m -3) within the electron source, and the less dense external tokamak edge plasma (n edge ≈ 10 18 m -3) into which current is injected at the applied injector voltage, V inj. Experiments on the Pegasus spherical tokamak show the injected current, I inj, increases with V inj according to the standard double layer scaling I injmore » ~ V inj 3/2 at low current and transitions to I inj ~ V inj 1/2 at high currents. In this high current regime, sheath expansion and/or space charge neutralization impose limits on the beam density n b ~ I inj/V inj 1/2. For low tokamak edge density n edge and high I inj, the inferred beam density n b is consistent with the requirement n b ≤ n edge imposed by space-charge neutralization of the beam in the tokamak edge plasma. At sufficient edge density, n b ~ n arc is observed, consistent with a limit to n b imposed by expansion of the double layer sheath. These results suggest that n arc is a viable control actuator for the source impedance.« less

Identification of new turbulence contributions to plasma transport and confinement in spherical tokamak regime

NASA Astrophysics Data System (ADS)

Wang, W. X.; Ethier, S.; Ren, Y.; Kaye, S.; Chen, J.; Startsev, E.; Lu, Z.; Li, Z. Q.

2015-10-01

Highly distinct features of spherical tokamaks (ST), such as National Spherical Torus eXperiment (NSTX) and NSTX-U, result in a different fusion plasma regime with unique physics properties compared to conventional tokamaks. Nonlinear global gyrokinetic simulations critical for addressing turbulence and transport physics in the ST regime have led to new insights. The drift wave Kelvin-Helmholtz (KH) instability characterized by intrinsic mode asymmetry is identified in strongly rotating NSTX L-mode plasmas. While the strong E ×B shear associated with the rotation leads to a reduction in KH/ion temperature gradient turbulence, the remaining fluctuations can produce a significant ion thermal transport that is comparable to the experimental level in the outer core region (with no "transport shortfall"). The other new, important turbulence source identified in NSTX is the dissipative trapped electron mode (DTEM), which is believed to play little role in conventional tokamak regime. Due to the high fraction of trapped electrons, long wavelength DTEMs peaking around kθρs˜0.1 are destabilized in NSTX collisionality regime by electron density and temperature gradients achieved there. Surprisingly, the E ×B shear stabilization effect on DTEM is remarkably weak, which makes it a major turbulence source in the ST regime dominant over collisionless TEM (CTEM). The latter, on the other hand, is subject to strong collisional and E ×B shear suppression in NSTX. DTEM is shown to produce significant particle, energy and toroidal momentum transport, in agreement with experimental levels in NSTX H-modes. Moreover, DTEM-driven transport in NSTX parametric regime is found to increase with electron collision frequency, providing one possible source for the scaling of confinement time observed in NSTX H-modes. Most interestingly, the existence of a turbulence-free regime in the collision-induced CTEM to DTEM transition, corresponding to a minimum plasma transport in advanced ST

Identification of new turbulence contributions to plasma transport and confinement in spherical tokamak regime

DOE PAGES

Wang, W. X.; Ethier, S.; Ren, Y.; ...

2015-10-15

Highly distinct features of spherical tokamaks (ST), such as National Spherical Torus eXperiment (NSTX) and NSTX-U, result in a different fusion plasma regime with unique physics properties compared to conventional tokamaks. Nonlinear global gyrokinetic simulations critical for addressing turbulence and transport physics in the ST regime have led to new insights. The drift wave Kelvin-Helmholtz (KH) instability characterized by intrinsic mode asymmetry is identified in strongly rotating NSTX L-mode plasmas. While the strong E x B shear associated with the rotation leads to a reduction in KH/ion temperature gradient turbulence, the remaining fluctuations can produce a significant ion thermal transportmore » that is comparable to the experimental level in the outer core region (with no "transport shortfall"). The other new, important turbulence source identified in NSTX is the dissipative trapped electron mode (DTEM), which is believed to play little role in conventional tokamak regime. Due to the high fraction of trapped electrons, long wavelength DTEMs peaking around k θρs ~ 0.1 are destabilized in NSTX collisionality regime by electron density and temperature gradients achieved there. Surprisingly, the E x B shear stabilization effect on DTEM is remarkably weak, which makes it a major turbulence source in the ST regime dominant over collisionless TEM (CTEM). The latter, on the other hand, is subject to strong collisional and E x B shear suppression in NSTX. DTEM is shown to produce significant particle, energy and toroidal momentum transport, in agreement with experimental levels in NSTX H-modes. Furthermore, DTEM-driven transport in NSTX parametric regime is found to increase with electron collision frequency, providing one possible source for the scaling of confinement time observed in NSTX H-modes. Most interestingly, the existence of a turbulence-free regime in the collision-induced CTEM to DTEM transition, corresponding to a minimum plasma transport in

Interaction of external n = 1 magnetic fields with the sawtooth instability in low- q RFX-mod and DIII-D tokamaks

DOE PAGES

Piron, C.; Martin, P.; Bonfiglio, D.; ...

2016-08-11

External n = 1 magnetic fields are applied in RFX-mod and DIII-D low safety factor Tokamak plasmas to investigate their interaction with the internal MHD dynamics and in particular with the sawtooth instability. In these experiments the applied magnetic fields cause a reduction of both the sawtooth amplitude and period, leading to an overall stabilizing effect on the oscillations. In RFX-mod sawteeth eventually disappear and are replaced by a stationary m = 1, n = 1 helical equilibrium without an increase in disruptivity. However toroidal rotation is significantly reduced in these plasmas, thus it is likely that the sawtooth mitigationmore » in these experiments is due to the combination of the helically deformed core and the reduced rotation. The former effect is qualitatively well reproduced by nonlinear MHD simulations performed with the PIXIE3D code. The results obtained in these RFX-mod experiments motivated similar ones in DIII-D L-mode diverted Tokamak plasmas at low q 95. These experiments succeeded in reproducing the sawtooth mitigation with the approach developed in RFX-mod. In DIII-D this effect is correlated with a clear increase of the n = 1 plasma response, that indicates an enhancement of the coupling to the marginally stable n = 1 external kink, as simulations with the linear MHD code IPEC suggest. A significant rotation braking in the plasma core is also observed in DIII-D. Finally, numerical calculations of the neoclassical toroidal viscosity (NTV) carried out with PENT identify this torque as a possible contributor for this effect.« less

Recognition of Core- and Polymerase-derived immunogenic peptides included in novel therapeutic vaccine by T cells from Chinese chronic hepatitis B patients.

PubMed

Huang, D; Sansas, B; Jiang, J H; Gong, Q M; Jin, G D; Calais, V; Yu, D M; Zhu, M Y; Wei, D; Zhang, D H; Inchauspé, G; Zhang, X X; Zhu, R

2017-11-01

Chronic hepatitis B (CHB) is one of the major public health challenges in the world. Due to a strong interplay between specific T-cell immunity and elimination of hepatitis B virus (HBV), efforts to develop novel immunotherapeutics are gaining attention. TG1050, a novel immunotherapy, has shown efficacy in an animal study. To support the clinical development of TG1050 in China, specific immunity to the fusion antigens of TG1050 was assessed in Chinese patients. One hundred and thirty subjects were divided into three groups as CHB patients, HBV spontaneous resolvers, and CHB patients with HBsAg loss after antiviral treatment. HBV-specific T-cell responses to pools of HBV Core or Polymerase genotype D peptides included in TG1050 were evaluated. HBV Core- or Polymerase-specific cells were detected in peripheral blood mononuclear cells (PBMCs) from the different cohorts. The frequencies and intensities of HBV Core-specific immune responses were significantly lower in CHB patients than in HBsAg loss subjects. In CHB patients, a dominant pool derived from Polymerase (Pol1) was the most immunogenic. CHB patients with low viral loads (<10 6 IU/mL) were more likely to have a positive response specific to the Core peptide pool. Overall, genotype D-derived peptides included in TG1050 could raise broad and functional T-cell responses in PBMCs from Chinese CHB patients infected with genotype B/C isolates. Core-specific immunogenic domains appeared as "hot spots" with the capacity to differentiate between CHB vs HBsAg loss subjects. These observations support the extended application and associated immune monitoring of TG1050 in China. © 2017 The Authors. Journal of Viral Hepatitis Published by John Wiley & Sons Ltd.

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.

gC1qR expression in chimpanzees with resolved and chronic infection: Potential role of HCV core/gC1qR-mediated T cell suppression in the outcome of HCV infection

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

Yao Zhiqang; Shata, Mohamed Tarek; Tricoche, Nancy

2006-03-15

Chimpanzee is a unique animal model for HCV infection, in which about 50% of infections resolve spontaneously. It has been reported that the magnitude of T cell responses to HCV core in recovered chimpanzees is greater than that in chronically infected ones. However, the mechanism(s) by which the chimpanzees with resolved infection overcome core-mediated immunosuppression remains unknown. In this study, we examined the effect of HCV core on T cell responsiveness in chimpanzees with resolved and chronic HCV infection. We found that core protein strongly inhibited T cell activation and proliferation in chimpanzees with chronic infection, while this inhibition wasmore » limited in chimpanzees with resolved infection. Notably, the level of gC1qR, as well as the binding of core protein, on the surface of T cells was lower in recovered chimpanzees when compared to chimpanzees with chronic HCV infection. Intriguingly, the observed differences in gC1qR expression levels and susceptibility to core-induced suppression amongst HCV-chronically infected and recovered chimpanzees were observed prior to HCV challenge, suggesting a possible genetic determination of the outcome of infection. These findings suggest that gC1qR expression on the surface of T cells is crucial for HCV core-mediated T cell suppression and viral clearance, and that represents a novel mechanism by which a virus usurps host machinery for persistence.« less

Influence of multiband sign-changing superconductivity on vortex cores and vortex pinning in stoichiometric high- T c   CaKFe 4 As 4

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

Fente, Anton; Meier, William R.; Kong, Tai

We use a scanning tunneling microscope to study the superconducting density of states and vortex lattice of single crystals of CaKFe 4As 4. This material has a critical temperature of T c = 35 K, one of the highest among stoichiometric iron based superconductors (FeBSCs), and is comparable to T c found near optimal doping in other FeBSCs. We observe quasiparticle scattering from defects with a pattern related to interband scattering between zone centered hole sheets. We measure the tunneling conductance in vortex cores and find a peak due to Caroli–de Gennes–Matricon bound states. The peak is located above themore » Fermi level, showing that CaKFe 4As 4 is a clean superconductor with vortex core bound states close to the so-called extreme quantum limit. We identify locations where the superconducting order parameter is strongly suppressed due to pair breaking. Vortices are pinned at these locations, and the length scale of the suppression of the order parameter is of order of the vortex core size. Finally, as a consequence, the vortex lattice is disordered up to 8 T.« less

Influence of multiband sign-changing superconductivity on vortex cores and vortex pinning in stoichiometric high- T c   CaKFe 4 As 4

DOE PAGES

Fente, Anton; Meier, William R.; Kong, Tai; ...

2018-04-02

We use a scanning tunneling microscope to study the superconducting density of states and vortex lattice of single crystals of CaKFe 4As 4. This material has a critical temperature of T c = 35 K, one of the highest among stoichiometric iron based superconductors (FeBSCs), and is comparable to T c found near optimal doping in other FeBSCs. We observe quasiparticle scattering from defects with a pattern related to interband scattering between zone centered hole sheets. We measure the tunneling conductance in vortex cores and find a peak due to Caroli–de Gennes–Matricon bound states. The peak is located above themore » Fermi level, showing that CaKFe 4As 4 is a clean superconductor with vortex core bound states close to the so-called extreme quantum limit. We identify locations where the superconducting order parameter is strongly suppressed due to pair breaking. Vortices are pinned at these locations, and the length scale of the suppression of the order parameter is of order of the vortex core size. Finally, as a consequence, the vortex lattice is disordered up to 8 T.« less

A Neutral Beam for the Lithium Tokamak eXperiment Upgrade (LTX-U)

NASA Astrophysics Data System (ADS)

Merino, Enrique; Majeski, Richard; Kaita, Robert; Kozub, Thomas; Boyle, Dennis; Schmitt, John; Smirnov, Artem

2015-11-01

Neutral beam injection into tokamaks is a proven method of plasma heating and fueling. In LTX, high confinement discharges have been achieved with low-recycling lithium walls. To further improve plasma performance, a neutral beam (NB) will be installed as part of an upgrade to LTX (LTX-U). The NB will provide core plasma fueling with up to 700 kW of injected power. Requirements for accommodating the NB include the addition of injection and beam-dump ports onto the vessel and enhancement of the vacuum vessel pumping capability. Because the NB can also serve as a source of neutrals for charge-exchange recombination spectroscopy, ``active'' spectroscopic diagnostics will also be developed. An overview of these plans and other improvements for upgrading LTX to LTX-U will be presented. Supported by US DOE contracts DE-AC02-09CH11466 and DE-AC52-07NA27344.

Effect of change in core width and core refractive index on the switching behavior of a nonlinear Mach-Zehnder interferometer

NASA Astrophysics Data System (ADS)

Srivastava, Arpita; Medhekar, Sarang

2012-09-01

The effect of variation of core width and core refractive index on output versus input (O-I) and transmission coefficient versus input (T-I) characteristics of a nonlinear Mach-Zehnder interferometer (NMZI) were investigated for the first time. Beam propagation method has been used for the present investigation. Change of core width and/or core refractive index adds extra liberty for changing the operating power of an NMZI to a desired value. Moreover, it is revealed for the first time that use of only the O-I or T-I characteristic presents an incomplete picture of NMZI switching; both O-I and T-I characteristics of both balanced/imbalanced NMZIs are indispensible for complete understanding of NMZI switching.

Enhancement of First Wall Damage in Iter Type Tokamak due to Lenr Effects

NASA Astrophysics Data System (ADS)

Lipson, Andrei G.; Miley, George H.; Momota, Hiromu

In recent experiments with pulsed periodic high current (J ~ 300-500 mA/cm2) D2-glow discharge at deuteron energies as low as 0.8-2.45 keV a large DD-reaction yield has been obtained. Thick target yield measurement show unusually high DD-reaction enhancement (at Ed = 1 keV the yield is about nine orders of magnitude larger than that deduced from standard Bosch and Halle extrapolation of DD-reaction cross-section to lower energies) The results obtained in these LENR experiments with glow discharge suggest nonnegligible edge plasma effects in the ITER TOKAMAK that were previously ignored. In the case of the ITER DT plasma core, we here estimate the DT reaction yield at the metal edge due to plasma ion bombardment of the first wall and/or divertor materials.

Electron Cyclotron Current Drive Efficiency in General Tokamak Geometry and Its Application to Advanced Tokamak Plasmas

NASA Astrophysics Data System (ADS)

Lin-Liu, Y. R.; Chan, V. S.; Luce, T. C.; Prater, R.

1998-11-01

Owing to relativistic mass shift in the cyclotron resonance condition, a simple and accurate interpolation formula for estimating the current drive efficiency, such as those(S.C. Chiu et al.), Nucl. Fusion 29, 2175 (1989).^,(D.A. Ehst and C.F.F. Karney, Nucl. Fusion 31), 1933 (1991). commonly used in FWCD, is not available in the case of ECCD. In this work, we model ECCD using the adjoint techniques. A semi-analytic adjoint function appropriate for general tokamak geometry is obtained using Fisch's relativistic collision model. Predictions of off-axis ECCD qualitatively and semi-quantitatively agrees with those of Cohen,(R.H. Cohen, Phys. Fluids 30), 2442 (1987). currently implemented in the raytracing code TORAY. The dependences of the current drive efficiency on the wave launch configuration and the plasma parameters will be presented. Strong absorption of the wave away from the resonance layer is shown to be an important factor in optimizing the off-axis ECCD for application to advanced tokamak operations.

A self-consistent model of an isothermal tokamak

NASA Astrophysics Data System (ADS)

McNamara, Steven; Lilley, Matthew

2014-10-01

Continued progress in liquid lithium coating technologies have made the development of a beam driven tokamak with minimal edge recycling a feasibly possibility. Such devices are characterised by improved confinement due to their inherent stability and the suppression of thermal conduction. Particle and energy confinement become intrinsically linked and the plasma thermal energy content is governed by the injected beam. A self-consistent model of a purely beam fuelled isothermal tokamak is presented, including calculations of the density profile, bulk species temperature ratios and the fusion output. Stability considerations constrain the operating parameters and regions of stable operation are identified and their suitability to potential reactor applications discussed.

What is the fate of runaway positrons in tokamaks?

DOE PAGES

Liu, Jian; Qin, Hong; Fisch, Nathaniel J.; ...

2014-06-19

In this study, massive runaway positrons are generated by runaway electrons in tokamaks. The fate of these positrons encodes valuable information about the runaway dynamics. The phase space dynamics of a runaway position is investigated using a Lagrangian that incorporates the tokamak geometry, loop voltage, radiation and collisional effects. It is found numerically that runaway positrons will drift out of the plasma to annihilate on the first wall, with an in-plasma annihilation possibility less than 0.1%. The dynamics of runaway positrons provides signatures that can be observed as diagnostic tools.

What is the fate of runaway positrons in tokamaks?

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

Liu, Jian; Qin, Hong, E-mail: hongqin@ustc.edu.cn; Princeton Plasma Physics Laboratory, Princeton University, Princeton, New Jersey 08543

2014-06-15

Massive runaway positrons are generated by runaway electrons in tokamaks. The fate of these positrons encodes valuable information about the runaway dynamics. The phase space dynamics of a runaway position is investigated using a Lagrangian that incorporates the tokamak geometry, loop voltage, radiation and collisional effects. It is found numerically that runaway positrons will drift out of the plasma to annihilate on the first wall, with an in-plasma annihilation possibility less than 0.1%. The dynamics of runaway positrons provides signatures that can be observed as diagnostic tools.

A survey of electron Bernstein wave heating and current drive potential for spherical tokamaks

NASA Astrophysics Data System (ADS)

Urban, Jakub; Decker, Joan; Peysson, Yves; Preinhaelter, Josef; Shevchenko, Vladimir; Taylor, Gary; Vahala, Linda; Vahala, George

2011-08-01

The electron Bernstein wave (EBW) is typically the only wave in the electron cyclotron (EC) range that can be applied in spherical tokamaks for heating and current drive (H&CD). Spherical tokamaks (STs) operate generally in high-β regimes, in which the usual EC O- and X-modes are cut off. In this case, EBWs seem to be the only option that can provide features similar to the EC waves—controllable localized H&CD that can be used for core plasma heating as well as for accurate plasma stabilization. The EBW is a quasi-electrostatic wave that can be excited by mode conversion from a suitably launched O- or X-mode; its propagation further inside the plasma is strongly influenced by the plasma parameters. These rather awkward properties make its application somewhat more difficult. In this paper we perform an extensive numerical study of EBW H&CD performance in four typical ST plasmas (NSTX L- and H-mode, MAST Upgrade, NHTX). Coupled ray-tracing (AMR) and Fokker-Planck (LUKE) codes are employed to simulate EBWs of varying frequencies and launch conditions, which are the fundamental EBW parameters that can be chosen and controlled. Our results indicate that an efficient and universal EBW H&CD system is indeed viable. In particular, power can be deposited and current reasonably efficiently driven across the whole plasma radius. Such a system could be controlled by a suitably chosen launching antenna vertical position and would also be sufficiently robust.

Currents in the DIII-D Tokamak

NASA Astrophysics Data System (ADS)

Azari, A.; Eidietis, N. W.

2012-10-01

Loss of vertical control of an elongated tokamak plasma results in a vertical displacement event (VDE) which can induce large currents on open field lines and exert high JxB forces on in-vessel components. An array of first-wall tile current monitors on DIII-D provides direct measurement of the poloidal halo currents. These measurements are analyzed to create a database of halo current magnitude and asymmetry, which are found to lie within the ranges seen by numerous other tokamaks in the ITPA Disruption Database. In addition, an analysis of halo asymmetry rotation is presented, as rotation at the resonance frequencies of in-vessel components could lead to significant amplification of the halo forces. Halo current rotation is found to be far more prevalent in old (1997-2002) DIII-D halo current data than recent data (2009), perhaps due to a change in divertor geometry over that time.

The Physics of Tokamak Start-up

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

D. Mueller

Tokamak start-up on present-day devices usually relies on inductively induced voltage from a central solenoid. In some cases inductive startup is assisted with auxiliary power from electron cyclotron radio frequency heating. ITER, the National Spherical Torus eXperiment Upgrade and JT60, now under construction, will make use of the understanding gained from present-day devices to ensure successful start-up. Design of a spherical tokamak (ST) with DT capability for nuclear component testing would require an alternative to a central solenoid because the small central column in an ST has insufficient space to provide shielding for the insulators in the solenoid. Alternative start-upmore » techniques such as induction using outer poloidal field coils, electron Bernstein wave start-up, coaxial helicity injection and point source helicity injection have been used with success, but require demonstration of scaling to higher plasma current.« less

The physics of tokamak start-up

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

Mueller, D.

Tokamak start-up on present-day devices usually relies on inductively induced voltage from a central solenoid. In some cases, inductive startup is assisted with auxiliary power from electron cyclotron radio frequency heating. International Thermonuclear Experimental Reactor, the National Spherical Torus Experiment Upgrade and JT60, now under construction, will make use of the understanding gained from present-day devices to ensure successful start-up. Design of a spherical tokamak (ST) with DT capability for nuclear component testing would require an alternative to a central solenoid because the small central column in an ST has insufficient space to provide shielding for the insulators in themore » solenoid. Alternative start-up techniques such as induction using outer poloidal field coils, electron Bernstein wave start-up, coaxial helicity injection, and point source helicity injection have been used with success, but require demonstration of scaling to higher plasma current.« less

Statistical Analysis Software for the TRS-80 Microcomputer.

DTIC Science & Technology

1981-09-01

1160 3 W640 : KSa3 ; GOSUB 760 1170 CLS : PRINT"Is the variance known?" 1190 ZP64: TI"’T" 1 T2S-N : CIOSU1 710 1190 IF 70- en " OR mO"N" 0070 1250 1200...Paul Isbell September 1981 C--) Thesis Advisor: Charles F. Taylor, Jr. LLU __ Approved for public release; distribution unlimited C. 3 - ? 01 12 032...ACCUIiON no . 3 . 1CIPIIT*S CATALOG NMUNIRO, 4. TITLE (god Su feiej) S. TYPE OP MI[PORT 6 PEIOo COVI[EO Stastical Analysis Software for the Master’s Thesis

The ARIES Advanced And Conservative Tokamak (ACT) Power Plant Study

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

Kessel, C. E.; Poli, F. M.; Ghantous, K.

2014-03-05

Tokamak power plants are studied with advanced and conservative design philosophies in order to identify the impacts on the resulting designs and to provide guidance to critical research needs. Incorporating updated physics understanding, and using more sophisticated engineering and physics analysis, the tokamak configurations have developed a more credible basis compared to older studies. The advanced configuration assumes a self-cooled lead lithium (SCLL) blanket concept with SiC composite structural material with 58% thermal conversion efficiency. This plasma has a major radius of 6.25 m, a toroidal field of 6.0 T, a q95 of 4.5, a βN total of 5.75, Hmore » 98 of 1.65, n/nGr of 1.0, and peak divertor heat flux of 13.7 MW/m 2. The conservative configuration assumes a dual coolant lead lithium (DCLL) blanket concept with ferritic steel structural material and helium coolant, achieving a thermal conversion efficiency of 45%. The plasma major radius is 9.75 m, a toroidal field of 8.75 T, a q95 of 8.0, a βN total of 2.5, H 98 of 1.25, n/n Gr of 1.3, and peak divertor heat flux of 10 MW/m 2. The divertor heat flux treatment with a narrow power scrape-off width has driven the plasmas to larger major radius. Edge and divertor plasma simulations are targeting a basis for high radiated power fraction in the divertor, which is necessary for solutions to keep the peak heat flux in the range of 10-15 MW/m 2. Combinations of the advanced and conservative approaches show intermediate sizes. A new systems code using a database approach has been used and shows that the operating point is really an operating zone with some range of plasma and engineering parameters and very similar costs of electricity. Papers in this issue provide more detailed discussion of the work summarized here.« less

Profile measurements of the electron temperature on the ASDEX Upgrade, COMPASS, and ISTTOK tokamak using Thomson scattering, triple, and ball-pen probes

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

Adamek, J., E-mail: adamek@ipp.cas.cz; Horacek, J.; Bilkova, P.

The ball-pen probe (BPP) technique is used successfully to make profile measurements of the electron temperature on the ASDEX Upgrade (Axially Symmetric Divertor Experiment), COMPASS (COMPact ASSembly), and ISTTOK (Instituto Superior Tecnico TOKamak) tokamak. The electron temperature is provided by a combination of the BPP potential (Φ{sub BPP}) and the floating potential (V{sub fl}) of the Langmuir probe (LP), which is compared with the Thomson scattering diagnostic on ASDEX Upgrade and COMPASS. Excellent agreement between the two diagnostics is obtained for circular and diverted plasmas and different heating mechanisms (Ohmic, NBI, ECRH) in deuterium discharges with the same formula T{submore » e} = (Φ{sub BPP} − V{sub fl})/2.2. The comparative measurements of the electron temperature using BPP/LP and triple probe (TP) techniques on the ISTTOK tokamak show good agreement of averaged values only inside the separatrix. It was also found that the TP provides the electron temperature with significantly higher standard deviation than BPP/LP. However, the resulting values of both techniques are well in the phase with the maximum of cross-correlation function being 0.8.« less

Plasma current start-up using the lower hybrid wave on the TST-2 spherical tokamak

NASA Astrophysics Data System (ADS)

Takase, Y.; Ejiri, A.; Inada, T.; Moeller, C. P.; Shinya, T.; Tsujii, N.; Yajima, S.; Furui, H.; Homma, H.; Imamura, K.; Nakamura, K.; Nakamura, K.; Sonehara, M.; Takeuchi, T.; Togashi, H.; Tsuda, S.; Yoshida, Y.

2015-12-01

Non-inductive plasma current start-up, ramp-up and sustainment by waves in the lower hybrid wave (LHW) frequency range at 200 MHz were investigated on the TST-2 spherical tokamak (R0 ≤ 0.38 m, a ≤ 0.25 m, Bt0 ≤ 0.3T, Ip ≤ 0.14 MA). Experimental results obtained using three types of antenna were compared. Both the highest plasma current (Ip = 18 kA) and the highest current drive figure of merit ηCD≡n¯eIpR0/PRF=1.4 ×1017 A/W/m2 were achieved using the capacitively-coupled combline (CCC) antenna, designed to excite the LHW with a sharp and highly directional wavenumber spectrum. For Ip greater than about 5 kA, high energy electrons accelerated by the LHW become the dominant carrier of plasma current. The low value of ηCD observed so far are believed to be caused by a rapid loss of energetic electrons and parasitic losses of the LHW energy in the plasma periphery. ηCD is expected to improve by an order of magnitude by increasing the plasma current to improve energetic electron confinement. In addition, edge power losses are expected to be reduced by increasing the toroidal magnetic field to improve wave accessibility to the plasma core, and by launching the LHW from the inboard upper region of the torus to achieve better single-pass absorption.

Development of a Method for Local Electron Temperature and Density Measurements in the Divertor of the JET Tokamak

NASA Technical Reports Server (NTRS)

Jupen, C.; Meigs, A.; Bhatia, A. K.; Brezinsek, S.; OMullane, M.

2004-01-01

Plasma volume recombination in the divertor, a process in which charged particles recombine to neutral atoms, contributes to plasma detachment and hence cooling at the divertor target region. Detachment has been observed at JET and other tokamaks and is known to occur at low electron temperatures (T(sub e)<1 eV) and at high electron density (n(sub e)>10(exp 20)/m(exp 3)). The ability to measure such low temperatures is therefore of interest for modelling the divertor. In present work we report development of a new spectroscopic technique for investigation of local electron density (n(sub e)) and temperature (T,) in the outer divertor at JET.

Parameter exploration for a Compact Advanced Tokamak DEMO

NASA Astrophysics Data System (ADS)

Weisberg, D. B.; Buttery, R. J.; Ferron, J. R.; Garofalo, A. M.; Snyder, P. B.; Turnbull, A. D.; Holcomb, C. T.; McClenaghan, J.; Canik, J.; Park, J.-M.

2017-10-01

A new parameter study has explored a range of design points to assess the physics feasibility for a compact 200MWe advanced tokamak DEMO that combines high beta (βN < 4) and high toroidal field (BT = 6 - 7 T). A unique aspect of this study is the use of a FASTRAN modeling suite that combines integrated transport, pedestal, stability, and heating & current drive calculations to predict steady-state solutions with neutral beam and helicon powered current drive. This study has identified a range of design solutions in a compact (R0 = 4 m), high-field (BT = 6 - 7 T), strongly-shaped (κ = 2 , δ = 0.6) device. Unlike previous proposals, C-AT DEMO takes advantage of high-beta operation as well as emerging advances in magnet technology to demonstrate net electric production in a moderately sized machine. We present results showing that the large bootstrap fraction and low recirculating power enabled by high normalized beta can achieve tolerable heat and neutron load with good H-mode access. The prediction of operating points with simultaneously achieved high-confinement (H98 < 1.3), high-density (fGW < 1.3), and high-beta warrants additional assessment of this approach towards a cost-attractive DEMO device. Work supported by the US DOE under DE-FC02-04ER54698.

Simulations of Tokamak Edge Turbulence Including Self-Consistent Zonal Flows

NASA Astrophysics Data System (ADS)

Cohen, Bruce; Umansky, Maxim

2013-10-01

Progress on simulations of electromagnetic drift-resistive ballooning turbulence in the tokamak edge is summarized in this mini-conference talk. A more detailed report on this work is presented in a poster at this conference. This work extends our previous work to include self-consistent zonal flows and their effects. The previous work addressed the simulation of L-mode tokamak edge turbulence using the turbulence code BOUT. The calculations used realistic single-null geometry and plasma parameters of the DIII-D tokamak and produced fluctuation amplitudes, fluctuation spectra, and particle and thermal fluxes that compare favorably to experimental data. In the effect of sheared ExB poloidal rotation is included with an imposed static radial electric field fitted to experimental data. In the new work here we include the radial electric field self-consistently driven by the microturbulence, which contributes to the sheared ExB poloidal rotation (zonal flow generation). We present simulations with/without zonal flows for both cylindrical geometry, as in the UCLA Large Plasma Device, and for the DIII-D tokamak L-mode cases in to quantify the influence of self-consistent zonal flows on the microturbulence and the concomitant transport. This work was performed under the auspices of the US Department of Energy under contract DE-AC52-07NA27344 at the Lawrence Livermore National Laboratory.

Three-dimensional analysis of tokamaks and stellarators

PubMed Central

Garabedian, Paul R.

2008-01-01

The NSTAB equilibrium and stability code and the TRAN Monte Carlo transport code furnish a simple but effective numerical simulation of essential features of present tokamak and stellarator experiments. When the mesh size is comparable to the island width, an accurate radial difference scheme in conservation form captures magnetic islands successfully despite a nested surface hypothesis imposed by the mathematics. Three-dimensional asymmetries in bifurcated numerical solutions of the axially symmetric tokamak problem are relevant to the observation of unstable neoclassical tearing modes and edge localized modes in experiments. Islands in compact stellarators with quasiaxial symmetry are easier to control, so these configurations will become good candidates for magnetic fusion if difficulties with safety and stability are encountered in the International Thermonuclear Experimental Reactor (ITER) project. PMID:18768807

Simulations of Turbulence in Tokamak Edge and Effects of Self-Consistent Zonal Flows

NASA Astrophysics Data System (ADS)

Cohen, Bruce; Umansky, Maxim

2013-10-01

Progress is reported on simulations of electromagnetic drift-resistive ballooning turbulence in the tokamak edge. This extends previous work to include self-consistent zonal flows and their effects. The previous work addressed simulation of L-mode tokamak edge turbulence using the turbulence code BOUT that solves Braginskii-based plasma fluid equations in tokamak edge domain. The calculations use realistic single-null geometry and plasma parameters of the DIII-D tokamak and produce fluctuation amplitudes, fluctuation spectra, and particle and thermal fluxes that compare favorably to experimental data. In the effect of sheared ExB poloidal rotation is included with an imposed static radial electric field fitted to experimental data. In the new work here we include the radial electric field self-consistently driven by the microturbulence, which contributes to the sheared ExB poloidal rotation (zonal flow generation). We present simulations with/without zonal flows for both cylindrical geometry, as in the UCLA Large Plasma Device, and for the DIII-D tokamak L-mode cases in to quantify the influence of self-consistent zonal flows on the microturbulence and the concomitant transport. This work was performed under the auspices of the U.S. Department of Energy under contract DE-AC52-07NA27344 at the Lawrence Livermore National Laboratory.

Features of self-organized plasma physics in tokamaks

NASA Astrophysics Data System (ADS)

Razumova, K. A.

2018-01-01

The history of investigations the role of self-organization processes in tokamak plasma confinement is presented. It was experimentally shown that the normalized pressure profile is the same for different tokamaks. Instead of the conventional Fick equation, where the thermal flux is proportional to a pressure gradient, processes in the plasma are well described by the Dyabilanin’s energy balance equation, in which the heat flux is proportional to the difference of normalized gradients for self-consistent and real pressure profiles. The transport coefficient depends on the values of heat flux, which compensates distortion of the pressure profile with external impacts. Radiative cooling of the plasma edge decreases the heat flux and improves the confinement.

DOUBLE code simulations of emissivities of fast neutrals for different plasma observation view-lines of neutral particle analyzers on the COMPASS tokamak

NASA Astrophysics Data System (ADS)

Mitosinkova, K.; Tomes, M.; Stockel, J.; Varju, J.; Stano, M.

2018-03-01

Neutral particle analyzers (NPA) measure line-integrated energy spectra of fast neutral atoms escaping the tokamak plasma, which are a product of charge-exchange (CX) collisions of plasma ions with background neutrals. They can observe variations in the ion temperature T i of non-thermal fast ions created by additional plasma heating. However, the plasma column which a fast atom has to pass through must be sufficiently short in comparison with the fast atom’s mean-free-path. Tokamak COMPASS is currently equipped with one NPA installed at a tangential mid-plane port. This orientation is optimal for observing non-thermal fast ions. However, in this configuration the signal at energies useful for T i derivation is lost in noise due to the too long fast atoms’ trajectories. Thus, a second NPA is planned to be connected for the purpose of measuring T i. We analyzed different possible view-lines (perpendicular mid-plane, tangential mid-plane, and top view) for the second NPA using the DOUBLE Monte-Carlo code and compared the results with the performance of the present NPA with tangential orientation. The DOUBLE code provides fast-atoms’ emissivity functions along the NPA view-line. The position of the median of these emissivity functions is related to the location from where the measured signal originates. Further, we compared the difference between the real central T i used as a DOUBLE code input and the T iCX derived from the exponential decay of simulated energy spectra. The advantages and disadvantages of each NPA location are discussed.

Core binding factors are necessary for natural killer cell development and cooperate with Notch signaling during T-cell specification

PubMed Central

Guo, Yalin; Maillard, Ivan; Chakraborti, Sankhamala; Rothenberg, Ellen V.

2008-01-01

CBFβ is the non-DNA binding subunit of the core binding factors (CBFs). Mice with reduced CBFβ levels display profound, early defects in T-cell but not B-cell development. Here we show that CBFβ is also required at very early stages of natural killer (NK)–cell development. We also demonstrate that T-cell development aborts during specification, as the expression of Gata3 and Tcf7, which encode key regulators of T lineage specification, is substantially reduced, as are functional thymic progenitors. Constitutively active Notch or IL-7 signaling cannot restore T-cell expansion or differentiation of CBFβ insufficient cells, nor can overexpression of Runx1 or CBFβ overcome a lack of Notch signaling. Therefore, the ability of the prethymic cell to respond appropriately to Notch is dependent on CBFβ, and both signals converge to activate the T-cell developmental program. PMID:18390836

How much does a tokamak reactor cost?

NASA Astrophysics Data System (ADS)

Freidberg, J.; Cerfon, A.; Ballinger, S.; Barber, J.; Dogra, A.; McCarthy, W.; Milanese, L.; Mouratidis, T.; Redman, W.; Sandberg, A.; Segal, D.; Simpson, R.; Sorensen, C.; Zhou, M.

2017-10-01

The cost of a fusion reactor is of critical importance to its ultimate acceptability as a commercial source of electricity. While there are general rules of thumb for scaling both overnight cost and levelized cost of electricity the corresponding relations are not very accurate or universally agreed upon. We have carried out a series of scaling studies of tokamak reactor costs based on reasonably sophisticated plasma and engineering models. The analysis is largely analytic, requiring only a simple numerical code, thus allowing a very large number of designs. Importantly, the studies are aimed at plasma physicists rather than fusion engineers. The goals are to assess the pros and cons of steady state burning plasma experiments and reactors. One specific set of results discusses the benefits of higher magnetic fields, now possible because of the recent development of high T rare earth superconductors (REBCO); with this goal in mind, we calculate quantitative expressions, including both scaling and multiplicative constants, for cost and major radius as a function of central magnetic field.

Measurement of turbulent electron temperature fluctuations on the ASDEX Upgrade tokamak using correlated electron cyclotron emission

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

Freethy, S. J., E-mail: simon.freethy@ipp.mpg.de; Plasma Science and Fusion Center, Massachusetts Institute of Technology, Cambridge, Massachusetts 02139; Conway, G. D.

2016-11-15

Turbulent temperature fluctuations are measured on the ASDEX Upgrade tokamak using pairs of closely spaced, narrow-band heterodyne radiometer channels and a standard correlation technique. The pre-detection spacing and bandwidth of the radiometer channel pairs is chosen such that they are physically separated less than a turbulent correlation length, but do not overlap. The radiometer has 4 fixed filter frequency channels and two tunable filter channels for added flexibility in the measurement position. Relative temperature fluctuation amplitudes are observed in a helium plasma to be δT/T = (0.76 ± 0.02)%, (0.67 ± 0.02)%, and (0.59 ± 0.03)% at normalised toroidal fluxmore » radius of ρ{sub tor} = 0.82, 0.75, and 0.68, respectively.« less

The Experiment of Modulated Toroidal Current on HT-7 and HT-6M Tokamak

NASA Astrophysics Data System (ADS)

Mao, Jian-shan; P, Phillips; Luo, Jia-rong; Xu, Yu-hong; Zhao, Jun-yu; Zhang, Xian-mei; Wan, Bao-nian; Zhang, Shou-yin; Jie, Yin-xian; Wu, Zhen-wei; Hu, Li-qun; Liu, Sheng-xia; Shi, Yue-jiang; Li, Jian-gang; HT-6M; HT-7 Group

2003-02-01

The Experiments of Modulated Toroidal Current were done on the HT-6M tokamak and HT-7 superconducting tokamak. The toroidal current was modulated by programming the Ohmic heating field. Modulation of the plasma current has been used successfully to suppress MHD activity in discharges near the density limit where large MHD m = 2 tearing modes were suppressed by sufficiently large plasma current oscillations. The improved Ohmic confinement phase was observed during modulating toroidal current (MTC) on the Hefei Tokamak-6M (HT-6M) and Hefei superconducting Tokamak-7 (HT-7). A toroidal frequency-modulated current, induced by a modulated loop voltage, was added on the plasma equilibrium current. The ratio of A.C. amplitude of plasma current to the main plasma current ΔIp/Ip is about 12%-30%. The different formats of the frequency-modulated toroidal current were compared.

Hepatitis C Virus Core Mutations Associated with False-Negative Serological Results for Genotype 3a Core Antigen

PubMed Central

Dunford, Linda; Freitas, Ines; Holder, Paul; Nguyen, Lan Anh; O'Gorman, Joanne; Connell, Jeff; Carr, Michael; Hall, William; De Gascun, Cillian

2015-01-01

Genetic characterization of the genotype 3a (GT3a) hepatitis C virus (HCV) core region from HCV core antigen (HCVcAg)-negative/RNA-positive cases and HCVcAg-positive/RNA-positive controls identified significant associations between the substitutions A48T and T49A/P and failure to detect HCVcAg (P < 0.05). Polymorphisms at residues 48 and 49 in the core protein are present across all major epidemic and endemic GTs. These findings have implications for HCV diagnosis, particularly in low-income regions in which GT3a HCV is endemic. PMID:25994168

Electron cyclotron emission from nonthermal tokamak plasmas

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

Harvey, R.W.; O'Brien, M.R.; Rozhdestvensky, V.V.

1993-02-01

Electron cyclotron emission can be a sensitive indicator of nonthermal electron distributions. A new, comprehensive ray-tracing and cyclotron emission code that is aimed at predicting and interpreting the cyclotron emission from tokamak plasmas is described. The radiation transfer equation is solved along Wentzel--Kramers--Brillouin (WKB) rays using a fully relativistic calculation of the emission and absorption from electron distributions that are gyrotropic and toroidally symmetric, but may be otherwise arbitrary functions of the constants of motion. Using a radial array of electron distributions obtained from a bounce-averaged Fokker--Planck code modeling dc electron field and electron cyclotron heating effects, the cyclotron emissionmore » spectra are obtained. A pronounced strong nonthermal cyclotron emission feature that occurs at frequencies relativistically downshifted to second harmonic cyclotron frequencies outside the tokamak is calculated, in agreement with experimental results from the DIII-D [J. L. Luxon and L. G. Davies, Fusion Technol. [bold 8], 441 (1985)] and FT-1 [D. G. Bulyginsky [ital et] [ital al]., in [ital Proceedings] [ital of] [ital the] 15[ital th] [ital European] [ital Conference] [ital on] [ital Controlled] [ital Fusion] [ital and] [ital Plasma] [ital Heating], Dubrovnik, 1988 (European Physical Society, Petit-Lancy, 1988), Vol. 12B, Part II, p. 823] tokamaks. The calculations indicate the presence of a strong loss mechanism that operates on electrons in the 100--150 keV energy range.« less

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

The composition of Earth's core from equations of state, metal-silicate partitioning, and core formation modeling

NASA Astrophysics Data System (ADS)

Fischer, Rebecca; Campbell, Andrew; Ciesla, Fred

2016-04-01

The Earth accreted in a series of increasingly large and violent collisions. Simultaneously, the metallic core segregated from the silicate mantle, acquiring its modern composition through high pressure (P), high temperature (T) partitioning reactions. Here we present a model that couples these aspects of early planetary evolution, building on recent accretion simulations and metal-silicate partitioning experiments, constrained by density measurements of Fe-rich alloys. Previously, the equations of state of FeO, Fe-9Si, Fe-16Si, and FeSi were measured to megabar pressures and several thousand K using a laser-heated diamond anvil cell. With these equations of state, we determined that the core's density can be reproduced through the addition of 11.3 +/- 0.6 wt% silicon or 8.1 +/- 1.1 wt% oxygen to an Fe-Ni alloy (Fischer et al., 2011, 2014). Metal-silicate partitioning experiments of Ni, Co, V, Cr, Si, and O have been performed in a diamond anvil cell to 100 GPa and 5700 K, allowing the effects of P, T, and composition on the partitioning behaviors of these elements to be parameterized (Fischer et al., 2015; Siebert et al., 2012). Here we apply those experimental results to model Earth's core formation, using N-body simulations to describe the delivery, masses, and original locations of planetary building blocks (Fischer and Ciesla, 2014). As planets accrete, their core and mantle compositions are modified by high P-T reactions with each collision (Rubie et al., 2011). For partial equilibration of the mantle at 55% of the evolving core-mantle boundary pressure and the liquidus temperature, we find that the core contains 5.4 wt% Si and 1.9 wt% O. This composition is consistent with the seismologically-inferred density of Earth's core, based on comparisons to our equations of state, and indicate that the core cannot contain more than ~2 wt% S or C. Earth analogues experience 1.2 +/- 0.2 log units of oxidation during accretion, due to both the effects of high P-T

A model of energetic ion effects on pressure driven tearing modes in tokamaks

DOE PAGES

Halfmoon, M. R.; Brennan, D. P.

2017-06-05

Here, the effects that energetic trapped ions have on linear resistive magnetohydrodynamic (MHD) instabilities are studied in a reduced model that captures the essential physics driving or damping the modes through variations in the magnetic shear. The drift-kinetic orbital interaction of a slowing down distribution of trapped energetic ions with a resistive MHD instability is integrated to a scalar contribution to the perturbed pressure, and entered into an asymptotic matching formalism for the resistive MHD dispersion relation. Toroidal magnetic field line curvature is included to model trapping in the particle distribution, in an otherwise cylindrical model. The focus is onmore » a configuration that is driven unstable to the m/n = 2/1 mode by increasing pressure, where m is the poloidal mode number and n is the toroidal. The particles and pressure can affect the mode both in the core region where there can be low and reversed shear and outside the resonant surface in significant positive shear. The results show that the energetic ions damp and stabilize the mode when orbiting in significant positive shear, increasing the marginal stability boundary. However, the inner core region contribution with low and reversed shear can drive the mode unstable. This effect of shear on the energetic ion pressure contribution is found to be consistent with the literature. These results explain the observation that the 2/1 mode was found to be damped and stabilized by energetic ions in delta δf-MHD simulations of tokamak experiments with positive shear throughout, while the 2/1 mode was found to be driven unstable in simulations of experiments with weakly reversed shear in the core. This is also found to be consistent with related experimental observations of the stability of the 2/1 mode changing significantly with core shear.« less

A model of energetic ion effects on pressure driven tearing modes in tokamaks

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

Halfmoon, M. R.; Brennan, D. P.

Here, the effects that energetic trapped ions have on linear resistive magnetohydrodynamic (MHD) instabilities are studied in a reduced model that captures the essential physics driving or damping the modes through variations in the magnetic shear. The drift-kinetic orbital interaction of a slowing down distribution of trapped energetic ions with a resistive MHD instability is integrated to a scalar contribution to the perturbed pressure, and entered into an asymptotic matching formalism for the resistive MHD dispersion relation. Toroidal magnetic field line curvature is included to model trapping in the particle distribution, in an otherwise cylindrical model. The focus is onmore » a configuration that is driven unstable to the m/n = 2/1 mode by increasing pressure, where m is the poloidal mode number and n is the toroidal. The particles and pressure can affect the mode both in the core region where there can be low and reversed shear and outside the resonant surface in significant positive shear. The results show that the energetic ions damp and stabilize the mode when orbiting in significant positive shear, increasing the marginal stability boundary. However, the inner core region contribution with low and reversed shear can drive the mode unstable. This effect of shear on the energetic ion pressure contribution is found to be consistent with the literature. These results explain the observation that the 2/1 mode was found to be damped and stabilized by energetic ions in delta δf-MHD simulations of tokamak experiments with positive shear throughout, while the 2/1 mode was found to be driven unstable in simulations of experiments with weakly reversed shear in the core. This is also found to be consistent with related experimental observations of the stability of the 2/1 mode changing significantly with core shear.« less

Physics of GAM-initiated L-H transition in a tokamak

NASA Astrophysics Data System (ADS)

Askinazi, L. G.; Belokurov, A. A.; Bulanin, V. V.; Gurchenko, A. D.; Gusakov, E. Z.; Kiviniemi, T. P.; Lebedev, S. V.; Kornev, V. A.; Korpilo, T.; Krikunov, S. V.; Leerink, S.; Machielsen, M.; Niskala, P.; Petrov, A. V.; Tukachinsky, A. S.; Yashin, A. Yu; Zhubr, N. A.

2017-01-01

Based on experimental observations using the TUMAN-3M and FT-2 tokamaks, and the results of gyrokinetic modeling of the interplay between turbulence and the geodesic acoustic mode (GAM) in these installations, a simple model is proposed for the analysis of the conditions required for L-H transition triggering by a burst of radial electric field oscillations in a tokamak. In the framework of this model, one-dimensional density evolution is considered to be governed by an anomalous diffusion coefficient dependent on radial electric field shear. The radial electric field is taken as the sum of the oscillating term and the quasi-stationary one determined by density and ion temperature gradients through a neoclassical formula. If the oscillating field parameters (amplitude, frequency, etc) are properly adjusted, a transport barrier forms at the plasma periphery and sustains after the oscillations are switched off, manifesting a transition into the high confinement mode with a strong inhomogeneous radial electric field and suppressed transport at the plasma edge. The electric field oscillation parameters required for L-H transition triggering are compared with the GAM parameters observed at the TUMAN-3M (in the discharges with ohmic L-H transition) and FT-2 tokamaks (where no clear L-H transition was observed). It is concluded based on this comparison that the GAM may act as a trigger for the L-H transition, provided that certain conditions for GAM oscillation and tokamak discharge are met.

Optimal Control Techniques for ResistiveWall Modes in Tokamaks

NASA Astrophysics Data System (ADS)

Clement, Mitchell Dobbs Pearson

Tokamaks can excite kink modes that can lock or nearly lock to the vacuum vessel wall, and whose rotation frequencies and growth rates vary in time but are generally inversely proportional to the magnetic flux diffusion time of the vacuum vessel wall. This magnetohydrodynamic (MHD) instability is pressure limiting in tokamaks and is called the Resistive Wall Mode (RWM). Future tokamaks that are expected to operate as fusion reactors will be required to maximize plasma pressure in order to maximize fusion performance. The DIII-D tokamak is equipped with electromagnetic control coils, both inside and outside of its vacuum vessel, which create magnetic fields that are small by comparison to the machine's equilibrium field but are able to dynamically counteract the RWM. Presently for RWM feedback, DIII-D uses its interior control coils using a classical proportional gain only controller to achieve high plasma pressure. Future advanced tokamak designs will not likely have the luxury of interior control coils and a proportional gain algorithm is not expected to be effective with external control coils. The computer code VALEN was designed to calculate the performance of an MHD feedback control system in an arbitrary geometry. VALEN models the perturbed magnetic field from a single MHD instability and its interaction with surrounding conducting structures using a finite element approach. A linear quadratic gaussian (LQG) control, or H 2 optimal control, algorithm based on the VALEN model for RWM feedback was developed for use with DIII-D's external control coil set. The algorithm is implemented on a platform that combines a graphics processing unit (GPU) for real-time control computation with low latency digital input/output control hardware and operates in parallel with the DIII-D Plasma Control System (PCS). Simulations and experiments showed that modern control techniques performed better, using 77% less current, than classical techniques when using coils external to

Developing snowflake divertor physics basis in the DIII-D, NSTX and NSTX-U tokamaks aimed at the divertor power exhaust solution [Snowflake divertor experiments in the DIII-D, NSTX and NSTX-U tokamaks aimed at the development of the divertor power exhaust solution

DOE PAGES

Soukhanovskii, V. A.; Allen, S. L.; Fenstermacher, M. E.; ...

2016-06-02

Experimental results from the National Spherical Torus Experiment (NSTX), a medium-size spherical tokamak with a compact divertor, and DIII-D, a large conventional aspect ratio tokamak, demonstrate that the snowflake (SF) divertor configuration may provide a promising solution for mitigating divertor heat loads and target plate erosion compatible with core H-mode confinement in future fusion devices, where the standard radiative divertor solution may be inadequate. In NSTX, where the initial high-power SF experiment were performed, the SF divertor was compatible with H-mode confinement, and led to the destabilization of large ELMs. However, a stable partial detachment of the outer strike pointmore » was also achieved where inter-ELM peak heat flux was reduced by factors 3-5, and peak ELM heat flux was reduced by up to 80% (cf. standard divertor). The DIII-D studies show the SF divertor enables significant power spreading in attached and radiative divertor conditions. Results include: compatibility with the core and pedestal, peak inter-ELM divertor heat flux reduction due to geometry at lower n e, and ELM energy and divertor peak heat flux reduction, especially prominent in radiative D 2-seeded SF divertor, and nearly complete power detachment and broader radiated power distribution in the radiative D 2-seeded SF divertor at P SOL = 3 - 4 MW. A variety of SF configurations can be supported by the divertor coil set in NSTX Upgrade. Edge transport modeling with the multi-fluid edge transport code UEDGE shows that the radiative SF divertor can successfully reduce peak divertor heat flux for the projected P SOL ≃9 MW case. In conclusion, the radiative SF divertor with carbon impurity provides a wider n e operating window, 50% less argon is needed in the impurity-seeded SF configuration to achieve similar q peak reduction factors (cf. standard divertor).« less

Nucleosome core particles containing a poly(dA.dT) sequence element exhibit a locally distorted DNA structure.

PubMed

Bao, Yunhe; White, Cindy L; Luger, Karolin

2006-08-25

Poly(dA.dT) DNA sequence elements are thought to promote transcription by either excluding nucleosomes or by altering their structural or dynamic properties. Here, the stability and structure of a defined nucleosome core particle containing a 16 base-pair poly(dA.dT) element (A16 NCP) was investigated. The A16 NCP requires a significantly higher temperature for histone octamer sliding in vitro compared to comparable nucleosomes that do not contain a poly(dA.dT) element. Fluorescence resonance energy transfer showed that the interactions between the nucleosomal DNA ends and the histone octamer were destabilized in A16 NCP. The crystal structure of A16 NCP was determined to a resolution of 3.2 A. The overall structure was maintained except for local deviations in DNA conformation. These results are consistent with previous in vivo and in vitro observations that poly(dA.dT) elements cause only modest changes in DNA accessibility and modest increases in steady-state transcription levels.

Inductive flux usage and its optimization in tokamak operation

DOE PAGES

Luce, Timothy C.; Humphreys, David A.; Jackson, Gary L.; ...

2014-07-30

The energy flow from the poloidal field coils of a tokamak to the electromagnetic and kinetic stored energy of the plasma are considered in the context of optimizing the operation of ITER. The goal is to optimize the flux usage in order to allow the longest possible burn in ITER at the desired conditions to meet the physics objectives (500 MW fusion power with energy gain of 10). A mathematical formulation of the energy flow is derived and applied to experiments in the DIII-D tokamak that simulate the ITER design shape and relevant normalized current and pressure. The rate ofmore » rise of the plasma current was varied, and the fastest stable current rise is found to be the optimum for flux usage in DIII-D. A method to project the results to ITER is formulated. The constraints of the ITER poloidal field coil set yield an optimum at ramp rates slower than the maximum stable rate for plasmas similar to the DIII-D plasmas. Finally, experiments in present-day tokamaks for further optimization of the current rise and validation of the projections are suggested.« less

High Confinement Mode and Edge Localized Mode Characteristics in a Near-Unity Aspect Ratio Tokamak.

PubMed

Thome, K E; Bongard, M W; Barr, J L; Bodner, G M; Burke, M G; Fonck, R J; Kriete, D M; Perry, J M; Schlossberg, D J

2016-04-29

Tokamak experiments at near-unity aspect ratio A≲1.2 offer new insights into the self-organized H-mode plasma confinement regime. In contrast to conventional A∼3 plasmas, the L-H power threshold P_{LH} is ∼15× higher than scaling predictions, and it is insensitive to magnetic topology, consistent with modeling. Edge localized mode (ELM) instabilities shift to lower toroidal mode numbers as A decreases. These ultralow-A operations enable heretofore inaccessible J_{edge}(R,t) measurements through an ELM that show a complex multimodal collapse and the ejection of a current-carrying filament.

High confinement mode and edge localized mode characteristics in a near-unity aspect ratio tokamak

DOE PAGES

Thome, Kathreen E.; Bongard, Michael W.; Barr, Jayson L.; ...

2016-04-27

Tokamak experiments at near-unity aspect ratio A ≲ 1.2 offer new insights into the self-organized H-mode plasma confinement regime. In contrast to conventional A ~ 3 plasmas, the L–H power threshold P LH is ~15× higher than scaling predictions, and it is insensitive to magnetic topology, consistent with modeling. Edge localized mode (ELM) instabilities shift to lower toroidal mode numbers as A decreases. Furthermore, these ultralow-A operations enable heretofore inaccessible J edge(R,t) measurements through an ELM that show a complex multimodal collapse and the ejection of a current-carrying filament.

A New Generation of Real-Time Systems in the JET Tokamak

NASA Astrophysics Data System (ADS)

Alves, Diogo; Neto, Andre C.; Valcarcel, Daniel F.; Felton, Robert; Lopez, Juan M.; Barbalace, Antonio; Boncagni, Luca; Card, Peter; De Tommasi, Gianmaria; Goodyear, Alex; Jachmich, Stefan; Lomas, Peter J.; Maviglia, Francesco; McCullen, Paul; Murari, Andrea; Rainford, Mark; Reux, Cedric; Rimini, Fernanda; Sartori, Filippo; Stephen, Adam V.; Vega, Jesus; Vitelli, Riccardo; Zabeo, Luca; Zastrow, Klaus-Dieter

2014-04-01

Recently, a new recipe for developing and deploying real-time systems has become increasingly adopted in the JET tokamak. Powered by the advent of x86 multi-core technology and the reliability of JET's well established Real-Time Data Network (RTDN) to handle all real-time I/O, an official Linux vanilla kernel has been demonstrated to be able to provide real-time performance to user-space applications that are required to meet stringent timing constraints. In particular, a careful rearrangement of the Interrupt ReQuests' (IRQs) affinities together with the kernel's CPU isolation mechanism allows one to obtain either soft or hard real-time behavior depending on the synchronization mechanism adopted. Finally, the Multithreaded Application Real-Time executor (MARTe) framework is used for building applications particularly optimised for exploring multi-core architectures. In the past year, four new systems based on this philosophy have been installed and are now part of JET's routine operation. The focus of the present work is on the configuration aspects that enable these new systems' real-time capability. Details are given about the common real-time configuration of these systems, followed by a brief description of each system together with results regarding their real-time performance. A cycle time jitter analysis of a user-space MARTe based application synchronizing over a network is also presented. The goal is to compare its deterministic performance while running on a vanilla and on a Messaging Real time Grid (MRG) Linux kernel.

Stainless steel blanket concept for tokamaks

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

Karbowski, J.S.; Lee, A.Y.; Prevenslik, T.V.

1979-01-25

The purpose of this joint ORNL/Westinghouse Program is to develop a design concept for a tokamak reactor blanket system which satisfies engineering requirements for a utility environment. While previous blanket studies have focused primarily on performance issues (thermal, neutronic, and structural), this study has emphasized consideration of reliability, fabricability, and lifetime.

Tokamak power reactor ignition and time dependent fractional power operation

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

Vold, E.L.; Mau, T.K.; Conn, R.W.

1986-06-01

A flexible time-dependent and zero-dimensional plasma burn code with radial profiles was developed and employed to study the fractional power operation and the thermal burn control options for an INTOR-sized tokamak reactor. The code includes alpha thermalization and a time-dependent transport loss which can be represented by any one of several currently popular scaling laws for energy confinement time. Ignition parameters were found to vary widely in density-temperature (n-T) space for the range of scaling laws examined. Critical ignition issues were found to include the extent of confinement time degradation by alpha heating, the ratio of ion to electron transportmore » power loss, and effect of auxiliary heating on confinement. Feedback control of the auxiliary power and ion fuel sources are shown to provide thermal stability near the ignition curve.« less

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.

Who will save the tokamak - Harry Potter, Arnold Schwarzenegger, or Shaquille O'Neil?

NASA Astrophysics Data System (ADS)

Freidberg, J.; Mangiarotti, F.; Minervini, J.

2014-10-01

The tokamak is the current leading contender for a fusion power reactor. The reason for the preeminence of the tokamak is its high quality plasma physics performance relative to other concepts. Even so, it is well known that the tokamak must still overcome two basic physics challenges before becoming viable as a DEMO and ultimately a reactor: (1) the achievement of non-inductive steady state operation, and (2) the achievement of robust disruption free operation. These are in addition to the PMI problems faced by all concepts. The work presented here demonstrates by means of a simple but highly credible analytic calculation that a ``standard'' tokamak cannot lead to a reactor - it is just not possible to simultaneously satisfy all the plasma physics plus engineering constraints. Three possible solutions, some more well-known than others, to the problem are analyzed. These visual image generating solutions are defined as (1) the Harry Potter solution, (2) the Arnold Schwarzenegger solution, and (3) the Shaquille O'Neil solution. Each solution will be described both qualitatively and quantitatively at the meeting.

Formation and evolution of quasi-interchange convection cell on the HL-2A tokamak

NASA Astrophysics Data System (ADS)

Shi, P. W.; Chen, W.; Shi, Z. B.; Duan, X. R.; Zhong, W. L.; Jiang, M.; Yang, Z. C.; Yu, L. M.; Wen, J.; Liang, A. S.; Yu, D. L.; Liu, Y.; Yang, Q. W.

2018-06-01

Formation and evolution of quasi-interchange convection cell have been observed for the first time by an electron cyclotron emission imaging (ECEI) system on the HL-2A tokamak. The instability with mode numbers of m/n = -1/-1 and a frequency of 2-4 kHz propagates in the electron diamagnetic drift direction. The mode downgrades the inner stored energy WE and enhances the divertor Hα signal, which indicates a degradation of plasma confinement. The ECEI reveals that the colder core temperature fluctuation distorts to a crescent shape, while a bubble originates from hot fluctuation gradually and draws into the concave side of the crescent. A finger-like structure can be observed during the convection process, and it presents a typical feature of the quasi-interchange model. The radial structure of the interchange mode obtained from electron cyclotron emission is quite narrow at the high field side but relatively wider at the low field side. Further analysis suggests that the discharge parameters are favorable for the excitation of interchange mode, and the stability criterion is violated in the core region. Moreover, the experimental frequency agrees with that given by the linear dispersion relation of interchange instability. The large grow rate suggests that the mode grows on a very short time scale.

Summary of Apollo; A D- sup 3 He tokamak reactor design

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

Kulcinski, G.L.; Blanchard, T.P.; El-Guebaly, L.A.

1992-07-01

In this paper, the key features of Apollo, a conceptual D-{sup 3}He tokamak reactor for commercial electricity production, are summarized. The 1000-MW (electric) design utilizes direct conversion of transport, neutron, and bremsstrahlung radiation power. The direct conversion method uses reactants, and the thermal conversion cycle uses an organic coolant. Apollo operates in the first-stability regime, with a major radius of 7.89 m, a peak magnetic field on the toroidal field coils of 19.3 T, a 53-MA plasma current, and a 6.7% beta value. The low neutron production of the D-{sup 3}He fuel cycle greatly reduces the radiation damage rate andmore » allows a full-lifetime first wall and structure made of standard steels with only slight modifications to reduce activation levels.« less

Suppression of runaway electrons with a resonant magnetic perturbation in MST tokamak plasmas

NASA Astrophysics Data System (ADS)

Munaretto, Stefano; Chapman, B. E.; Almagri, A. F.; Cornille, B. S.; Dubois, A. M.; Goetz, J. A.; McCollam, K. J.; Sovinec, C. R.

2016-10-01

Runaway electrons generated in MST tokamak plasmas are now being probed with resonant magnetic perturbations (RMP's). An RMP with m =3 strongly suppresses the runaway electrons. Initial modeling of these plasmas with NIMROD shows the degradation of flux surfaces with an m =3 RMP, which may account for the runaway electron suppression. These MST tokamak plasmas have Bt =0.14 T, Ip =50kA, and q(a) =2.2, with a bulk electron density and temperature of 5x1017 m-3 and 150 eV. Runaway electrons are detected via x-ray emission. The RMP is produced by a poloidal array of 32 saddle coils at the narrow vertical insulated cut in MST's thick conducting shell. Each RMP has a single m but a broad n spectrum. A sufficiently strong m =3 RMP completely suppresses the runaway electrons, while a comparable m =1 RMP has little effect. The impact of the RMP's on the magnetic topology of these plasmas is being studied with the nonlinear MHD code, NIMROD. With an m =3 RMP, stochasticity is introduced in the outer third of the plasma. No such change is observed with the m =1 RMP. NIMROD also predicts regularly occurring sawtooth oscillations with a period comparable to MHD activity observed in the experiment. Work supported by USDOE.

Development of DEMO-FNS tokamak for fusion and hybrid technologies

NASA Astrophysics Data System (ADS)

Kuteev, B. V.; Azizov, E. A.; Alexeev, P. N.; Ignatiev, V. V.; Subbotin, S. A.; Tsibulskiy, V. F.

2015-07-01

The history of fusion-fission hybrid systems based on a tokamak device as an extremely efficient DT-fusion neutron source has passed through several periods of ample research activity in the world since the very beginning of fusion research in the 1950s. Recently, a new roadmap of the hybrid program has been proposed with the goal to build a pilot hybrid plant (PHP) in Russia by 2030. Development of the DEMO-FNS tokamak for fusion and hybrid technologies, which is planned to be built by 2023, is the key milestone on the path to the PHP. This facility is in the phase of conceptual design aimed at providing feasibility studies for a full set of steady state tokamak technologies at a fusion energy gain factor Q ˜ 1, fusion power of ˜40 MW and opportunities for testing a wide range of hybrid technologies with the emphasis on continuous nuclide processing in molten salts. This paper describes the project motivations, its current status and the key issues of the design.

Integrated Tokamak modeling: When physics informs engineering and research planning

NASA Astrophysics Data System (ADS)

Poli, Francesca Maria

2018-05-01

Modeling tokamaks enables a deeper understanding of how to run and control our experiments and how to design stable and reliable reactors. We model tokamaks to understand the nonlinear dynamics of plasmas embedded in magnetic fields and contained by finite size, conducting structures, and the interplay between turbulence, magneto-hydrodynamic instabilities, and wave propagation. This tutorial guides through the components of a tokamak simulator, highlighting how high-fidelity simulations can guide the development of reduced models that can be used to understand how the dynamics at a small scale and short time scales affects macroscopic transport and global stability of plasmas. It discusses the important role that reduced models have in the modeling of an entire plasma discharge from startup to termination, the limits of these models, and how they can be improved. It discusses the important role that efficient workflows have in the coupling between codes, in the validation of models against experiments and in the verification of theoretical models. Finally, it reviews the status of integrated modeling and addresses the gaps and needs towards predictions of future devices and fusion reactors.

Cutoff Probe for Tokamak SOL Measurement

NASA Astrophysics Data System (ADS)

Na, Byung-Keun; You, Kwang-Ho; Kim, Dae-Woong; You, Shin-Jae; Kim, Jung-Hyung; Chang, Hong-Young

2013-09-01

Since a cutoff probe was developed, there have been a lot of improvements in methodology and analysis for low temperature plasmas. However, in order to apply the cutoff probe to the Tokamak scrape-off layer (SOL), three important issues should be solved - speed, thermal protection, and short-distance (a few mm) wave propagation in magnetized plasmas. In this presentation, the improvement of cutoff probe for Tokamak is shown. The above problems can be solved using the following methods: (a) the cutoff probe can be used with short impulse of a few nano-seconds for speed improvement. (b) Ceramic covers were used for thermal protection. (c) In magnetized plasmas, the cutoff peak can be analyzed using circuit modeling and CST Microwave studio. To verify the proposed methods, the cutoff probe was applied to a Helicon plasma, and the results were compared to laser Thomson scattering results. Based on the result in the Helicon plasma, the cutoff probe will be applied to far-SOL region at the KSTAR 2013 campaign, and SOL region at the later campaign.

Electron Cyclotron Radiation, Related Power Loss, and Passive Current Drive in Tokamaks: A Review

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

Fidone, Ignazio; Giruzzi, Gerardo; Granata, Giovanni

2001-01-15

A critical review on emission of weakly damped, high-harmonics electron cyclotron radiation, the related synchrotron power loss, and passive current drive in tokamaks with a fish-scale first wall is presented. First, the properties of overlapping harmonics are discussed using general analytical formulas and numerical applications. Next, the radiation power loss and efficiency of passive current drive in tokamak reactors are derived for the asymmetric fish-scale first wall. The radiation power loss is determined by the direction-averaged reflection coefficient {sigma}{sub 0} and the passive current drive by the differential reflectivity {delta}{sigma}/(1 - {sigma}{sub 0}). Finally, the problem of experimental investigations ofmore » the high harmonics radiation spectra, of {sigma}{sub 0} and {delta}{sigma}/(1 - {sigma}{sub 0}) in existing and next-step tokamaks, is discussed. Accurate measurements of the radiation spectra and the fish-scale reflectivity can be performed at arbitrary electron temperature using a partial fish-scale structure located near the tokamak equatorial plane.« less

Chaotic density fluctuations in L-mode plasmas of the DIII-D tokamak

DOE PAGES

Maggs, J. E.; Rhodes, Terry L.; Morales, G. J.

2015-03-05

Analysis of the time series obtained with the Doppler backscattering system (DBS) in the DIII-D tokamak shows that intermediate wave number plasma density fluctuations in low confinement (L-mode) tokamak plasmas are chaotic. Here, the supporting evidence is based on the shape of the power spectrum; the location of the signal in the complexity-entropy plane (C-H plane); and the population of the corresponding Bandt-Pompe probability distributions.

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.

Experimental pathways to understand and avoid high-Z impurity contamination from ICRF heating in tokamaks

NASA Astrophysics Data System (ADS)

Reinke, Matthew

2016-10-01

Recent results from Alcator C-Mod and JET demonstrate progress in understanding and mitigating core high-Z impurity contamination linked to ICRF heating in tokamaks with high-Z PFCs. Theory has identified two likely mechanisms: impurity sources due to sputtering enhanced by RF-rectified sheaths and greater cross-field SOL transport due to ExB convective cells. New experiments on Alcator C-Mod and JET demonstrate convective cell transport is likely a sub-dominant effect, despite directly observing ExB flows from rectified RF fields on C-Mod. Trace N2 introduced in the far SOL on field lines connected to and well away from an active ICRF antenna result in similar levels of core nitrogen, indicating local RF-driven transport is weak. This suggests the core high-Z density, nZ,core, is determined by sheath-induced sputtering and RF-independent SOL transport, allowing further reductions through antenna design. ICRF heating on C-Mod uses a unique, field aligned (FAA) and a pair of conventional, toroidally aligned (TAA) antennas. The FAA is designed to reduce rectified voltages relative to the TAA, and the impact of sheath-induced sputtering is explored by observing nZ,core while varying the TAA/FAA heating mix. A reduction of approximately 50% in core high-Z content is seen in L-modes when using the FAA and high-Z sources at the antenna limiter are effectively eliminated, indicating the remaining RF-driven source is away from the limiter. A drop in nZ,core may also be realized by locating the RF antenna on the inboard side where SOL transport aids impurity screening. New C-Mod experiments demonstrate up to a factor of 5 reduction in core nitrogen when N2 is injected on the high-field side as compared to low-field side impurity fueling. Varying the magnetic topology helps to elucidate the SOL transport physics responsible, laying a physics basis for inboard RF antenna placement. This work is supported by U.S. DOE Award DE-FC02-99ER54512, using Alcator C-Mod and carried out

Macroscopic erosion of divertor and first wall armour in future tokamaks

NASA Astrophysics Data System (ADS)

Würz, H.; Bazylev, B.; Landman, I.; Pestchanyi, S.; Safronov, V.

2002-12-01

Sputtering, evaporation and macroscopic erosion determine the lifetime of the 'in vessel' armour materials CFC, tungsten and beryllium presently under discussion for future tokamaks. For CFC armour macroscopic erosion means brittle destruction and dust formation whereas for metallic armour melt layer erosion by melt motion and droplet splashing. Available results on macroscopic erosion from hot plasma and e-beam simulation experiments and from tokamaks are critically evaluated and a comprehensive discussion of experimental and numerical macroscopic erosion and its extrapolation to future tokamaks is given. Shielding of divertor armour materials by their own vapor exists during plasma disruptions. The evolving plasma shield protects the armour from high heat loads, absorbs the incoming energy and reradiates it volumetrically thus reducing drastically the deposited energy. As a result, vertical target erosion by vaporization turns out to be of the order of a few microns per disruption event and macroscopic erosion becomes the dominant erosion source.

Lower Hybrid wave edge power loss quantification on the Alcator C-Mod tokamak

NASA Astrophysics Data System (ADS)

Faust, I. C.

2015-11-01

For the first time, the power deposition of Lower Hybrid RF waves into the edge plasma of a diverted tokamak has been systematically quantified. Edge deposition represents a parasitic loss of power that can greatly impact the use and efficiency of Lower Hybrid Current Drive (LHCD) at reactor-relevant densities. Through the use of a unique set of fast time resolution edge diagnostics, including innovative fast-thermocouples, an extensive set of Langmuir probes, and a Lyα ionization camera, the toroidal, poloidal and radial structure of the power deposition has been simultaneously determined. Power modulation was used to directly isolate the RF effects due to the prompt (t <τE) response of the scrape-off-layer (SOL) plasma to LHRF power. LHRF power was found to absorb more strongly in the edge at higher densities. It is found that a majority of this edge-deposited power is promptly conducted to the divertor. This correlates with the loss of current drive efficiency at high density previously observed on Alcator C-Mod, and displaying characteristics that contrast with the local RF edge absorption seen on other tokamaks. Measurements of ionization in the active divertor show dramatic changes due to LHRF power, implying that divertor region can be key for the LHRF edge power deposition physics. These observations support the existence a loss mechanism near the edge for LHRF at high density (ne > 1 . 0 .1020 [m-3]). Results will be shown addressing the distribution of power within the SOL, including the toroidal symmetry and radial distribution. These characteristics are important for deducing the cause of the reduced LHCD efficiency at high density and motivates the tailoring of wave propagation to minimize SOL interaction, for example, through the use of high-field-side launch. This work was performed on the Alcator C-Mod tokamak, a DoE Office of Science user facility, and is supported by USDoE award DE-FC02-99ER54512.

Preliminary measurements of neutrons from the D-D reaction in the COMPASS tokamak

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

Dankowski, J., E-mail: jan.dankowski@ifj.edu.pl; Kurowski, A.; Twarog, D.

Recent results of measured fast neutrons created in the D-D reaction on the COMPASS tokamak during ohmic discharges are presented in this paper. Two different type detectors were used during experiment. He-3 detectors and bubble detectors as a support. The measurements are an introduction for neutron diagnostic on tokamak COMPASS and monitoring neutrons during discharges with Neutral Beam Injection (NBI). The He-3 counters and bubble detectors were located in two positions near tokamak vacuum chamber at a distance less than 40 cm to the centre of plasma. The neutrons flux was observed in ohmic discharges. However, analysis of our resultsmore » does not indicate any clear source of neutrons production during ohmic discharges.« less

Constrained ripple optimization of Tokamak bundle divertors

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

Hively, L.M.; Rome, J.A.; Lynch, V.E.

1983-02-01

Magnetic field ripple from a tokamak bundle divertor is localized to a small toroidal sector and must be treated differently from the usual (distributed) toroidal field (TF) coil ripple. Generally, in a tokamak with an unoptimized divertor design, all of the banana-trapped fast ions are quickly lost due to banana drift diffusion or to trapping between the 1/R variation in absolute value vector B ..xi.. B and local field maxima due to the divertor. A computer code has been written to optimize automatically on-axis ripple subject to these constraints, while varying up to nine design parameters. Optimum configurations have lowmore » on-axis ripple (<0.2%) so that, now, most banana-trapped fast ions are confined. Only those ions with banana tips near the outside region (absolute value theta < or equal to 45/sup 0/) are lost. However, because finite-sized TF coils have not been used in this study, the flux bundle is not expanded.« less

Comparative Chromosome Painting and NOR Distribution Suggest a Complex Hybrid Origin of Triploid Lepidodactylus lugubris (Gekkonidae).

PubMed

Trifonov, Vladimir A; Paoletti, Alessio; Caputo Barucchi, Vincenzo; Kalinina, Tatiana; O'Brien, Patricia C M; Ferguson-Smith, Malcolm A; Giovannotti, Massimo

2015-01-01

Parthenogenesis, unisexuality and triploidy are interesting but poorly studied phenomena occurring in some reptile species. The mourning gecko (Lepidodactylus lugubris) represents a complex of diploid and triploid parthenogenetic mostly all-female populations (males occur quite rarely) widely distributed in coastal areas of the Indian and Pacific Oceans. Here, we study karyotypes of a male and two female L. lugubris (LLU) triploid individuals (3n = 66) using comparative painting with Gekko japonicus, Hemidactylus turcicus and H. platyurus chromosome specific probes to visualize the homologous regions and to reveal genus specific rearrangements. Also, we applied a 28S ribosomal DNA probe and Ag-staining to detect nucleolus organizer regions (NORs). Our results suggest that the karyotype of L. lugubris underwent a chromosome fission and a fusion after its divergence from a common ancestor of the Gekko-Hemidactylus group. The NORs were found to be located on one out of three homologs on each of LLU8, LLU15 and LLU18, thus further confirming a hybrid origin of triploid individuals. It seems that three different bisexual populations might have contributed to the origin of this triploid parthenogenetic population. We postulate that the heterozygosity in NOR localization is maintained in the triploid clone studied by the absence of recombination as described in whiptail lizards. The pattern of NOR localizations and homologous regions in males and females, as well as the absence of other detectable karyotypic differences, suggest that males arise spontaneously in all female populations and do not arise from independent hybridizations with different species.

Radial and poloidal correlation reflectometry on Experimental Advanced Superconducting Tokamak

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

Qu, Hao; Zhang, Tao; Han, Xiang

2015-08-15

An X-mode polarized V band (50 GHz–75 GHz) radial and poloidal correlation reflectometry is designed and installed on Experimental Advanced Superconducting Tokamak (EAST). Two frequency synthesizers (12 GHz–19 GHz) are used as sources. Signals from the sources are up-converted to V band using active quadruplers and then coupled together for launching through one single pyramidal antenna. Two poloidally separated antennae are installed to receive the reflected waves from plasma. This reflectometry system can be used for radial and poloidal correlation measurement of the electron density fluctuation. In ohmically heated plasma, the radial correlation length is about 1.5 cm measured bymore » the system. The poloidal correlation analysis provides a means to estimate the fluctuation velocity perpendicular to the main magnetic field. In the present paper, the distance between two poloidal probing points is calculated with ray-tracing code and the propagation time is deduced from cross-phase spectrum. Fluctuation velocity perpendicular to the main magnetic field in the core of ohmically heated plasma is about from −1 km/s to −3 km/s.« less

Electromagnetically driven westward drift and inner-core superrotation in Earth's core.

PubMed

Livermore, Philip W; Hollerbach, Rainer; Jackson, Andrew

2013-10-01

A 3D numerical model of the earth's core with a viscosity two orders of magnitude lower than the state of the art suggests a link between the observed westward drift of the magnetic field and superrotation of the inner core. In our model, the axial electromagnetic torque has a dominant influence only at the surface and in the deepest reaches of the core, where it respectively drives a broad westward flow rising to an axisymmetric equatorial jet and imparts an eastward-directed torque on the solid inner core. Subtle changes in the structure of the internal magnetic field may alter not just the magnitude but the direction of these torques. This not only suggests that the quasi-oscillatory nature of inner-core superrotation [Tkalčić H, Young M, Bodin T, Ngo S, Sambridge M (2013) The shuffling rotation of the earth's inner core revealed by earthquake doublets. Nat Geosci 6:497-502.] may be driven by decadal changes in the magnetic field, but further that historical periods in which the field exhibited eastward drift were contemporaneous with a westward inner-core rotation. The model further indicates a strong internal shear layer on the tangent cylinder that may be a source of torsional waves inside the core.

Physics and performance of the I-mode regime over an expanded operating space on Alcator C-Mod

NASA Astrophysics Data System (ADS)

Hubbard, A. E.; Baek, S.-G.; Brunner, D.; Creely, A. J.; Cziegler, I.; Edlund, E.; Hughes, J. W.; LaBombard, B.; Lin, Y.; Liu, Z.; Marmar, E. S.; Reinke, M. L.; Rice, J. E.; Sorbom, B.; Sung, C.; Terry, J.; Theiler, C.; Tolman, E. A.; Walk, J. R.; White, A. E.; Whyte, D.; Wolfe, S. M.; Wukitch, S.; Xu, X. Q.; the Alcator C-Mod Team

2017-12-01

New results on the I-mode regime of operation on the Alcator C-Mod tokamak are reported. This ELM-free regime features high energy confinement and a steep temperature pedestal, while particle confinement remains at L-mode levels, giving stationary density and avoiding impurity accumulation. I-mode has now been obtained over nearly all of the magnetic fields and currents possible in this high field tokamak (I p 0.55-1.7 MA, B T 2.8-8 T) using a configuration with B  ×  ∇ B drift away from the X-point. Results at 8 T confirm that the L-I power threshold varies only weakly with B T, and that the power range for I-mode increases with B T; no 8 T discharges transitioned to H-mode. Parameter dependences of energy confinement are investigated. Core transport simulations are giving insight into the observed turbulence reduction, profile stiffness and confinement improvement. Pedestal models explain the observed stability to ELMs, and can simulate the observed weakly coherent mode. Conditions for I-H transitions have complex dependences on density as well as power. I-modes have now been maintained in near-DN configurations, leading to improved divertor power flux sharing. Prospects for I-mode on future fusion devices such as ITER and ARC are encouraging. Further experiments on other tokamaks are needed to improve confidence in extrapolation.

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

Modeling of Resistive Wall Modes in Tokamak and Reversed Field Pinch Configurations of KTX

NASA Astrophysics Data System (ADS)

Han, Rui; Zhu, Ping; Bai, Wei; Lan, Tao; Liu, Wandong

2016-10-01

Resistive wall mode is believed to be one of the leading causes for macroscopic degradation of plasma confinement in tokamaks and reversed field pinches (RFP). In this study, we evaluate the linear RWM instability of Keda Torus eXperiment (KTX) in both tokamak and RFP configurations. For the tokamak configuration, the extended MHD code NIMROD is employed for calculating the dependence of the RWM growth rate on the position and conductivity of the vacuum wall for a model tokamak equilibrium of KTX in the large aspect-ratio approximation. For the RFP configuration, the standard formulation of dispersion relation for RWM based on the MHD energy principle has been evaluated for a cylindrical α- Θ model of KTX plasma equilibrium, in an effort to investigate the effects of thin wall on the RWM in KTX. Full MHD calculations of RWM in the RFP configuration of KTX using the NIMROD code are also being developed. Supported by National Magnetic Confinement Fusion Science Program of China Grant Nos. 2014GB124002, 2015GB101004, 2011GB106000, and 2011GB106003.

Optimization study of normal conductor tokamak for commercial neutron source

NASA Astrophysics Data System (ADS)

Fujita, T.; Sakai, R.; Okamoto, A.

2017-05-01

The optimum conceptual design of tokamak with normal conductor coils was studied for minimizing the cost for producing a given neutron flux by using a system code, PEC. It is assumed that the fusion neutrons are used for burning transuranics from the fission reactor spent fuel in the blanket and a fraction of the generated electric power is circulated to opearate the tokamak with moderate plasma fusion gain. The plasma performance was assumed to be moderate ones; {β\\text{N}}~∼ ~3{--}4 in the aspect ratio A~=~2{--}3 and {{H}98y2}~=~1 . The circulating power is an important factor affecting the cost. Though decreasing the aspect ratio is useful to raise the plasma beta and decrease the toroidal field, the maximum field in the coil starts to rise in the very low aspect ratio range and then the circulating power increases with decrease in the plasma aspect ratio A below A~∼ ~2 , while the construction cost increases with A . As a result, the cost per neutron has its minimum around A~∼ ~2.2 , namely, between ST and the conventional tokamak. The average circulating power fraction is expected to be ~51%.

Integrated Tokamak modeling: When physics informs engineering and research planning

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

Poli, Francesca Maria

Modeling tokamaks enables a deeper understanding of how to run and control our experiments and how to design stable and reliable reactors. We model tokamaks to understand the nonlinear dynamics of plasmas embedded in magnetic fields and contained by finite size, conducting structures, and the interplay between turbulence, magneto-hydrodynamic instabilities, and wave propagation. This tutorial guides through the components of a tokamak simulator, highlighting how high-fidelity simulations can guide the development of reduced models that can be used to understand how the dynamics at a small scale and short time scales affects macroscopic transport and global stability of plasmas. Itmore » discusses the important role that reduced models have in the modeling of an entire plasma discharge from startup to termination, the limits of these models, and how they can be improved. It discusses the important role that efficient workflows have in the coupling between codes, in the validation of models against experiments and in the verification of theoretical models. Finally, it reviews the status of integrated modeling and addresses the gaps and needs towards predictions of future devices and fusion reactors.« less

Integrated Tokamak modeling: When physics informs engineering and research planning

DOE PAGES

Poli, Francesca Maria

2018-05-01

Modeling tokamaks enables a deeper understanding of how to run and control our experiments and how to design stable and reliable reactors. We model tokamaks to understand the nonlinear dynamics of plasmas embedded in magnetic fields and contained by finite size, conducting structures, and the interplay between turbulence, magneto-hydrodynamic instabilities, and wave propagation. This tutorial guides through the components of a tokamak simulator, highlighting how high-fidelity simulations can guide the development of reduced models that can be used to understand how the dynamics at a small scale and short time scales affects macroscopic transport and global stability of plasmas. Itmore » discusses the important role that reduced models have in the modeling of an entire plasma discharge from startup to termination, the limits of these models, and how they can be improved. It discusses the important role that efficient workflows have in the coupling between codes, in the validation of models against experiments and in the verification of theoretical models. Finally, it reviews the status of integrated modeling and addresses the gaps and needs towards predictions of future devices and fusion reactors.« less

Simulation of EAST vertical displacement events by tokamak simulation code

NASA Astrophysics Data System (ADS)

Qiu, Qinglai; Xiao, Bingjia; Guo, Yong; Liu, Lei; Xing, Zhe; Humphreys, D. A.

2016-10-01

Vertical instability is a potentially serious hazard for elongated plasma. In this paper, the tokamak simulation code (TSC) is used to simulate vertical displacement events (VDE) on the experimental advanced superconducting tokamak (EAST). Key parameters from simulations, including plasma current, plasma shape and position, flux contours and magnetic measurements match experimental data well. The growth rates simulated by TSC are in good agreement with TokSys results. In addition to modeling the free drift, an EAST fast vertical control model enables TSC to simulate the course of VDE recovery. The trajectories of the plasma current center and control currents on internal coils (IC) fit experimental data well.

Plasma shaping effects on tokamak scrape-off layer turbulence

NASA Astrophysics Data System (ADS)

Riva, Fabio; Lanti, Emmanuel; Jolliet, Sébastien; Ricci, Paolo

2017-03-01

The impact of plasma shaping on tokamak scrape-off layer (SOL) turbulence is investigated. The drift-reduced Braginskii equations are written for arbitrary magnetic geometries, and an analytical equilibrium model is used to introduce the dependence of turbulence equations on tokamak inverse aspect ratio (ε ), Shafranov’s shift (Δ), elongation (κ), and triangularity (δ). A linear study of plasma shaping effects on the growth rate of resistive ballooning modes (RBMs) and resistive drift waves (RDWs) reveals that RBMs are strongly stabilized by elongation and negative triangularity, while RDWs are only slightly stabilized in non-circular magnetic geometries. Assuming that the linear instabilities saturate due to nonlinear local flattening of the plasma gradient, the equilibrium gradient pressure length {L}p=-{p}e/{{\

Effect of 3D magnetic perturbations on divertor conditions and detachment in tokamak and stellarator

DOE PAGES

Ahn, J. -W.; Briesemester, A. R.; Kobayashi, M.; ...

2017-06-22

Enhanced perpendicular heat and momentum transport induces parallel pressure loss leading to divertor detachment, which can be produced by the increase of density in 2D tokamaks. However, in the 3D configurations such as tokamaks with 3D fields and stellarators, the fraction of perpendicular transport can be higher even in a lower density regime, which could lead to the early transition to detachment without passing through the high-recycling regime. 3D fields applied to the limiter tokamak plasmas produce edge stochastic layers close to the last closed flux surface (LCFS), which can allow for enhanced perpendicular transport and indeed the absence ofmore » high recycling regime and early detachment have been observed in TEXTOR and Tore Supra. However, in the X-point divertor tokamaks with the applied 3D fields, the parallel transport is still dominant and the detachment facilitation has not been observed yet. Rather, 3D fields affected detachment adversely under certain conditions, either by preventing detachment onset as seen in DIII-D or by re-attaching the existing detached plasma as shown in NSTX. The possible way for strong 3D effects to induce access to the early detachment in divertor tokamaks appears to be via significant perpendicular loss of parallel momentum by frictional force for the counter-streaming flows between neighboring flow channels in the divertor. In principle, the adjacent lobes in the 3D divertor tokamak may generate the counter-streaming flow channels. However, an EMC3-EIRENE simulation for ITER H-mode plasmas demonstrated that screened RMP leads to significantly reduced counter-flows near the divertor target, therefore the momentum loss effect leading to detachment facilitation is expected to be small. This is consistent with the observation in LHD, which showed screening (amplification) of RMP fields in the attachment (stable detachment) case. In conclusion, work for optimal parameter window for best divertor operation scenario is

Effect of 3D magnetic perturbations on divertor conditions and detachment in tokamak and stellarator

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

Ahn, J. -W.; Briesemester, A. R.; Kobayashi, M.

Enhanced perpendicular heat and momentum transport induces parallel pressure loss leading to divertor detachment, which can be produced by the increase of density in 2D tokamaks. However, in the 3D configurations such as tokamaks with 3D fields and stellarators, the fraction of perpendicular transport can be higher even in a lower density regime, which could lead to the early transition to detachment without passing through the high-recycling regime. 3D fields applied to the limiter tokamak plasmas produce edge stochastic layers close to the last closed flux surface (LCFS), which can allow for enhanced perpendicular transport and indeed the absence ofmore » high recycling regime and early detachment have been observed in TEXTOR and Tore Supra. However, in the X-point divertor tokamaks with the applied 3D fields, the parallel transport is still dominant and the detachment facilitation has not been observed yet. Rather, 3D fields affected detachment adversely under certain conditions, either by preventing detachment onset as seen in DIII-D or by re-attaching the existing detached plasma as shown in NSTX. The possible way for strong 3D effects to induce access to the early detachment in divertor tokamaks appears to be via significant perpendicular loss of parallel momentum by frictional force for the counter-streaming flows between neighboring flow channels in the divertor. In principle, the adjacent lobes in the 3D divertor tokamak may generate the counter-streaming flow channels. However, an EMC3-EIRENE simulation for ITER H-mode plasmas demonstrated that screened RMP leads to significantly reduced counter-flows near the divertor target, therefore the momentum loss effect leading to detachment facilitation is expected to be small. This is consistent with the observation in LHD, which showed screening (amplification) of RMP fields in the attachment (stable detachment) case. In conclusion, work for optimal parameter window for best divertor operation scenario is

Plasma rotation and transport in MAST spherical tokamak

NASA Astrophysics Data System (ADS)

Field, A. R.; Michael, C.; Akers, R. J.; Candy, J.; Colyer, G.; Guttenfelder, W.; Ghim, Y.-c.; Roach, C. M.; Saarelma, S.; MAST Team

2011-06-01

The formation of internal transport barriers (ITBs) is investigated in MAST spherical tokamak plasmas. The relative importance of equilibrium flow shear and magnetic shear in their formation and evolution is investigated using data from high-resolution kinetic- and q-profile diagnostics. In L-mode plasmas, with co-current directed NBI heating, ITBs in the momentum and ion thermal channels form in the negative shear region just inside qmin. In the ITB region the anomalous ion thermal transport is suppressed, with ion thermal transport close to the neo-classical level, although the electron transport remains anomalous. Linear stability analysis with the gyro-kinetic code GS2 shows that all electrostatic micro-instabilities are stable in the negative magnetic shear region in the core, both with and without flow shear. Outside the ITB, in the region of positive magnetic shear and relatively weak flow shear, electrostatic micro-instabilities become unstable over a wide range of wave numbers. Flow shear reduces the linear growth rates of low-k modes but suppression of ITG modes is incomplete, which is consistent with the observed anomalous ion transport in this region; however, flow shear has little impact on growth rates of high-k, electron-scale modes. With counter-NBI ITBs of greater radial extent form outside qmin due to the broader profile of E × B flow shear produced by the greater prompt fast-ion loss torque.

Developing physics basis for the snowflake divertor in the DIII-D tokamak

DOE PAGES

Soukhanovskii, V. A.; Allen, S. L.; Fenstermacher, M. E.; ...

2018-02-01

Recent DIII-D results demonstrate that the snowflake (SF) divertor geometry (cf. standard divertor) enables significant manipulation of divertor heat transport for heat spreading and reduction in attached and radiative divertor regimes, between and during edge localized modes (ELMs), while maintaining good H-mode confinement. Snowflake divertor configurations have been realized in the DIII-D tokamak for several seconds in H-mode discharges with heating power PNBImore » $$\\leqslant$$ 4-5 MW and a range of plasma currents Ip = 0.8-1.2 MA. In this work, inter-ELM transport and radiative SF divertor properties are studied. Significant impact of geometric properties on SOL and divertor plasma parameters, including increased poloidal magnetic flux expansion, divertor magnetic field line length and divertor volume, is confirmed. In the SF-minus configuration, heat deposition is affected by the geometry, and peak divertor heat fluxes are significantly reduced. In the SF-plus and near-exact SF configurations, divertor peak heat flux reduction and outer strike point heat flux profile broadening are observed. Inter-ELM sharing of power and particle fluxes between the main and additional snowflake divertor strike points has been demonstrated. The additional strike points typically receive up to 10-15% of total outer divertor power. Measurements of electron pressure and poloidal beta !p support the theoretically proposed churning mode that is driven by toroidal curvature and vertical pressure gradient in the weak poloidal field region. A comparison of the 4-4.5 MW NBI-heated H-mode plasmas with radiative SF divertor and the standard radiative divertor (both induced with additional gas puffing) shows a nearly complete power detachment and broader divertor radiated power distribution in the SF, as compared to a partial detachment and peaked localized radiation in the standard divertor. However, insignificant difference in the detachment onset w.r.t. density between the SF and

Developing physics basis for the snowflake divertor in the DIII-D tokamak

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

Soukhanovskii, V. A.; Allen, S. L.; Fenstermacher, M. E.

Recent DIII-D results demonstrate that the snowflake (SF) divertor geometry (cf. standard divertor) enables significant manipulation of divertor heat transport for heat spreading and reduction in attached and radiative divertor regimes, between and during edge localized modes (ELMs), while maintaining good H-mode confinement. Snowflake divertor configurations have been realized in the DIII-D tokamak for several seconds in H-mode discharges with heating power PNBImore » $$\\leqslant$$ 4-5 MW and a range of plasma currents Ip = 0.8-1.2 MA. In this work, inter-ELM transport and radiative SF divertor properties are studied. Significant impact of geometric properties on SOL and divertor plasma parameters, including increased poloidal magnetic flux expansion, divertor magnetic field line length and divertor volume, is confirmed. In the SF-minus configuration, heat deposition is affected by the geometry, and peak divertor heat fluxes are significantly reduced. In the SF-plus and near-exact SF configurations, divertor peak heat flux reduction and outer strike point heat flux profile broadening are observed. Inter-ELM sharing of power and particle fluxes between the main and additional snowflake divertor strike points has been demonstrated. The additional strike points typically receive up to 10-15% of total outer divertor power. Measurements of electron pressure and poloidal beta !p support the theoretically proposed churning mode that is driven by toroidal curvature and vertical pressure gradient in the weak poloidal field region. A comparison of the 4-4.5 MW NBI-heated H-mode plasmas with radiative SF divertor and the standard radiative divertor (both induced with additional gas puffing) shows a nearly complete power detachment and broader divertor radiated power distribution in the SF, as compared to a partial detachment and peaked localized radiation in the standard divertor. However, insignificant difference in the detachment onset w.r.t. density between the SF and

Synergistic cross-scale coupling of turbulence in a tokamak plasma

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

Howard, N. T., E-mail: nthoward@psfc.mit.edu; Holland, C.; White, A. E.

2014-11-15

For the first time, nonlinear gyrokinetic simulations spanning both the ion and electron spatio-temporal scales have been performed with realistic electron mass ratio ((m{sub D}∕m{sub e}){sup 1∕2 }= 60.0), realistic geometry, and all experimental inputs, demonstrating the coexistence and synergy of ion (k{sub θ}ρ{sub s}∼O(1.0)) and electron-scale (k{sub θ}ρ{sub e}∼O(1.0)) turbulence in the core of a tokamak plasma. All multi-scale simulations utilized the GYRO code [J. Candy and R. E. Waltz, J. Comput. Phys. 186, 545 (2003)] to study the coupling of ion and electron-scale turbulence in the core (r/a = 0.6) of an Alcator C-Mod L-mode discharge shown previously to exhibit an under-predictionmore » of the electron heat flux when using simulations only including ion-scale turbulence. Electron-scale turbulence is found to play a dominant role in setting the electron heat flux level and radially elongated (k{sub r} ≪ k{sub θ}) “streamers” are found to coexist with ion-scale eddies in experimental plasma conditions. Inclusion of electron-scale turbulence in these simulations is found to increase both ion and electron heat flux levels by enhancing the transport at the ion-scale while also driving electron heat flux at sub-ρ{sub i} scales. The combined increases in the low and high-k driven electron heat flux may explain previously observed discrepancies between simulated and experimental electron heat fluxes and indicates a complex interaction of short and long wavelength turbulence.« less

Confinement properties of tokamak plasmas with extended regions of low magnetic shear

NASA Astrophysics Data System (ADS)

Graves, J. P.; Cooper, W. A.; Kleiner, A.; Raghunathan, M.; Neto, E.; Nicolas, T.; Lanthaler, S.; Patten, H.; Pfefferle, D.; Brunetti, D.; Lutjens, H.

2017-10-01

Extended regions of low magnetic shear can be advantageous to tokamak plasmas. But the core and edge can be susceptible to non-resonant ideal fluctuations due to the weakened restoring force associated with magnetic field line bending. This contribution shows how saturated non-linear phenomenology, such as 1 / 1 Long Lived Modes, and Edge Harmonic Oscillations associated with QH-modes, can be modelled accurately using the non-linear stability code XTOR, the free boundary 3D equilibrium code VMEC, and non-linear analytic theory. That the equilibrium approach is valid is particularly valuable because it enables advanced particle confinement studies to be undertaken in the ordinarily difficult environment of strongly 3D magnetic fields. The VENUS-LEVIS code exploits the Fourier description of the VMEC equilibrium fields, such that full Lorenzian and guiding centre approximated differential operators in curvilinear angular coordinates can be evaluated analytically. Consequently, the confinement properties of minority ions such as energetic particles and high Z impurities can be calculated accurately over slowing down timescales in experimentally relevant 3D plasmas.

Tungsten coating by ATC plasma spraying on CFC for WEST tokamak

NASA Astrophysics Data System (ADS)

Firdaouss, M.; Desgranges, C.; Hernandez, C.; Mateus, C.; Maier, H.; Böswirth, B.; Greuner, H.; Samaille, F.; Bucalossi, J.; Missirlian, M.

2017-12-01

In the field of fusion experiments using a tokamak, the plasma facing components (PFC) are the closest object to the hot plasma. Due to the plasma-wall interaction, the material composing the PFC may enter the plasma and disturb the experiments. In the past, the main material for PFC was carbon (CFC, graphite), while the future reactors like ITER will be fully metallic, in particular tungsten. The Tore Supra tokamak has been transformed in an x-point divertor fusion device within the frame of the WEST (W (tungsten) Environment in Steady-state Tokamak) project in order to have plasma conditions close to those expected in ITER. The PFC other than the divertor has been coated with W to transform Tore Supra into a fully metallic environment. Different coating techniques have been selected for different kind of PFC. This paper gives an overview on the coating process used for the antennae protection limiter, the associated validation programme and concludes on the adequacy of the W coating with the WEST experimental programme requirements and gives perspectives on the development to be pursued.

Interaction of external n  =  1 magnetic fields with the sawtooth instability in low-q RFX-mod and DIII-D tokamaks

NASA Astrophysics Data System (ADS)

Piron, C.; Martin, P.; Bonfiglio, D.; Hanson, J.; Logan, N. C.; Paz-Soldan, C.; Piovesan, P.; Turco, F.; Bialek, J.; Franz, P.; Jackson, G.; Lanctot, M. J.; Navratil, G. A.; Okabayashi, M.; Strait, E.; Terranova, D.; Turnbull, A.

2016-10-01

External n  =  1 magnetic fields are applied in RFX-mod and DIII-D low safety factor Tokamak plasmas to investigate their interaction with the internal MHD dynamics and in particular with the sawtooth instability. In these experiments the applied magnetic fields cause a reduction of both the sawtooth amplitude and period, leading to an overall stabilizing effect on the oscillations. In RFX-mod sawteeth eventually disappear and are replaced by a stationary m  =  1, n  =  1 helical equilibrium without an increase in disruptivity. However toroidal rotation is significantly reduced in these plasmas, thus it is likely that the sawtooth mitigation in these experiments is due to the combination of the helically deformed core and the reduced rotation. The former effect is qualitatively well reproduced by nonlinear MHD simulations performed with the PIXIE3D code. The results obtained in these RFX-mod experiments motivated similar ones in DIII-D L-mode diverted Tokamak plasmas at low q 95. These experiments succeeded in reproducing the sawtooth mitigation with the approach developed in RFX-mod. In DIII-D this effect is correlated with a clear increase of the n  =  1 plasma response, that indicates an enhancement of the coupling to the marginally stable n  =  1 external kink, as simulations with the linear MHD code IPEC suggest. A significant rotation braking in the plasma core is also observed in DIII-D. Numerical calculations of the neoclassical toroidal viscosity (NTV) carried out with PENT identify this torque as a possible contributor for this effect.

Stationary Flowing Liquid Lithium (SFLiLi) systems for tokamaks

NASA Astrophysics Data System (ADS)

Zakharov, Leonid; Gentile, Charles; Roquemore, Lane

2013-10-01

The present approach to magnetic fusion which relies on high recycling plasma-wall interaction has exhausted itself at the level of TFTR, JET, JT-60 devices with no realistic path to the burning plasma. Instead, magnetic fusion needs a return to its original idea of insulation of the plasma from the wall, which was the dominant approach in the 1970s and upon implementations has a clear path to the DEMO device with PDT ~= 100 MW and Qelectric > 1 . The SFLiLi systems of this talk is the technology tool for implementation of the guiding idea of magnetic fusion. It utilizes the unique properties of flowing LiLi to pump plasma particles and, thus, insulate plasma from the walls. The necessary flow rate, ~= 1 g3/s, is very small, thus, making the use of lithium practical and consistent with safety requirements. The talk describes how chemical activity of LiLi, which is the major technology challenge of using LiLi in tokamaks, is addressed by SFLiLi systems at the level of already performed (HT-7) experiment, and in ongoing implementations for a prototype of SFLiLi for tokamak divertors and the mid-plane limiter for EAST tokamak (to be tested in the next experimental campaign). This work is supported by US DoE contract No. DE-AC02-09-CH11466.

Characterization of Fe3O4/SiO2/Gd2O(CO3)2 core/shell/shell nanoparticles as T1 and T2 dual mode MRI contrast agent.

PubMed

Yang, Meicheng; Gao, Lipeng; Liu, Kai; Luo, Chunhua; Wang, Yiting; Yu, Lei; Peng, Hui; Zhang, Wen

2015-01-01

Core/shell/shell structured Fe3O4/SiO2/Gd2O(CO3)2 nanoparticles were successfully synthesized. Their properties as a new type of T1-T2 dual model contrast agent for magnetic resonance imaging were investigated. Due to the introduce of a separating SiO2 layer, the magnetic coupling between Gd2O(CO3)2 and Fe3O4 could be modulated by the thickness of SiO2 layer and produce appropriate T1 and T2 signal. Additionally, the existence of Gd(3+) enhances the transverse relaxivity of Fe3O4 possibly because of the magnetic coupling between Gd(3+) and Fe3O4. The Fe3O4/SiO2/Gd2O(CO3)2 nanoparticles exhibit good biocompatibility, showing great potential for biomedical applications. Copyright © 2014 Elsevier B.V. All rights reserved.

Isotope effects on L-H threshold and confinement in tokamak plasmas

NASA Astrophysics Data System (ADS)

Maggi, C. F.; Weisen, H.; Hillesheim, J. C.; Chankin, A.; Delabie, E.; Horvath, L.; Auriemma, F.; Carvalho, I. S.; Corrigan, G.; Flanagan, J.; Garzotti, L.; Keeling, D.; King, D.; Lerche, E.; Lorenzini, R.; Maslov, M.; Menmuir, S.; Saarelma, S.; Sips, A. C. C.; Solano, E. R.; Belonohy, E.; Casson, F. J.; Challis, C.; Giroud, C.; Parail, V.; Silva, C.; Valisa, M.; Contributors, JET

2018-01-01

The dependence of plasma transport and confinement on the main hydrogenic ion isotope mass is of fundamental importance for understanding turbulent transport and, therefore, for accurate extrapolations of confinement from present tokamak experiments, which typically use a single hydrogen isotope, to burning plasmas such as ITER, which will operate in deuterium-tritium mixtures. Knowledge of the dependence of plasma properties and edge transport barrier formation on main ion species is critical in view of the initial, low-activation phase of ITER operations in hydrogen or helium and of its implications on the subsequent operation in deuterium-tritium. The favourable scaling of global energy confinement time with isotope mass, which has been observed in many tokamak experiments, remains largely unexplained theoretically. Moreover, the mass scaling observed in experiments varies depending on the plasma edge conditions. In preparation for upcoming deuterium-tritium experiments in the JET tokamak with the ITER-like Be/W Wall (JET-ILW), a thorough experimental investigation of isotope effects in hydrogen, deuterium and tritium plasmas is being carried out, in order to provide stringent tests of plasma energy, particle and momentum transport models. Recent hydrogen and deuterium isotope experiments in JET-ILW on L-H power threshold, L-mode and H-mode confinement are reviewed and discussed in the context of past and more recent isotope experiments in tokamak plasmas, highlighting common elements as well as contrasting observations that have been reported. The experimental findings are discussed in the context of fundamental aspects of plasma transport models.

H-mode achievement and edge features in RFX-mod tokamak operation

NASA Astrophysics Data System (ADS)

Spolaore, M.; Cavazzana, R.; Marrelli, L.; Carraro, L.; Franz, P.; Spagnolo, S.; Zaniol, B.; Zuin, M.; Cordaro, L.; Dal Bello, S.; De Masi, G.; Ferro, A.; Finotti, C.; Grando, L.; Grenfell, G.; Innocente, P.; Kudlacek, O.; Marchiori, G.; Martines, E.; Momo, B.; Paccagnella, R.; Piovesan, P.; Piron, C.; Puiatti, M. E.; Recchia, M.; Scarin, P.; Taliercio, C.; Vianello, N.; Zanotto, L.

2017-11-01

The RFX-mod experiment is a fusion device designed to operate as a reversed field pinch (RFP), with a major radius R = 2 m and a minor radius a = 0.459 m. Its high versatility recently allowed operating it also as an ohmic tokamak, allowing comparative studies between the two configurations in the same device. The device is equipped with a state of the art MHD mode feedback control system providing a magnetic boundary effective control, by applying resonant or non-resonant magnetic perturbations (MP), both in RFP and in tokamak configurations. In the fusion community the application of MPs is widely studied as a promising tool to limit the impact of plasma filaments and ELMs (edge localized modes) on plasma facing components. An important new research line is the exploitation of the RFX-mod active control system for ELM mitigation studies. As a first step in this direction, this paper presents the most recent achievements in term of RFX-mod tokamak explored scenarios, which allowed the first investigation of the ohmic and edge biasing induced H-mode. The production of D-shaped tokamak discharges and the design and deployment of an insertable polarized electrode were accomplished. Reproducible H-mode phases were obtained with insertable electrode negative biasing in single null discharges, representing an unexplored scenario with this technique. Important modifications of the edge plasma density and flow properties are observed. During the achieved H-mode ELM-like electromagnetic composite filamentary structures are observed. They are characterized by clear vorticity and parallel current density patterns.

Plasma density injection and flow during coaxial helicity injection in a tokamak

NASA Astrophysics Data System (ADS)

Hooper, E. B.

2018-02-01

Whole device, resistive MHD simulations of spheromaks and tokamaks have used a large diffusion coefficient that maintains a nearly constant density throughout the device. In the present work, helicity and plasma are coinjected into a low-density plasma in a tokamak with a small diffusion coefficient. As in previous simulations [Hooper et al., Phys. Plasmas 20, 092510 (2013)], a flux bubble is formed, which expands to fill the tokamak volume. The injected plasma is non-uniform inside the bubble. The flow pattern is analyzed; when the simulation is not axisymmetric, an n = 1 mode on the surface of the bubble generates leakage of plasma into the low-density volume. Closed flux is generated following injection, as in experiments and previous simulations. The result provides a more detailed physics analysis of the injection, including density non-uniformities in the plasma that may affect its use as a startup plasma [Raman et al., Phys. Rev. Lett. 97, 175002 (2006)].

Sensitivity of magnetic field-line pitch angle measurements to sawtooth events in tokamaks

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

Ko, J., E-mail: jinseok@nfri.re.kr

2016-11-15

The sensitivity of the pitch angle profiles measured by the motional Stark effect (MSE) diagnostic to the evolution of the safety factor, q, profiles during the tokamak sawtooth events has been investigated for Korea Superconducting Tokamak Advanced Research (KSTAR). An analytic relation between the tokamak pitch angle, γ, and q estimates that Δγ ∼ 0.1° is required for detecting Δq ∼ 0.05 near the magnetic axis (not at the magnetic axis, though). The pitch angle becomes less sensitive to the same Δq for the middle and outer regions of the plasma (Δγ ∼ 0.5°). At the magnetic axis, it ismore » not straightforward to directly relate the γ sensitivity to Δq since the gradient of γ(R), where R is the major radius of the tokamak, is involved. Many of the MSE data obtained from the 2015 KSTAR campaign, when calibrated carefully, can meet these requirements with the time integration down to 10 ms. The analysis with the measured data shows that the pitch angle profiles and their gradients near the magnetic axis can resolve the change of the q profiles including the central safety factor, q{sub 0}, during the sawtooth events.« less

Electromagnetically driven westward drift and inner-core superrotation in Earth’s core

PubMed Central

Livermore, Philip W.; Hollerbach, Rainer; Jackson, Andrew

2013-01-01

A 3D numerical model of the earth’s core with a viscosity two orders of magnitude lower than the state of the art suggests a link between the observed westward drift of the magnetic field and superrotation of the inner core. In our model, the axial electromagnetic torque has a dominant influence only at the surface and in the deepest reaches of the core, where it respectively drives a broad westward flow rising to an axisymmetric equatorial jet and imparts an eastward-directed torque on the solid inner core. Subtle changes in the structure of the internal magnetic field may alter not just the magnitude but the direction of these torques. This not only suggests that the quasi-oscillatory nature of inner-core superrotation [Tkalčić H, Young M, Bodin T, Ngo S, Sambridge M (2013) The shuffling rotation of the earth’s inner core revealed by earthquake doublets. Nat Geosci 6:497–502.] may be driven by decadal changes in the magnetic field, but further that historical periods in which the field exhibited eastward drift were contemporaneous with a westward inner-core rotation. The model further indicates a strong internal shear layer on the tangent cylinder that may be a source of torsional waves inside the core. PMID:24043841

ADX: A high Power Density, Advanced RF-Driven Divertor Test Tokamak for PMI studies

NASA Astrophysics Data System (ADS)

Whyte, Dennis; ADX Team

2015-11-01

The MIT PSFC and collaborators are proposing an advanced divertor experiment, ADX; a divertor test tokamak dedicated to address critical gaps in plasma-material interactions (PMI) science, and the world fusion research program, on the pathway to FNSF/DEMO. Basic ADX design features are motivated and discussed. In order to assess the widest range of advanced divertor concepts, a large fraction (>50%) of the toroidal field volume is purpose-built with innovative magnetic topology control and flexibility for assessing different surfaces, including liquids. ADX features high B-field (>6 Tesla) and high global power density (P/S ~ 1.5 MW/m2) in order to access the full range of parallel heat flux and divertor plasma pressures foreseen for reactors, while simultaneously assessing the effect of highly dissipative divertors on core plasma/pedestal. Various options for efficiently achieving high field are being assessed including the use of Alcator technology (cryogenic cooled copper) and high-temperature superconductors. The experimental platform would also explore advanced lower hybrid current drive and ion-cyclotron range of frequency actuators located at the high-field side; a location which is predicted to greatly reduce the PMI effects on the launcher while minimally perturbing the core plasma. The synergistic effects of high-field launchers with high total B on current and flow drive can thus be studied in reactor-relevant boundary plasmas.

Core Composition and the Magnetic Field of Mercury

NASA Astrophysics Data System (ADS)

Spohn, T.; Breuer, D.

2005-05-01

The density of Mercury suggests a core of approximately 1800 km radius and a mantle of approximately 600 km thickness. Convection in the mantle is often claimed to be capable of freezing the core over the lifetime of the solar system if the core is nearly pure iron. The thermal history calculations of Stevenson et al. (1983) and Schubert et al. (1988) suggest that about 5 weight-% sulphur are required to lower the core liquidus sufficiently to prevent complete freezing of the core and maintain a significant fluid outer core shell. Other candidates for a light alloying element require similarly large concentrations. The requirement of a significant concentration of volatile elements in the core is likely to be at variance with cosmochemical arguments for a mostly refractory, volatile poor composition of the planet. We have re-addressed the question of the freezing of Mercury's core using parameterized convection models based on the stagnant lid theory of planetary mantle convection. We have compared these results to earlier calculations (Conzelmann and Spohn, 1999) of Hermian mantle convection using a finite-amplitude convection code. We find consistently that the stagnant lid tends to thermally insulate the deep interior and we find mantle and core temperatures significantly larger than those calculated by Stevenson et al. (1983) and Schubert et al. (1988). As a consequence we find fluid outer core shells for reasonable mantle rheology parameters even for compositions with as little as 0.1 weight-% sulphur. Stevenson, D.J., T. Spohn, and G. Schubert. Icarus, 54, 466, 1983. Schubert, G. M.N. Ross, D.J. Stevenson, and T. Spohn, in Mercury, F. Vilas, C.R. Chapman and M.S. Matthews, eds., p.429, 1988. Conzelmann, V. and T. Spohn, Bull. Am. Astr. Soc., 31, 1102, 1999.

Common Core Literacy Lesson Plans: Ready-to-Use Resources, 9-12

ERIC Educational Resources Information Center

Davis, Lauren

2013-01-01

Schools nationwide are transitioning to the Common Core--our advice to you: Be prepared, but don't go it alone! Our new book, "Common Core Literacy Lesson Plans: Ready-to-Use Resources, 9-12," shows you that teaching the Common Core State Standards in high school doesn't have to be intimidating! This easy-to-use guide meets the particular needs of…

Common Core Literacy Lesson Plans: Ready-to-Use Resources, 6-8

ERIC Educational Resources Information Center

Davis, Lauren

2013-01-01

Schools nationwide are transitioning to the Common Core--our advice to you: Don't go it alone! Our new book, "Common Core Literacy Lesson Plans: Ready-to-Use Resources, 6-8," shows you that teaching the Common Core State Standards in the middle grades doesn't have to be intimidating! This easy-to-use guide meets the particular needs of middle…

Common Core Literacy Lesson Plans: Ready-to-Use Resources, K-5

ERIC Educational Resources Information Center

Davis, Lauren

2013-01-01

Schools nationwide are transitioning to the Common Core--our advice to you: Don't go it alone! Our new book, "Common Core Literacy Lesson Plans: Ready-to-Use Resources, K-5," shows you that teaching the Common Core State Standards in the elementary grades doesn't have to be intimidating! This easy-to-use guide provides model lesson plans for…

The computation in diagnostics for tokamaks: systems, designs, approaches

NASA Astrophysics Data System (ADS)

Krawczyk, Rafał; Linczuk, Paweł; Czarski, Tomasz; Wojeński, Andrzej; Chernyshova, Maryna; Poźniak, Krzysztof; Kolasiński, Piotr; Kasprowicz, Grzegorz; Zabołotny, Wojciech; Kowalska-Strzeciwilk, Ewa; Malinowski, Karol; Gaska, Michał

2017-08-01

The requirements given for GEM (Gaseous Electron Multiplier) detector based acquisition system for plasma impurities diagnostics triggered a need for the development of a specialized software and hardware architecture. The amount of computations with latency and throughput restrictions cause that an advanced solution is sought for. In order to provide a mechanism fitting the designated tokamaks, an insight into existing solutions was necessary. In the article there is discussed architecture of systems used for plasma diagnostics and in related scientific fields. The developed solution is compared and contrasted with other diagnostic and control systems. Particular attention is payed to specific requirements for plasma impurities diagnostics in tokamak thermal fusion reactor. Subsequently, the details are presented that justified the choice of the system architecture and the discussion on various approaches is given.

Neoclassical theory inside transport barriers in tokamaks

NASA Astrophysics Data System (ADS)

Shaing, K. C.; Hsu, C. T.

2012-02-01

Inside the transport barriers in tokamaks, ion energy losses sometimes are smaller than the value predicted by the standard neoclassical theory. This improvement can be understood in terms of the orbit squeezing theory in addition to the sonic poloidal E ×B Mach number Up,m that pushes the tips of the trapped particles to the higher energy. In general, Up,m also includes the poloidal component of the parallel mass flow speed. These physics mechanisms are the corner stones for the transition theory of the low confinement mode (L-mode) to the high confinement mode (H-mode) in tokamaks. Here, detailed transport fluxes in the banana regime are presented using the parallel viscous forces calculated earlier. It is found, as expected, that effects of orbit squeezing and the sonic Up,m reduce the ion heat conductivity. The former reduces it by a factor of |S|3/2 and the later by a factor of R(Up ,m2)exp(-Up ,m2) with R(Up ,m2), a rational function. Here, S is the orbit squeezing factor.

Recent Doppler Backscattering results from EAST tokamak

NASA Astrophysics Data System (ADS)

Zhou, Chu; Liu, Adi; Zhang, Xiaohui; Hu, Jianqiang; Wang, Mingyuan; Yu, Changxuan; Liu, Wandong; Li, Hong; Lan, Tao; Sun, Xuan; Xie, Jinlin; Ding, Weixing; CAS Key Laboratory of Geospace Environment, University of Science and Technology of China Team; Department of Physics and Astronomy, University of California at Los Angeles Collaboration

2013-10-01

A Doppler reflectometer system has recently been installed in the EAST tokamak. It includes two separated systems, one for Q-band and the other for V-band. The optical system consists of a fixed flat mirror and a steerable parabolic mirror, which enabling the measurement of perpendicular wave number in the range of 4-22/cm, with the wave number resolution around 2/cm, while the radial location can cover the whole minor radius for L mode and the whole pedestal for H mode on EAST. A 2D Gaussion Ray tracing code is used to calculate the scattering location, the perpendicular wave number and the resolution. In EAST last experimental campaign the Doppler shifted signals have been obtained and the radial profiles of the perpendicular propagation velocity during L-mode and H-mode are calculated. The Er evolution during L-H and H-L transition have also been measured. The two separated systems are also used as a poloidal coherent system together to study the GAM in EAST tokamak.

Magnetic diagnostics for the lithium tokamak experiment.

PubMed

Berzak, L; Kaita, R; Kozub, T; Majeski, R; Zakharov, L

2008-10-01

The lithium tokamak experiment (LTX) is a spherical tokamak with R(0)=0.4 m, a=0.26 m, B(TF) approximately 3.4 kG, I(P) approximately 400 kA, and pulse length approximately 0.25 s. The focus of LTX is to investigate the novel low-recycling lithium wall operating regime for magnetically confined plasmas. This regime is reached by placing an in-vessel shell conformal to the plasma last closed flux surface. The shell is heated and then coated with liquid lithium. An extensive array of magnetic diagnostics is available to characterize the experiment, including 80 Mirnov coils (single and double axis, internal and external to the shell), 34 flux loops, 3 Rogowskii coils, and a diamagnetic loop. Diagnostics are specifically located to account for the presence of a secondary conducting surface and engineered to withstand both high temperatures and incidental contact with liquid lithium. The diagnostic set is therefore fabricated from robust materials with heat and lithium resistance and is designed for electrical isolation from the shell and to provide the data required for highly constrained equilibrium reconstructions.

Far-infrared laser diagnostics on the HT-6M tokamak

NASA Astrophysics Data System (ADS)

Gao, X.; Lu, H. J.; Guo, Q. L.; Wan, Y. X.; Tong, X. D.

1995-01-01

A multichannel far-infrared (FIR) hydrogen cyanide (HCN) laser interferometer was developed to measure plasma electron density profile on the HT-6M tokamak. The structure of the seven-channel FIR laser interferometer is described. The laser source used in the interferometer was a continuous-wave glow discharge HCN laser with a cavity length of 3.4 m and power output of about 100 mW at 337 μm. The detection sensitivity was 1/15 fringe with a temporal resolution of 0.1 ms. Experimental results were measured by the seven-channel FIR HCN laser interferometer with edge Ohmic heating, a pumping limiter, and ion cyclotron resonant heating on the HT-6M tokamak are reported.

Measurement of eddy-current distribution in the vacuum vessel of the Sino-UNIted Spherical Tokamak.

PubMed

Li, G; Tan, Y; Liu, Y Q

2015-08-01

Eddy currents have an important effect on tokamak plasma equilibrium and control of magneto hydrodynamic activity. The vacuum vessel of the Sino-UNIted Spherical Tokamak is separated into two hemispherical sections by a toroidal insulating barrier. Consequently, the characteristics of eddy currents are more complex than those found in a standard tokamak. Thus, it is necessary to measure and analyze the eddy-current distribution. In this study, we propose an experimental method for measuring the eddy-current distribution in a vacuum vessel. By placing a flexible printed circuit board with magnetic probes onto the external surface of the vacuum vessel to measure the magnetic field parallel to the surface and then subtracting the magnetic field generated by the vertical-field coils, the magnetic field due to the eddy current can be obtained, and its distribution can be determined. We successfully applied this method to the Sino-UNIted Spherical Tokamak, and thus, we obtained the eddy-current distribution despite the presence of the magnetic field generated by the external coils.

Snowflake divertor experiments in the DIII-D, NSTX, and NSTX-U tokamaks aimed at the development of the divertor power exhaust solution

DOE PAGES

Soukhanovskii, V. A.; Allen, S. L.; Fenstermacher, M. E.; ...

2016-11-16

Experimental results from the National Spherical Torus Experiment (NSTX), a medium-size spherical tokamak with a compact divertor, and DIII-D, a large conventional aspect ratio tokamak, demonstrate that the snowflake (SF) divertor configuration may provide a promising solution for mitigating divertor heat loads and target plate erosion compatible with core H-mode confinement in the future fusion devices, where the standard radiative divertor solution may be inadequate. In NSTX, where the initial high-power SF experiment was performed, the SF divertor was compatible with H-mode confinement, and led to the destabilization of large Edge Localized Modes (ELMs). However, a stable partial detachment ofmore » the outer strike point was also achieved where inter-ELM peak heat flux was reduced by factors 3-5, and peak ELM heat flux was reduced by up to 80% (see standard divertor). The DIII-D studies show the SF divertor enables significant power spreading in attached and radiative divertor conditions. Results include: compatibility with the core and pedestal, peak inter-ELM divertor heat flux reduction due to geometry at lower ne, and ELM energy and divertor peak heat flux reduction, especially prominent in radiative D 2-seeded SF divertor, and nearly complete power detachment and broader radiated power distribution in the radiative D 2-seeded SF divertor at PSOL = 3 - 4 MW. A variety of SF configurations can be supported by the divertor coil set in NSTX Upgrade. Edge transport modeling with the multifluid edge transport code UEDGE shows that the radiative SF divertor can successfully reduce peak divertor heat flux for the projected PSOL ≃ 9 MW case. Furthermore, the radiative SF divertor with carbon impurity provides a wider ne operating window, 50% less argon is needed in the impurity-seeded SF configuration to achieve similar q peak reduction factors (see standard divertor).« less

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.

Advances in Dust Detection and Removal for Tokamaks

NASA Astrophysics Data System (ADS)

Campos, A.; Skinner, C. H.; Roquemore, A. L.; Leisure, J. O. V.; Wagner, S.

2008-11-01

Dust diagnostics and removal techniques are vital for the safe operation of next step fusion devices such as ITER. An electrostatic dust detector[1] developed in the laboratory is being applied to NSTX. In the tokamak environment, large particles or fibres can fall on the grid potentially causing a permanent short. We report on the development of a gas puff system that uses helium to clear such particles from the detector. Experiments with varying nozzle designs, backing pressures, puff durations, and exit flow orientations have obtained an optimal configuration that effectively removes particles from a 25 cm^2 area. Dust removal from next step tokamaks will be required to meet regulatory dust limits. A tripolar grid of fine interdigitated traces has been designed that generates an electrostatic travelling wave for conveying dust particles to a ``drain.'' First trials have shown particle motion in optical microscope images. [1] C. H. Skinner et al., J. Nucl. Mater., 376 (2008) 29.

The prospects for magnetohydrodynamic stability in advanced tokamak regimes

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

Manickam, J.; Chance, M.S.; Jardin, S.C.

1994-05-01

Stability analysis of advanced regime tokamaks is presented. Here advanced regimes are defined to include configurations where the ratio of the bootstrap current, [ital I][sub BS], to the total plasma current, [ital I][sub [ital p

Numerical simulation of the multiple core localized low shear toroidal Alfvenic eigenmodes

NASA Astrophysics Data System (ADS)

Wang, Wenjia; Zhou, Deng; Hu, Youjun; Ming, Yue

2018-03-01

In modern tokamak experiments, scenarios with weak central magnetic shear has been proposed. It is necessary to study the Alfvenic mode activities in such scenarios. Theoretical researches have predicted the multiplicity of core-localized toroidally induced Alfvenic eigenmodes for ɛ/s > 1, where ɛ is the inverse aspect ratio and s is magnetic shear. We numerically investigate the existence of multiplicity of core-localized TAEs and mode characteristics using NOVA code in the present work. We firstly verify the existence of the multiplicity for zero beta plasma and the even mode at the forbidden zone. For finite beta plasma, the mode parities become more distinguishable, and the frequencies of odd modes are close to the upper tip of the continuum, while the frequencies of even modes are close to the lower tip of the continuum. Their frequencies are well separated by the forbidden zone. With the increasing value of ɛ/s, more modes with multiple radial nodes will appear, which is in agreement with theoretical prediction. The discrepancy between theoretical prediction and our numerical simulation is also discussed in the main text.

Study of Globus-M Tokamak Poloidal System and Plasma Position Control

NASA Astrophysics Data System (ADS)

Dokuka, V. N.; Korenev, P. S.; Mitrishkin, Yu. V.; Pavlova, E. A.; Patrov, M. I.; Khayrutdinov, R. R.

2017-12-01

In order to provide efficient performance of tokamaks with vertically elongated plasma position, control systems for limited and diverted plasma configuration are required. The accuracy, stability, speed of response, and reliability of plasma position control as well as plasma shape and current control depend on the performance of the control system. Therefore, the problem of the development of such systems is an important and actual task in modern tokamaks. In this study, the measured signals from the magnetic loops and Rogowski coils are used to reconstruct the plasma equilibrium, for which linear models in small deviations are constructed. We apply methods of the H∞-optimization theory to the synthesize control system for vertical and horizontal position of plasma capable to working with structural uncertainty of the models of the plant. These systems are applied to the plasma-physical DINA code which is configured for the tokamak Globus-M plasma. The testing of the developed systems applied to the DINA code with Heaviside step functions have revealed the complex dynamics of plasma magnetic configurations. Being close to the bifurcation point in the parameter space of unstable plasma has made it possible to detect an abrupt change in the X-point position from the top to the bottom and vice versa. Development of the methods for reconstruction of plasma magnetic configurations and experience in designing plasma control systems with feedback for tokamaks provided an opportunity to synthesize new digital controllers for plasma vertical and horizontal position stabilization. It also allowed us to test the synthesized digital controllers in the closed loop of the control system with the DINA code as a nonlinear model of plasma.

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.

1984-02-01

Auxiliary rf heating of electrons before and during the current-rise phase of a large tokamak, such as the Fusion Engineering Device (R = 4.8 m, a = 1.3 m, sigma = 1.6, B/sub T/ = 3.62 T), 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 themore » upper hybrid resonance (UHR) region. This plasma conditioning permits a small radius (a/sub 0/ approx. = 0.2 to 0.4 m) current channel to be established with a relatively low initial loop voltage (less than or equal to 25 V as opposed to approx. 100 V without rf assist). During the subsequent plasma expansion and current ramp phase, a combination of rf heating (up to 5 MW) and current profile control leads to a substantial savings in volt-seconds by: (1) minimizing the resistive flux consumption; and (2) maintaining the internal flux at or near the flat profile limit.« less

Petrographic Analysis of Cores from Plant 42

DTIC Science & Technology

2016-10-01

ER D C TR -X X- D R AF T Petrographic Analysis of Cores from Plant 42 E n gi n ee r R es ea rc h a n d D ev el op m en t C en te r...E. Rae Reed-Gore, Kyle Klaus, and Robert D. Moser October 2016 Approved for public release; distribution is unlimited. The U.S. Army...ORIGINATOR. ii Figures and Tables Figures Figure 1. Test location map of AF Plant 42 with core number locations. ............................. 2

Core Fueling of DEMO by Direct Line Injection of High-Speed Pellets From the HFS

DOE PAGES

Frattolillo, Antonio; Baylor, Larry R.; Bombarda, Francesca; ...

2018-04-17

Pellet injection represents to date the most realistic candidate technology for core fueling of a demonstration fusion power reactor tokamak fusion reactor. Modeling of both pellet penetration and fuel deposition profiles, for different injection locations, indicates that effective core fuelling can be achieved launching pellets from the inboard high field side at speeds not less than ~ 1 km/s. Inboard pellet fueling is commonly achieved in present tokamaks, using curved guide tubes; however, this technology might be hampered at velocities ≥ 1 km/s. An innovative approach, aimed at identifying suitable inboard "direct line'' paths, to inject high-speed pellets (in themore » 3 to 4 km/s range), has recently been proposed as a potential complementary solution. The fuel deposition profiles achievable by this approach have been explored using the HPI2 simulation code. The results presented here show that there are possible geometrical schemes providing good fueling performance. The problem of neutron flux in a direct line-of-sight injection path is being investigated, though preliminary analyses indicate that, perhaps, this is not a serious problem. The identification and integration of straight injection paths suitably tilted may be a rather difficult task due to the many constraints and to interference with existing structures. The suitability of straight guide tubes to reduce the scatter cone of high-speed pellets is, therefore, of main interest. A preliminary investigation, aimed at addressing these technological issues, has recently been started. As a result, a possible implementation plan, using an existing Italian National Agency for New Technologies, Energy and Sustainable Economic Development-Oak Ridge National Laboratory facility is shortly outlined.« less

Core Fueling of DEMO by Direct Line Injection of High-Speed Pellets From the HFS

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

Frattolillo, Antonio; Baylor, Larry R.; Bombarda, Francesca

Pellet injection represents to date the most realistic candidate technology for core fueling of a demonstration fusion power reactor tokamak fusion reactor. Modeling of both pellet penetration and fuel deposition profiles, for different injection locations, indicates that effective core fuelling can be achieved launching pellets from the inboard high field side at speeds not less than ~ 1 km/s. Inboard pellet fueling is commonly achieved in present tokamaks, using curved guide tubes; however, this technology might be hampered at velocities ≥ 1 km/s. An innovative approach, aimed at identifying suitable inboard "direct line'' paths, to inject high-speed pellets (in themore » 3 to 4 km/s range), has recently been proposed as a potential complementary solution. The fuel deposition profiles achievable by this approach have been explored using the HPI2 simulation code. The results presented here show that there are possible geometrical schemes providing good fueling performance. The problem of neutron flux in a direct line-of-sight injection path is being investigated, though preliminary analyses indicate that, perhaps, this is not a serious problem. The identification and integration of straight injection paths suitably tilted may be a rather difficult task due to the many constraints and to interference with existing structures. The suitability of straight guide tubes to reduce the scatter cone of high-speed pellets is, therefore, of main interest. A preliminary investigation, aimed at addressing these technological issues, has recently been started. As a result, a possible implementation plan, using an existing Italian National Agency for New Technologies, Energy and Sustainable Economic Development-Oak Ridge National Laboratory facility is shortly outlined.« less

Progress Toward Steady State Tokamak Operation Exploiting the high bootstrap current fraction regime

NASA Astrophysics Data System (ADS)

Ren, Q.

2015-11-01

Recent DIII-D experiments have advanced the normalized fusion performance of the high bootstrap current fraction tokamak regime toward reactor-relevant steady state operation. The experiments, conducted by a joint team of researchers from the DIII-D and EAST tokamaks, developed a fully noninductive scenario that could be extended on EAST to a demonstration of long pulse steady-state tokamak operation. Fully noninductive plasmas with extremely high values of the poloidal beta, βp >= 4 , have been sustained at βT >= 2 % for long durations with excellent energy confinement quality (H98y,2 >= 1 . 5) and internal transport barriers (ITBs) generated at large minor radius (>= 0 . 6) in all channels (Te, Ti, ne, VTf). Large bootstrap fraction (fBS ~ 80 %) has been obtained with high βp. ITBs have been shown to be compatible with steady state operation. Because of the unusually large ITB radius, normalized pressure is not limited to low βN values by internal ITB-driven modes. βN up to ~4.3 has been obtained by optimizing the plasma-wall distance. The scenario is robust against several variations, including replacing some on-axis with off-axis neutral beam injection (NBI), adding electron cyclotron (EC) heating, and reducing the NBI torque by a factor of 2. This latter observation is particularly promising for extension of the scenario to EAST, where maximum power is obtained with balanced NBI injection, and to a reactor, expected to have low rotation. However, modeling of this regime has provided new challenges to state-of-the-art modeling capabilities: quasilinear models can dramatically underpredict the electron transport, and the Sauter bootstrap current can be insufficient. The analysis shows first-principle NEO is in good agreement with experiments for the bootstrap current calculation and ETG modes with a larger saturated amplitude or EM modes may provide the missing electron transport. Work supported in part by the US DOE under DE-FC02-04ER54698, DE-AC52-07NA

Neoclassical tearing mode seeding by coupling with infernal modes in low-shear tokamaks

NASA Astrophysics Data System (ADS)

Kleiner, A.; Graves, J. P.; Brunetti, D.; Cooper, W. A.; Halpern, F. D.; Luciani, J.-F.; Lütjens, H.

2016-09-01

A numerical and an analytical study of the triggering of resistive MHD modes in tokamak plasmas with low magnetic shear core is presented. Flat q profiles give rise to fast growing pressure driven MHD modes, such as infernal modes. It has been shown that infernal modes drive fast growing islands on neighbouring rational surfaces. Numerical simulations of such instabilities in a MAST-like configuration are performed with the initial value stability code XTOR-2F in the resistive frame. The evolution of magnetic islands are computed from XTOR-2F simulations and an analytical model is developed based on Rutherford’s theory in combination with a model of resistive infernal modes. The parameter {{Δ }\\prime} is extended from the linear phase to the non-linear phase. Additionally, the destabilising contribution due to a helically perturbed bootstrap current is considered. Comparing the numerical XTOR-2F simulations to the model, we find that coupling has a strong destabilising effect on (neoclassical) tearing modes and is able to seed 2/1 magnetic islands in situations when the standard NTM theory predicts stability.

Estimation of Electron Temperature on Glass Spherical Tokamak (GLAST)

NASA Astrophysics Data System (ADS)

Hussain, S.; Sadiq, M.; Shah, S. I. W.; GLAST Team

2015-03-01

Glass Spherical Tokamak (GLAST) is a small spherical tokamak indigenously developed in Pakistan with an insulating vacuum vessel. A commercially available 2.45 GHz magnetron is used as pre-ionization source for plasma current startup. Different diagnostic systems like Rogowski coils, magnetic probes, flux loops, Langmuir probe, fast imaging and emission spectroscopy are installed on the device. The plasma temperature inside of GLAST, at the time of maxima of plasma current, is estimated by taking into account the Spitzer resistivity calculations with some experimentally determined plasma parameters. The plasma resistance is calculated by using Ohm's law with plasma current and loop voltage as experimentally determined inputs. The plasma resistivity is then determined by using length and area of the plasma column. Finally, the average plasma electron temperature is predicted to be 12.65eV for taking neon (Ne) as a working gas.

Progress on the Implementation of a Neutral Beam for the Lithium Tokamak eXperiment-Beta

NASA Astrophysics Data System (ADS)

Merino, Enrique; Kozub, Thomas; Boyle, Dennis; Majeski, Richard; Kaita, Robert; Smirnov, Artem; Catalano, Ryan

2016-10-01

In the Lithium Tokamak eXperiment (LTX), good performance discharges have been achieved with reduced-recycling lithium walls. Two hydrogen neutral beams (NB) have been loaned to the LTX project by Tri-Alpha Energy, Inc. To further improve plasma parameters, one of these neutral beams is being installed as part of an upgrade to LTX (LTX-Beta). Current ohmic input power in LTX is less than 100 kW. The NB will provide core plasma fueling with up to 700 kW of injected power. Requirements for accommodating the NB include the addition of injection and beam-dump ports on the vessel, and their designs have been finalized. Progress has also been made on the NB power supplies, including the preparation of a new room to accommodate them. A description of these activities and the status of other improvements to LTX for LTX-Beta will be presented. Work supported by US DOE contracts DE-AC02- 09CH11466 and DE-AC05- 00OR22725.

Plasma cleaning of ITER edge Thomson scattering mock-up mirror in the EAST tokamak

NASA Astrophysics Data System (ADS)

Yan, Rong; Moser, Lucas; Wang, Baoguo; Peng, Jiao; Vorpahl, Christian; Leipold, Frank; Reichle, Roger; Ding, Rui; Chen, Junling; Mu, Lei; Steiner, Roland; Meyer, Ernst; Zhao, Mingzhong; Wu, Jinhua; Marot, Laurent

2018-02-01

First mirrors are the key element of all optical and laser diagnostics in ITER. Facing the plasma directly, the surface of the first mirrors could be sputtered by energetic particles or deposited with contaminants eroded from the first wall (tungsten and beryllium), which would result in the degradation of the reflectivity. The impurity deposits emphasize the necessity of the first mirror in situ cleaning for ITER. The mock-up first mirror system for ITER edge Thomson scattering diagnostics has been cleaned in EAST for the first time in a tokamak using radio frequency capacitively coupled plasma. The cleaning properties, namely the removal of contaminants and homogeneity of cleaning were investigated with molybdenum mirror insets (25 mm diameter) located at five positions over the mock-up plate (center to edge) on which 10 nm of aluminum oxide, used as beryllium proxy, were deposited. The cleaning efficiency was evaluated using energy dispersive x-ray spectroscopy, reflectivity measurements and x-ray photoelectron spectroscopy. Using argon or neon plasma without magnetic field in the laboratory and with a 1.7 T magnetic field in the EAST tokamak, the aluminum oxide films were homogeneously removed. The full recovery of the mirrors’ reflectivity was attained after cleaning in EAST with the magnetic field, and the cleaning efficiency was about 40 times higher than that without the magnetic field. All these results are promising for the plasma cleaning baseline scenario of ITER.

Structural and core parameters of the hot B subdwarf KPD 0629-0016 from CoRoT g-mode asteroseismology

NASA Astrophysics Data System (ADS)

Van Grootel, V.; Charpinet, S.; Fontaine, G.; Green, E. M.; Brassard, P.

2010-12-01

Context. The asteroseismic exploitation of long period, g-mode hot B subdwarf pulsators (sdBVs), undermined so far by limitations associated with ground-based observations, has now become possible, thanks to high quality data obtained from space such as those recently gathered with the CoRoT (COnvection, ROtation, and planetary Transits) satellite. Aims: We propose a detailed seismic analysis of the sdBVs star KPD 0629-0016, the first compact pulsator monitored with CoRoT, using the g-mode pulsations recently uncovered by that space-borne observatory during short run SRa03. Methods: We use a forward modeling approach on the basis of our latest sdB models, which are now suitable for the accurate computation of the g-mode pulsation properties. The simultaneous match of the independent periods observed in KPD 0629-0016 with those of the models leads objectively to the identification of the pulsation modes and, more importantly, to the determination of the structural and core parameters of the star. Results: The optimal model we found closely reproduces the 18 observed periods retained in our analysis at a 0.23% level on average. These are identified as low-degree (ℓ = 1 and 2), intermediate-order (k = -9 through -74) g-modes. The structural and core parameters for KPD 0629-0016 are the following (formal fitting errors only): Teff = 26 290 ± 530 K, log g = 5.450 ± 0.034, M_* = 0.471 ± 0.002 M⊙, log (Menv/M_*) = -2.42 ± 0.07, log (1-Mcore/M_*) = -0.27 ± 0.01, and Xcore(C+O) = 0.41 ± 0.01. We additionally derive an age of 42.6 ± 1.0 Myr after the zero-age extreme horizontal branch, the radius R = 0.214 ± 0.009 R⊙, the luminosity L = 19.7 ± 3.2 L⊙, the absolute magnitude MV = 4.23 ± 0.13, the reddening index E(B-V) = 0.128 ± 0.023, and the distance d = 1190 ± 115 pc. Conclusions: The advent of high-precision time-series photometry from space with instruments like CoRoT now allows as demonstrated with KPD 0629-0016 the full exploitation of g-modes as

High performance discharges in the Lithium Tokamak eXperiment with liquid lithium walls

DOE PAGES

Schmitt, J. C.; Bell, R. E.; Boyle, D. P.; ...

2015-05-15

The first-ever successful operation of a tokamak with a large area (40% of the total plasma surface area) liquid lithium wall has been achieved in the Lithium Tokamak eXperiment (LTX). These results were obtained with a new, electron beam-based lithium evaporation system, which can deposit a lithium coating on the limiting wall of LTX in a five-minute period. Preliminary analyses of diamagnetic and other data for discharges operated with a liquid lithium wall indicate that confinement times increased by 10 x compared to discharges with helium-dispersed solid lithium coatings. Ohmic energy confinement times with fresh lithium walls, solid and liquid,more » exceed several relevant empirical scaling expressions. Spectroscopic analysis of the discharges indicates that oxygen levels in the discharges limited on liquid lithium walls were significantly reduced compared to discharges limited on solid lithium walls. Finally, Tokamak operations with a full liquid lithium wall (85% of the total plasma surface area) have recently started.« less

Modeling of rapid shutdown in the DIII-D tokamak by core deposition of high-Z material

DOE PAGES

Izzo, Valerie A.; Parks, Paul B.

2017-06-22

MHD modeling of shell-pellet injection for disruption mitigation is carried out under the assumption of idealized delivery of the radiating payload to the core, neglecting the physics of shell ablation. The shell pellet method is designed to produce an inside-out thermal quench in which core thermal heat is radiated while outer flux surfaces remain intact, protecting the divertor from large conducted heat loads. In the simulation, good outer surfaces remain until the thermal quench is nearly complete, and a high radiated energy fraction is achieved. As a result, when the outermost surfaces are destroyed, runaway electron test orbits indicate thatmore » the rate of runaway electron loss is very fast compared with prior massive gas injection simulations, which is attributed to the very different current profile evolution that occurs with central cooling.« less

Runaway electrons in tokamaks

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

Liu, Chang

The generation of runaway electrons is a complex and important phenomenon that impacts many areas of plasma physics. Due to the decrease of electron collision frequency with increasing velocity, electrons under strong electric field can experience unlimited “runaway” acceleration. In tokamaks, runaway electrons can be produced in disruptions, due to the strong inductive electric field formed as the thermal energy of plasma gets rapidly lost. This population of runaway electrons can undergo an exponential growth, denoted the runaway electron avalanche, due to hard collisions between relativistic runaway electrons and low energy electrons. It is predicted that in a large tokamakmore » device like the International Thermonuclear Experimental Reactor (ITER), a runway electron beam generated in a disruption event can potentially cause severe damage to the device, which poses a significant challenge for ITER to achieve its mission. It is therefore extremely important to seek an effective mitigation mechanism for runaway electrons. Experimental efforts have been made to study the properties of runaway electrons in tokamaks, including their generation, diffusion, and radiation. In order to understand these experimental results, extensive theoretical and simulation studies of runaway electron physics are required. The main topic of this thesis is to study the wave particle interaction associated with runaway electron beams in tokamaks. The runaway electrons can emit and absorb electromagnetic waves through resonances, and can be diffused in momentum space by the waves. Initially, we address the Cherenkov radiation of runaway electrons, which originates from the polarization of the plasma medium. The energy and momentum loss of the Cherenkov radiation can be modeled by adding a correction to the Coulomb logarithm in the collisional drag force. Subsequently, we address pitch angle scattering caused by normal modes in the plasma, which are driven unstable by the

Anomalous transport scaling in the DIII-D tokamak matched by supercomputer simulation.

PubMed

Candy, J; Waltz, R E

2003-07-25

Gyrokinetic simulation of tokamak transport has evolved sufficiently to allow direct comparison of numerical results with experimental data. It is to be emphasized that only with the simultaneous inclusion of many distinct and complex effects can this comparison realistically be made. Until now, numerical studies of tokamak microturbulence have been restricted to either (a) flux tubes or (b) electrostatic fluctuations. Using a newly developed global electromagnetic solver, we have been able to recover via direct simulation the Bohm-like scaling observed in DIII-D L-mode discharges. We also match, well within experimental uncertainty, the measured energy diffusivities.

Experimental observation of m/n = 1/1 mode behaviour during sawtooth activity and its manifestations in tokamak plasmas

NASA Astrophysics Data System (ADS)

Udintsev, V. S.; Ottaviani, M.; Maget, P.; Giruzzi, G.; Ségui, J.-L.; Aniel, T.; Artaud, J. F.; Clairet, F.; Goniche, M.; Hoang, G. T.; Huysmans, G. T. A.; Imbeaux, F.; Joffrin, E.; Mazon, D.; Pecquet, A. L.; Sabot, R.; Sirinelli, A.; Vermare, L.; Tore Supra Team; Krämer-Flecken, A.; Koslowski, H. R.; TEXTOR Team; Donné, A. J. H.; Schüller, F. C.; Domier, C. W.; Luhmann, N. C., Jr.; Mirnov, S. V.

2005-08-01

To shed some light on the development of the fast m/n = 1/1 precursor to the sawtooth crash and its influence on plasma transport properties in the vicinity of the q = 1 surface, series of dedicated experiments have been conducted on the Tore Supra and TEXTOR tokamaks. It has been concluded that, before a crash, the hot core gets displaced with respect to the magnetic axis, drifts outwards by as much as 8-10 cm and may change its shape. Observation of the magnetic reconnection process has been made by means of electron cyclotron emission diagnostics. The heat pulse is seen far outside the inversion radius. The colder plasma develops a magnetic island on the former magnetic axis, after the hot core expulsion. Different kinds of behaviour of the m = 1 precursor before the crash, with respect to the displacement of the hot core and the duration of the oscillating phase, have been observed. An ideal kink model alone cannot be used for explanation; therefore, resistive effects play an important role in the mode development. Possible mechanisms that lead an m = 1 mode to such behaviour, and their links to the change in the central q-profile, are discussed. Results have been discussed in the light of various theoretical models of the sawtooth. This is an extended version of the paper presented at the 31st EPS Conf. on Plasma Physics (London, 28 June-2 July 2004).

Alpha-channeling simulation experiment in the DIII-D tokamak.

PubMed

Wong, K L; Budny, R; Nazikian, R; Petty, C C; Greenfield, C M; Heidbrink, W W; Ruskov, E

2004-08-20

Alfvén instabilities can reduce the central magnetic shear via redistribution of energetic ions. They can sustain a steady state internal transport barrier as demonstrated in this DIII-D tokamak experiment. Improvement in burning plasma performance based on this mechanism is discussed.

Phase Contrast Imaging on the HL-2A Tokamak

NASA Astrophysics Data System (ADS)

Yu, Yi; Gong, Shaobo; Xu, Min; Jiang, Wei; Zhong, Wulv; Shi, Zhongbin; Wang, Huajie; Wu, Yifan; Yuan, Boda; Lan, Tao; Ye, Minyou; Duan, Xuru; HL-2A Team

2016-10-01

In this article we present the design of a phase contrast imaging (PCI) system on the HL-2A tokamak. This diagnostic is developed to infer line integrated plasma density fluctuations by measuring the phase shift of an expanded CO2 laser beam passing through magnetically confined high temperature plasmas. This system is designed to diagnose plasma density fluctuations with the maximum wavenumber of 66 cm-1. The designed wavenumber resolution is 2.09cm-1, and the time resolution is higher than 0.2 μs. The broad kρs ranging from 0.34 to 13.37 makes it suitable for turbulence measurement. An upgraded PCI system is also discussed, which is designed for the HL-2M tokamak. Supported by the National Magnetic Confinement Fusion Energy Research Project (Grant No. 2015GB120002), the National Natural Science Foundation of China (Grant No. 11375053, 11105144, 10905057, 11535013).

Impact of helical boundary conditions in MHD modeling of RFP and tokamak plasmas

NASA Astrophysics Data System (ADS)

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

2012-10-01

Helical boundary conditions imposed by the active control system of the RFX-mod device provide a handle to govern the plasma dynamics in both RFP and Ohmic tokamak discharges [1]. By applying an edge radial magnetic field with proper helicity, it is possible to increase the persistence of the spontaneous helical RFP states at high current,and to stimulate them also at low current or high density. Helical BCs even allow to access helical states with different helicity than the spontaneous one [2]. In Ohmic tokamak operation at q(a)<2, the presence of the 2/1 RWM reduces the sawtoothing activity of the 1/1 internal kink, which takes a stationary snake-like character instead. Many of these features are qualitatively reproduced in 3D nonlinear MHD modeling. We study the impact of helical BCs on the MHD dynamics in both RFP and tokamak with two successfully benchmarked numerical tools, SpeCyl and PIXIE3D [3]. We recover the bifurcation from a sawtooth to a snake solution when imposing a 2/1 BC in the tokamak case and we interpret this as a toroidal/nonlinear coupling effect. We show that the bifurcation is more easily stimulated with a 1/1 BC.[4pt] [1] P. Piovesan, invited talk this meeting[0pt] [2] M. Veranda et al EPS-ICPP Conference (2012) P4.004[0pt] [3] D. Bonfiglio et al Phys. Plasmas (2010)

High Adjustable Shear Map for a Single-null Divertor Tokamak

NASA Astrophysics Data System (ADS)

Whitten, Michaelangelo; Lam, Maria; Punjabi, Alkesh

1996-11-01

An explicit map that has an adjustable shear s is x1 = x0 - k y0 [ ( 1 - y0 ) ( 1 + s y0 ) + s x ^21 ] y1 = y0 + k x1 [ 1 + s ( x^21 + y^20 ) ] Tokamak shear corresponds to negative s. Thus we can construct maps for variable shear for a single-null divertor tokamak (Punjabi A, Verma A and Boozer A, Phys Rev Lett), 3322, 69 (1992) ^, (Punjabi A, Verma A and Boozer A, J Plasma Phys), 52, 91 (1994). Here we present the results from an initial study of this map. This work is supported by US DOE OFES. Michelangelo Whitten is a HU CFRT Summer Fusion High School Workshop scholar from Bowie High School in El Paso, Texas. He is supported by NASA SHARP Plus Program.

Dust measurements in tokamaks (invited).

PubMed

Rudakov, D L; Yu, J H; Boedo, J A; Hollmann, E M; Krasheninnikov, S I; Moyer, R A; Muller, S H; Pigarov, A Yu; Rosenberg, M; Smirnov, R D; West, W P; Boivin, R L; Bray, B D; Brooks, N H; Hyatt, A W; Wong, C P C; Roquemore, A L; Skinner, C H; Solomon, W M; Ratynskaia, S; Fenstermacher, M E; Groth, M; Lasnier, C J; McLean, A G; Stangeby, P C

2008-10-01

Dust production and accumulation present potential safety and operational issues for the ITER. Dust diagnostics can be divided into two groups: diagnostics of dust on surfaces and diagnostics of dust in plasma. Diagnostics from both groups are employed in contemporary tokamaks; new diagnostics suitable for ITER are also being developed and tested. Dust accumulation in ITER is likely to occur in hidden areas, e.g., between tiles and under divertor baffles. A novel electrostatic dust detector for monitoring dust in these regions has been developed and tested at PPPL. In the DIII-D tokamak dust diagnostics include Mie scattering from Nd:YAG lasers, visible imaging, and spectroscopy. Laser scattering is able to resolve particles between 0.16 and 1.6 microm in diameter; using these data the total dust content in the edge plasmas and trends in the dust production rates within this size range have been established. Individual dust particles are observed by visible imaging using fast framing cameras, detecting dust particles of a few microns in diameter and larger. Dust velocities and trajectories can be determined in two-dimension with a single camera or three-dimension using multiple cameras, but determination of particle size is challenging. In order to calibrate diagnostics and benchmark dust dynamics modeling, precharacterized carbon dust has been injected into the lower divertor of DIII-D. Injected dust is seen by cameras, and spectroscopic diagnostics observe an increase in carbon line (CI, CII, C(2) dimer) and thermal continuum emissions from the injected dust. The latter observation can be used in the design of novel dust survey diagnostics.

Design and Fabrication of the Lithium Tokamak Experiment

NASA Astrophysics Data System (ADS)

Kozub, Thomas; Majeski, Richard; Kaita, Robert; Priniski, Craig; Zakharov, Leonid

2006-10-01

The design objective of the lithium tokamak experiment (LTX) is to investigate the equilibrium and stability of tokamak discharges with near-zero recycling. The construction of LTX incorporates the conversion of the existing current drive experiment (CDX) vessel into one with a nearly complete plasma facing surface of liquid lithium This paper will describe the design, fabrication, and installation activities required to convert CDX into LTX. The most significant new feature is the addition of a plasma facing liner on a shell that will be operated at 300 C to 400 C and covered with an evaporated layer of liquid lithium. The shell has been fabricated in-house from explosively bonded stainless steel on copper to a rather unique geometry to match the outer flux surface. Other significant device modifications include the construction of a new ohmic heating power system, rebuilding of the vacuum vessel, new lithium evaporators, additional diagnostics, modifications to the poloidal field coil geometry and their associated power supplies. Details on the progress of this conversion will be reported.

Impact of E × B flow shear on turbulence and resulting power fall-off width in H-mode plasmas in experimental advanced superconducting tokamak

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

Yang, Q. Q., E-mail: yangqq@ipp.ac.cn; Zhong, F. C., E-mail: gsxu@ipp.ac.cn, E-mail: fczhong@dhu.edu.cn; Jia, M. N.

2015-06-15

The power fall-off width in the H-mode scrape-off layer (SOL) in tokamaks shows a strong inverse dependence on the plasma current, which was noticed by both previous multi-machine scaling work [T. Eich et al., Nucl. Fusion 53, 093031 (2013)] and more recent work [L. Wang et al., Nucl. Fusion 54, 114002 (2014)] on the Experimental Advanced Superconducting Tokamak. To understand the underlying physics, probe measurements of three H-mode discharges with different plasma currents have been studied in this work. The results suggest that a higher plasma current is accompanied by a stronger E×B shear and a shorter radial correlation lengthmore » of turbulence in the SOL, thus resulting in a narrower power fall-off width. A simple model has also been applied to demonstrate the suppression effect of E×B shear on turbulence in the SOL and shows relatively good agreement with the experimental observations.« less

Self-Organized Stationary States of Tokamaks

DOE PAGES

Jardin, S. C.; Ferraro, N.; Krebs, I.

2015-11-17

We demonstrate that in a 3D resistive magnetohydrodynamic (MHD) simulation, for some parameters it is possible to form a stationary state in a tokamak where a saturated interchange mode in the center of the discharge drives a near helical flow pattern that acts to non-linearly sustain the configuration by adjusting the central loop voltage through a dynamo action. This could explain the physical mechanism for maintaining stationary non-sawtoothing “hybrid” discharges, often referred to as “flux-pumping”.

The non-resonant kink modes triggering strong sawtooth-like crashes in the EAST tokamak

NASA Astrophysics Data System (ADS)

Li, Erzhong; Igochine, V.; Dumbrajs, O.; Xu, L.; Chen, K.; Shi, T.; Hu, L.

2014-12-01

Evolution of the safety factor (q) profile during L-H transitions in the Experimental Advanced Superconducting Tokamak (EAST) was accompanied by strong core crashes prior to regular sawtooth behavior. These crashes appeared in the absence of q = 1 (q is the safety factor) rational surface inside the plasma. Analysis indicates that the m/n = 2/1 tearing mode is destabilized and phase-locked with the m/n = 1/1 non-resonant kink mode (the q = 1 rational surface is absent) due to the self-consistent evolution of plasma profiles as the L-H transition occurs (m and n are the poloidal and toroidal mode numbers, respectively). The growing m/n = 1/1 mode destabilizes the m/n = 2/2 kink mode which eventually triggers the strong crash due to an anomalous heat conductivity, as predicted by the transport model of stochastic magnetic fields using experimental parameters. It is also shown that the magnetic topology changes with the amplitude of m/n = 2/2 mode and the value of center safety factor in a reasonable range.

Measurement of H/D ratio and ion temperature on a HT-6M Tokamak

NASA Astrophysics Data System (ADS)

Wei, Lehan; Lin, Xiaodong

1997-01-01

By combining optical fibers with piezoelectric scanning Fabry-Perot interferometer, the profiles of Hα and Dα have been determined simultaneously in a single Tokamak discharge. Consequently, the ratio of hydrogen to deuterium and ion temperature are obtained. Not only is the uncertainty of shot-to-shot avoided, the results of the experiment indicate that this instrumentation has the advantage of rapid wavelength scanning, large dispersion, high resolution, and good adaptability of working in adverse circumstances such as at a Tokamak site.

Neutron and gamma flux distributions and their implications for radiation damage in the shielded superconducting core of a fusion power plant

NASA Astrophysics Data System (ADS)

Windsor, Colin G.; Morgan, J. Guy

2017-11-01

The neutron and gamma ray fluxes within the shielded high-temperature superconducting central columns of proposed spherical tokamak power plants have been studied using the MCNP Monte-Carlo code. The spatial, energy and angular variations of the fluxes over the shield and superconducting core are computed and used to specify experimental studies relevant to radiation damage and activation. The mean neutron and gamma fluxes, averaged over energy and angle, are shown to decay exponentially through the shield and then to remain roughly constant in the core region. The mean energy of neutrons is shown to decay more slowly than the neutron flux through the shield while the gamma energy is almost constant around 2 MeV. The differential neutron and gamma fluxes as a function of energy are examined. The neutron spectrum shows a fusion peak around 1 MeV changing at lower energies into an epithermal E -0.85 variation and at thermal energies to a Maxwellian distribution. The neutron and gamma energy spectra are defined for the outer surface of the superconducting core, relevant to damage studies. The inclusion of tungsten boride in the shield is shown to reduce energy deposition. A series of plasma scenarios with varying plasma major radii between 0.6 and 2.5 m was considered. Neutron and gamma fluxes are shown to decay exponentially with plasma radius, except at low shield thickness. Using the currently known experimental fluence limitations for high temperature superconductors, the continuous running time before the fluence limit is reached has been calculated to be days at 1.4 m major radius increasing to years at 2.2 m. This work helps validate the concept of the spherical tokamak route to fusion power by demonstrating that the neutron shielding required for long lifetime fusion power generation can be accommodated in a compact device.

Ageing of structural materials in tokamaks: TEXTOR liner study

NASA Astrophysics Data System (ADS)

Weckmann, A.; Petersson, P.; Rubel, M.; Fortuna-Zaleśna, E.; Zielinski, W.; Romelczyk-Baishya, B.; Grigore, E.; Ruset, C.; Kreter, A.

2017-12-01

After the final shut-down of the tokamak TEXTOR, all of its machine parts became accessible for comprehensive studies. This unique opportunity enabled the study of the Inconel 625 liner by a wide range of methods. The aim was to evaluate eventual alteration of surface and bulk characteristics from recessed wall elements that may influence the machine performance. The surface was covered with stratified layers consisting mainly of boron, carbon, oxygen, and in some cases also silicon. Wall conditioning and limiter materials hence predominantly define deposition on the liner. Deposited layers on recessed wall elements reach micrometre thickness within decades, peel off and may contribute to the dust inventory in tokamaks. Deuterium content was about 4,7 at% on average most probably due to wall conditioning with deuterated gas, and very low concentration in the Inconel substrate. Inconel 625 retained its mechanical strength despite 26 years of cyclic heating, stresses and particle bombardment.

Investigation of neutral particle dynamics in Aditya tokamak plasma with DEGAS2 code

NASA Astrophysics Data System (ADS)

Dey, Ritu; Ghosh, Joydeep; Chowdhuri, M. B.; Manchanda, R.; Banerjee, S.; Ramaiya, N.; Sharma, Deepti; Srinivasan, R.; Stotler, D. P.; Aditya Team

2017-08-01

Neutral particle behavior in Aditya tokamak, which has a circular poloidal ring limiter at one particular toroidal location, has been investigated using DEGAS2 code. The code is based on the calculation using Monte Carlo algorithms and is mainly used in tokamaks with divertor configuration. This code has been successfully implemented in Aditya tokamak with limiter configuration. The penetration of neutral hydrogen atom is studied with various atomic and molecular contributions and it is found that the maximum contribution comes from the dissociation processes. For the same, H α spectrum is also simulated and matched with the experimental one. The dominant contribution around 64% comes from molecular dissociation processes and neutral particle is generated by those processes have energy of ~2.0 eV. Furthermore, the variation of neutral hydrogen density and H α emissivity profile are analysed for various edge temperature profiles and found that there is not much changes in H α emission at the plasma edge with the variation of edge temperature (7-40 eV).

Investigation of neutral particle dynamics in Aditya tokamak plasma with DEGAS2 code

DOE PAGES

Dey, Ritu; Ghosh, Joydeep; Chowdhuri, M. B.; ...

2017-06-09

Neutral particle behavior in Aditya tokamak, which has a circular poloidal ring limiter at one particular toroidal location, has been investigated using DEGAS2 code. The code is based on the calculation using Monte Carlo algorithms and is mainly used in tokamaks with divertor configuration. This code has been successfully implemented in Aditya tokamak with limiter configuration. The penetration of neutral hydrogen atom is studied with various atomic and molecular contributions and it is found that the maximum contribution comes from the dissociation processes. For the same, H α spectrum is also simulated which was matched with the experimental one. Themore » dominant contribution around 64% comes from molecular dissociation processes and neutral particle is generated by those processes have energy of ~ 2.0 eV. Furthermore, the variation of neutral hydrogen density and H α emissivity profile are analysed for various edge temperature profiles and found that there is not much changes in H α emission at the plasma edge with the variation of edge temperature (7 to 40 eV).« less

Fast island phase identification for tearing mode feedback control on J-TEXT tokamak

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

Rao, B., E-mail: borao@hust.edu.cn; Li, D.; Hu, F. R.

A new method to control the tearing mode (TM) in tokamaks has been proposed [Q. Hu and Q. Yu, Nucl. Fusion 56, 034001 (5pp.) (2016)], according to which, the external resonant magnetic perturbation needs to be applied in certain magnetic island phase regions. Therefore, it is very important to identify the helical phase of magnetic islands in real time. The TM in tokamak plasmas is normally rotating and carries magnetic oscillations, which are known as Mirnov oscillations and can be detected by Mirnov probes. When the O-point or X-point of the magnetic island passes through the probe, the signal willmore » experience a zero-crossing. A poloidal Mirnov probe array and a corresponding island phase identification method are presented. A field-programmable gate array is used to provide the magnetic island helical phase in real time by using multichannel zero crossing detection. This system has been developed on the J-TEXT tokamak and works well. This paper introduces the establishment of the fast magnetic island phase identifying system.« less

Progress toward steady-state tokamak operation exploiting the high bootstrap current fraction regime

DOE PAGES

Ren, Q. L.; Garofalo, A. M.; Gong, X. Z.; ...

2016-06-20

Recent DIII-D experiments have increased the normalized fusion performance of the high bootstrap current fraction tokamak regime toward reactor-relevant steady state operation. The experiments, conducted by a joint team of researchers from the DIII-D and EAST tokamaks, developed a fully noninductive scenario that could be extended on EAST to a demonstration of long pulse steady-state tokamak operation. Improved understanding of scenario stability has led to the achievement of very high values of β p and β N despite strong ITBs. Good confinement has been achieved with reduced toroidal rotation. These high β p plasmas challenge the energy transport understanding, especiallymore » in the electron energy channel. A new turbulent transport model, named 2 TGLF-SAT1, has been developed which improves the transport prediction. Experiments extending results to long pulse on EAST, based on the physics basis developed at DIII-D, have been conducted. Finally, more investigations will be carried out on EAST with more additional auxiliary power to come online in the near term.« less

High-sensitivity O-glycomic analysis of mice deficient in core 2 β1,6-N-acetylglucosaminyltransferases

PubMed Central

Ismail, Mohd Nazri; Stone, Erica L; Panico, Maria; Lee, Seung Ho; Luu, Ying; Ramirez, Kevin; Ho, Samuel B; Fukuda, Minoru; Marth, Jamey D; Haslam, Stuart M; Dell, Anne

2011-01-01

Core 2 β1,6-N-acetylglucosaminyltransferase (C2GnT), which exists in three isoforms, C2GnT1, C2GnT2 and C2GnT3, is one of the key enzymes in the O-glycan biosynthetic pathway. These isoenzymes produce core 2 O-glycans and have been correlated with the biosynthesis of core 4 O-glycans and I-branches. Previously, we have reported mice with single and multiple deficiencies of C2GnT isoenzyme(s) and have evaluated the biological and structural consequences of the loss of core 2 function. We now present more comprehensive O-glycomic analyses of neutral and sialylated glycans expressed in the colon, small intestine, stomach, kidney, thyroid/trachea and thymus of wild-type, C2GnT2 and C2GnT3 single knockouts and the C2GnT1–3 triple knockout mice. Very high-quality data have emerged from our mass spectrometry techniques with the capability of detecting O-glycans up to at least 3500 Da. We were able to unambiguously elucidate the types of O-glycan core, branching location and residue linkages, which allowed us to exhaustively characterize structural changes in the knockout tissues. The C2GnT2 knockout mice suffered a major loss of core 2 O-glycans as well as glycans with I-branches on core 1 antennae especially in the stomach and the colon. In contrast, core 2 O-glycans still dominated the O-glycomic profile of most tissues in the C2GnT3 knockout mice. Analysis of the C2GnT triple knockout mice revealed a complete loss of both core 2 O-glycans and branched core 1 antennae, confirming that the three known isoenzymes are entirely responsible for producing these structures. Unexpectedly, O-linked mannosyl glycans are upregulated in the triple deficient stomach. In addition, our studies have revealed an interesting terminal structure detected on O-glycans of the colon tissues that is similar to the RM2 antigen from glycolipids. PMID:20855471

Effect of Magnetic Islands on Divertors in Tokamaks and Stellarators

NASA Astrophysics Data System (ADS)

Punjabi, Alkesh; Boozer, Allen

2017-10-01

Divertors are required for handling the plasma particle and heat exhausts on the walls in fusion plasmas. Relatively simple methods, models, and maps from field line Hamiltonian are developed to better understand the interaction of strong plasma shaping and magnetic islands on the size and behavior of the magnetic flux tubes that go from the plasma edge to the wall in non-axisymmetric system. This approach is applicable not only in tokamaks but also in stellarators. Stellarator diverters in which magnetic islands are dominant are called resonant and when shaping is dominant are called non-resonant. Optimized stellarators generally have sharp edges on their surface, but unlike the case for tokamaks these edges do not encircle the entire plasma, so they do not define an edge value for the rotational transform. The approach is used in the DIII-D tokamak. Computation results are consistent with the predictions of the models. Further simulations are being done to understand why the transition from an effective cubic to a linear increase in loss time and area of footprint occurs and whether this increase is discontinuous or not. This work is supported by the US DOE Grants DE-FG02-01ER54624 and DE-FG02-04ER54793 to Hampton University and DE-FG02-95ER54333 to Columbia University. This research used resources of the NERSC, supported by the Office of Science, US DOE, under Contract No. DE-AC02-05CH11231.

Electronic structure and intersubband magnetoabsorption spectra of CdSe/CdS core-shell nanowires

NASA Astrophysics Data System (ADS)

Xiong, Wen

2016-10-01

The electronic structures of CdSe/CdS core-shell nanowires are calculated based on the effective-mass theory, and it is found that the hole states in CdSe/CdS core-shell nanowires are strongly mixed, which are very different from the hole states in CdSe or CdS nanowires. In addition, we find the three highest hole states at the Γ point are almost localized in the CdSe core and the energies of the hole states in CdSe/CdS core-shell nanowires can be enhanced greatly when the core radius Rc increases and the total radius R is fixed. The degenerate hole states are split by the magnetic field, and the split energies will increase when |Jh | increases from 1/2 to 7/2, while they are almost not influenced by the change of the core radius Rc. The absorption spectra of CdSe/CdS core-shell nanowires at the Γ point are also studied in the magnetic field when the temperature T is considered, and we find there are only two peaks will arise if the core radius Rc and the temperature T increase. The intensity of each optical absorption can be considerably enhanced by increasing the core radius Rc when the temperature T is fixed, it is due to the increase of their optical transition matrix element. Meanwhile, the intensity of each optical absorption can be decreased when the temperature T increases and the core radius Rc is fixed, and this is because the Fermi-Dirac distribution function of the corresponding hole states will increase as the increase of the temperature T.

Transport of carbon ion test particles and hydrogen recycling in the plasma of the Columbia tokamak HBT'' (High Beta Tokamak)

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

Wang, Jian-Hua.

Carbon impurity ion transport is studied in the Columbia High Beta Tokamak (HBT), using a carbon tipped probe which is inserted into the plasma (n{sub e} {approx} 1 {minus} 5 {times} 10{sup 14} (cm{sup {minus}3}), T{sub e} {approx} 4 {minus} 10 (eV), B{sub t} {approx} 0.2 {minus} 0.4(T)). Carbon impurity light, mainly the strong lines of C{sub II}(4267A, emitted by the C{sup +} ions) and C{sub III} (4647A, emitted by the C{sup ++} ions), is formed by the ablation or sputtering of plasma ions and by the discharge of the carbon probe itself. The diffusion transport of the carbon ionsmore » is modeled by measuring the space-and-time dependent spectral light emission of the carbon ions with a collimated optical beam and photomultiplier. The point of emission can be observed in such a way as to sample regions along and transverse to the toroidal magnetic field. The carbon ion diffusion coefficients are obtained by fitting the data to a diffusion transport model. It is found that the diffusion of the carbon ions is classical'' and is controlled by the high collisionality of the HBT plasma; the diffusion is a two-dimensional problem and the expected dependence on the charge of the impurity ion is observed. The measurement of the spatial distribution of the H{sub {alpha}} emissivity was obtained by inverting the light signals from a 4-channel polychromator, the data were used to calculate the minor-radial influx, the density, and the recycling time of neutral hydrogen atoms or molecules. The calculation shows that the particle recycling time {tau}{sub p} is comparable with the plasma energy confinement time {tau}{sub E}; therefore, the recycling of the hot plasma ions with the cold neutrals from the walls is one of the main mechanisms for loss of plasma energy.« less

Runaway Electrons Modeling and Nanoparticle Plasma Jet Penetration into Tokamak Plasma

NASA Astrophysics Data System (ADS)

Galkin, S. A.; Bogatu, I. N.

2017-10-01

A novel idea to probe runaway electrons (REs) by superfast injection of high velocity nanoparticle plasma jet (NPPJ) from a plasma accelerator needs to be sustained by both RE dynamics modeling and simulation of NPPJ penetration through increasing tokamak magnetic field. We present our recent progress in both areas. RE simulation is based on the model, including Dreicer and ``avalanche'' mechanisms of RE generation, with emphasis on high Zeff effects. The high-density hyper-velocity C60 and BN NPPJ penetration through transversal B-field is conducted with the Hybrid Electro-Magnetic code (HEM-2D) in cylindrical coordinates, with 1/R B-field dependence for both DIII-D and ITER tokamaks. Work is supported in part by US DOE SBIR Grant.

Current Challenges in the First Principle Quantitative Modelling of the Lower Hybrid Current Drive in Tokamaks

NASA Astrophysics Data System (ADS)

Peysson, Y.; Bonoli, P. T.; Chen, J.; Garofalo, A.; Hillairet, J.; Li, M.; Qian, J.; Shiraiwa, S.; Decker, J.; Ding, B. J.; Ekedahl, A.; Goniche, M.; Zhai, X.

2017-10-01

The Lower Hybrid (LH) wave is widely used in existing tokamaks for tailoring current density profile or extending pulse duration to steady-state regimes. Its high efficiency makes it particularly attractive for a fusion reactor, leading to consider it for this purpose in ITER tokamak. Nevertheless, if basics of the LH wave in tokamak plasma are well known, quantitative modeling of experimental observations based on first principles remains a highly challenging exercise, despite considerable numerical efforts achieved so far. In this context, a rigorous methodology must be carried out in the simulations to identify the minimum number of physical mechanisms that must be considered to reproduce experimental shot to shot observations and also scalings (density, power spectrum). Based on recent simulations carried out for EAST, Alcator C-Mod and Tore Supra tokamaks, the state of the art in LH modeling is reviewed. The capability of fast electron bremsstrahlung, internal inductance li and LH driven current at zero loop voltage to constrain all together LH simulations is discussed, as well as the needs of further improvements (diagnostics, codes, LH model), for robust interpretative and predictive simulations.

Residual zonal flows in tokamaks and stellarators at arbitrary wavelengths

NASA Astrophysics Data System (ADS)

Monreal, Pedro; Calvo, Iván; Sánchez, Edilberto; Parra, Félix I.; Bustos, Andrés; Könies, Axel; Kleiber, Ralf; Görler, Tobias

2016-04-01

In the linear collisionless limit, a zonal potential perturbation in a toroidal plasma relaxes, in general, to a non-zero residual value. Expressions for the residual value in tokamak and stellarator geometries, and for arbitrary wavelengths, are derived. These expressions involve averages over the lowest order particle trajectories, that typically cannot be evaluated analytically. In this work, an efficient numerical method for the evaluation of such expressions is reported. It is shown that this method is faster than direct gyrokinetic simulations performed with the Gene and EUTERPE codes. Calculations of the residual value in stellarators are provided for much shorter wavelengths than previously available in the literature. Electrons must be treated kinetically in stellarators because, unlike in tokamaks, kinetic electrons modify the residual value even at long wavelengths. This effect, that had already been predicted theoretically, is confirmed by gyrokinetic simulations.

Cause and Effect of Feedback: Multiphase Gas in Cluster Cores Heated by AGN Jets

NASA Astrophysics Data System (ADS)

Gaspari, M.; Ruszkowski, M.; Sharma, P.

2012-02-01

Multiwavelength data indicate that the X-ray-emitting plasma in the cores of galaxy clusters is not cooling catastrophically. To a large extent, cooling is offset by heating due to active galactic nuclei (AGNs) via jets. The cool-core clusters, with cooler/denser plasmas, show multiphase gas and signs of some cooling in their cores. These observations suggest that the cool core is locally thermally unstable while maintaining global thermal equilibrium. Using high-resolution, three-dimensional simulations we study the formation of multiphase gas in cluster cores heated by collimated bipolar AGN jets. Our key conclusion is that spatially extended multiphase filaments form only when the instantaneous ratio of the thermal instability and free-fall timescales (t TI/t ff) falls below a critical threshold of ≈10. When this happens, dense cold gas decouples from the hot intracluster medium (ICM) phase and generates inhomogeneous and spatially extended Hα filaments. These cold gas clumps and filaments "rain" down onto the central regions of the core, forming a cold rotating torus and in part feeding the supermassive black hole. Consequently, the self-regulated feedback enhances AGN heating and the core returns to a higher entropy level with t TI/t ff > 10. Eventually, the core reaches quasi-stable global thermal equilibrium, and cold filaments condense out of the hot ICM whenever t TI/t ff <~ 10. This occurs despite the fact that the energy from AGN jets is supplied to the core in a highly anisotropic fashion. The effective spatial redistribution of heat is enabled in part by the turbulent motions in the wake of freely falling cold filaments. Increased AGN activity can locally reverse the cold gas flow, launching cold filamentary gas away from the cluster center. Our criterion for the condensation of spatially extended cold gas is in agreement with observations and previous idealized simulations.

Effects of 2HZ Imposed Bulk Flow Unsteadiness on Laminar/Turbulent Transition in a Straight Channel

DTIC Science & Technology

1989-12-01

near surfaces of operating turbines and airfoils . Non-equilibrium shear layer development near turbine passage surfaces and near airfoils is due to...13A *AbJ’IWH,n SW6 Figure 31. 61 jrr TTyrfljFITt r1I-r-r m~rnrrTrtrfT- rl rti ED 0 ) CDC inO > In- *LLo If.> w -F N E(no: Z 0U 12NWC LE >~ a~mLm z( LL...8217 u’U >4 > D .4 LUJILL"LLLpLLU .L A LJ L i ll ... L.... Figure 61 . 91 (U cu > LO I 4CU E a: (3w a: 4- a:. Qm jL .m I r . wr U > tLLLLLi w co c >3 0A/U

Comparison of 2D simulations of detached divertor plasmas with divertor Thomson measurements in the DIII-D tokamak

DOE PAGES

Rognlien, Thomas D.; McLean, Adam G.; Fenstermacher, Max E.; ...

2017-01-27

A modeling study is reported using new 2D data from DIII-D tokamak divertor plasmas and improved 2D transport model that includes large cross-field drifts for the numerically difficult H-mode regime. The data set, which spans a range of plasmas densities for both forward and reverse toroidal magnetic field (B t) over a range of plasma densities, is provided by divertor Thomson scattering (DTS). Measurements utilizing X-point sweeping give corresponding 2D profiles of electron temperature (T e) and density (n e) across both divertor legs for individual discharges. The calculations show the same features of in/out plasma asymmetries as measured inmore » the experiment, with the normal B t direction (ion ∇B drift toward the X-point) having higher n e and lower T e in the inner divertor leg than outer. Corresponding emission data for total radiated power shows a strong inner-divertor/outer-divertor asymmetry that is reproduced by the simulations. Furthermore, these 2D UEDGE transport simulations are enabled for steep-gradient H-mode conditions by newly implemented algorithms to control isolated grid-scale irregularities.« less

On current drive by Ohkawa mechanism of electron cyclotron wave in large inverse aspect ratio tokamaks

NASA Astrophysics Data System (ADS)

Zheng, Pingwei; Gong, Xueyu; Lu, Xingqiang; He, Lihua; Cao, Jingjia; Huang, Qianhong; Deng, Sheng

2018-03-01

A localized and efficient current drive method in the outer-half region of the tokamak with a large inverse aspect ratio is proposed via the Ohkawa mechanism of electron cyclotron (EC) waves. Further off-axis Ohkawa current drive (OKCD) via EC waves was investigated in high electron beta β e HL-2M-like tokamaks with a large inverse aspect ratio, and in EAST-like tokamaks with a low inverse aspect ratio. OKCD can be driven efficiently, and the driven current profile is spatially localized in the radial region, ranging from 0.62 to 0.85, where the large fraction of trapped electrons provides an excellent advantage for OKCD. Furthermore, the current drive efficiency increases with an increase in minor radius, and then drops when the minor radius beyond a certain value. The effect of trapped electrons greatly enhances the current driving capability of the OKCD mechanism. The highest current drive efficiency can reach 0.183 by adjusting the steering mirror to change the toroidal and poloidal incident angle, and the total driven current by OKCD can reach 20-32 kA MW-1 in HL-2M-like tokamaks. The current drive is less efficient for the EAST-like scenario due to the lower inverse aspect ratio. The results show that OKCD may be a valuable alternative current drive method in large inverse aspect ratio tokamaks, and the potential capabilities of OKCD can be used to suppress some important magnetohydrodynamics instabilities in the far off-axis region.

Measurements of localized core turbulence & turbulence suppression on DIII-D

NASA Astrophysics Data System (ADS)

Shafer, Morgan W.

The crucial dynamics of turbulent-driven cross-field transport in tokamak plasmas reside in the two-dimensional (2D) radial/poloidal plane. Thus, 2D measurements of turbulence are needed to test theoretical models and validate sophisticated gyrokinetic codes. Furthermore, measurements are important for understanding the role of turbulence suppression in enhanced confinement regimes. The Beam Emission Spectroscopy (BES) diagnostic on the DIII-D tokamak measures localized, long-wavelength (k⊥rho i≤1) density fluctuations in the 2D radial/poloidal plane and is suitable for these studies. Measurements of turbulence amplitude, S(kr,k theta) spectra, correlation lengths, decorrelation rates and group velocities are obtained via BES in the core (0.3< r/a <0.9) and compared to nonlinear gyrokinetic simulations from the GYRO code. The 2D measurements show a tilted eddy structure in the core that is consistent with ExB shear. The S(kr,ktheta) spectra are directly compared to GYRO simulations. These comparisons show the 2D structure is in reasonable agreement at r/a = 0.5 where the predicted turbulence amplitude and heat flux agree well with the measurements. However, the simulations show a strongly tilted eddy structure that extends to high-kr at r/a = 0.75, where the simulations under-predict the turbulence amplitude and heat flux. This is not observed in the experiment and suggests a possible over-exaggeration of an ExB or zonal flow shearing mechanism in the simulations. Measurements demonstrate local turbulence suppression near low-order rational q-surfaces at low magnetic shear. This interaction can lead to an Internal Transport Barrier (ITB) provided sufficient equilibrium ExB shear (largely due to the toroidal rotation of neutral beam heated rotating plasmas) sustains the barrier. Related GYRO simulations suggest these ITBs are triggered by zonal flows that form near the q = 2 surface. Consistent with the simulations, localized measurements demonstrate increased

Kinetic simulations of scrape-off layer physics in the DIII-D tokamak

DOE PAGES

Churchill, Randy M.; Canik, John M.; Chang, C. S.; ...

2016-12-27

Simulations using the fully kinetic code XGCa were undertaken to explore the impact of kinetic effects on scrape-off layer (SOL) physics in DIII-D H-mode plasmas. XGCa is a total- f, gyrokinetic code which self-consistently calculates the axisymmetric electrostatic potential and plasma dynamics, and includes modules for Monte Carlo neutral transport. Fluid simulations are normally used to simulate the SOL, due to its high collisionality. However, depending on plasma conditions, a number of discrepancies have been observed between experiment and leading SOL fluid codes (e.g. SOLPS), including underestimating outer target temperatures, radial electric field in the SOL, parallel ion SOL flowsmore » at the low field side, and impurity radiation. Many of these discrepancies may be linked to the fluid treatment, and might be resolved by including kinetic effects in SOL simulations. The XGCa simulation of the DIII-D tokamak in a nominally sheath-limited regime show many noteworthy features in the SOL. The density and ion temperature are higher at the low-field side, indicative of ion orbit loss. The SOL ion Mach flows are at experimentally relevant levels ( Mi ~0.5), with similar shapes and poloidal variation as observed in various tokamaks. Surprisingly, the ion Mach flows close to the sheath edge remain subsonic, in contrast to the typical fluid Bohm criterion requiring ion flows to be above sonic at the sheath edge. Related to this are the presence of elevated sheath potentials, eΔΦ/T e ~ 3–4, over most of the SOL, with regions in the near-SOL close to the separatrix having eΔΦ/Te > 4. Finally, these two results at the sheath edge are a consequence of non-Maxwellian features in the ions and electrons there.« less

Axisymmetric electrostatic magnetohydrodynamic oscillations in tokamaks with general cross-sections and toroidal flow

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

Chu, M. S.; Guo, Wenfeng

2016-06-15

The frequency spectrum and mode structure of axisymmetric electrostatic oscillations [the zonal flow (ZF), sound waves (SW), geodesic acoustic modes (GAM), and electrostatic mean flows (EMF)] in tokamaks with general cross-sections and toroidal flows are studied analytically using the electrostatic approximation for magnetohydrodynamic modes. These modes constitute the “electrostatic continua.” Starting from the energy principle for a tokamak plasma with toroidal rotation, we showed that these modes are completely stable. The ZF, the SW, and the EMF could all be viewed as special cases of the general GAM. The Euler equations for the general GAM are obtained and are solvedmore » analytically for both the low and high range of Mach numbers. The solution consists of the usual countable infinite set of eigen-modes with discrete eigen-frequencies, and two modes with lower frequencies. The countable infinite set is identified with the regular GAM. The lower frequency mode, which is also divergence free as the plasma rotation tends to zero, is identified as the ZF. The other lower (zero) frequency mode is a pure geodesic E×B flow and not divergence free is identified as the EMF. The frequency of the EMF is shown to be exactly 0 independent of plasma cross-section or its flow Mach number. We also show that in general, sound waves with no geodesic components are (almost) completely lost in tokamaks with a general cross-sectional shape. The exception is the special case of strict up-down symmetry. In this case, half of the GAMs would have no geodesic displacements. They are identified as the SW. Present day tokamaks, although not strictly up-down symmetric, usually are only slightly up-down asymmetric. They are expected to share the property with the up-down symmetric tokamak in that half of the GAMs would be more sound wave-like, i.e., have much weaker coupling to the geodesic components than the other half of non-sound-wave-like modes with stronger coupling to the

Core-core and core-valence correlation

NASA Technical Reports Server (NTRS)

Bauschlicher, Charles W., Jr.; Langhoff, Stephen R.; Taylor, Peter R.

1988-01-01

The effect of (1s) core correlation on properties and energy separations was analyzed using full configuration-interaction (FCI) calculations. The Be 1 S - 1 P, the C 3 P - 5 S and CH+ 1 Sigma + or - 1 Pi separations, and CH+ spectroscopic constants, dipole moment and 1 Sigma + - 1 Pi transition dipole moment were studied. The results of the FCI calculations are compared to those obtained using approximate methods. In addition, the generation of atomic natural orbital (ANO) basis sets, as a method for contracting a primitive basis set for both valence and core correlation, is discussed. When both core-core and core-valence correlation are included in the calculation, no suitable truncated CI approach consistently reproduces the FCI, and contraction of the basis set is very difficult. If the (nearly constant) core-core correlation is eliminated, and only the core-valence correlation is included, CASSCF/MRCI approached reproduce the FCI results and basis set contraction is significantly easier.

Neural network evaluation of tokamak current profiles for real time control

NASA Astrophysics Data System (ADS)

Wróblewski, Dariusz

1997-02-01

Active feedback control of the current profile, requiring real-time determination of the current profile parameters, is envisioned for tokamaks operating in enhanced confinement regimes. The distribution of toroidal current in a tokamak is now routinely evaluated based on external (magnetic probes, flux loops) and internal (motional Stark effect) measurements of the poloidal magnetic field. However, the analysis involves reconstruction of magnetohydrodynamic equilibrium and is too intensive computationally to be performed in real time. In the present study, a neural network is used to provide a mapping from the magnetic measurements (internal and external) to selected parameters of the safety factor profile. The single-pass, feedforward calculation of output of a trained neural network is very fast, making this approach particularly suitable for real-time applications. The network was trained on a large set of simulated equilibrium data for the DIII-D tokamak. The database encompasses a large variety of current profiles including the hollow current profiles important for reversed central shear operation. The parameters of safety factor profile (a quantity related to the current profile through the magnetic field tilt angle) estimated by the neural network include central safety factor, q0, minimum value of q, qmin, and the location of qmin. Very good performance of the trained neural network both for simulated test data and for experimental datais demonstrated.

Main-ion intrinsic toroidal rotation profile driven by residual stress torque from ion temperature gradient turbulence in the DIII-D tokamak

DOE PAGES

Grierson, B. A.; Wang, W. X.; Ethier, S.; ...

2017-01-06

Intrinsic toroidal rotation of the deuterium main ions in the core of the DIII-D tokamak is observed to transition from flat to hollow, forming an off-axis peak, above a threshold level of direct electron heating. Nonlinear gyrokinetic simulations show that the residual stress associated with electrostatic ion temperature gradient turbulence possesses the correct radial location and stress structure to cause the observed hollow rotation profile. Residual stress momentum flux in the gyrokinetic simulations is balanced by turbulent momentum diffusion, with negligible contributions from turbulent pinch. Finally, the prediction of the velocity profile by integrating the momentum balance equation produces amore » rotation profile that qualitatively and quantitatively agrees with the measured main-ion profile, demonstrating that fluctuation-induced residual stress can drive the observed intrinsic velocity profile.« less

Tokamak startup using point-source dc helicity injection.

PubMed

Battaglia, D J; Bongard, M W; Fonck, R J; Redd, A J; Sontag, A C

2009-06-05

Startup of a 0.1 MA tokamak plasma is demonstrated on the ultralow aspect ratio Pegasus Toroidal Experiment using three localized, high-current density sources mounted near the outboard midplane. The injected open field current relaxes via helicity-conserving magnetic turbulence into a tokamaklike magnetic topology where the maximum sustained plasma current is determined by helicity balance and the requirements for magnetic relaxation.

Energetic-ion-driven global instabilities in stellarator/helical plasmas and comparison with tokamak plasmas

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

Toi, K.; Ogawa, K.; Isobe, M.

2011-01-01

Comprehensive understanding of energetic-ion-driven global instabilities such as Alfven eigenmodes (AEs) and their impact on energetic ions and bulk plasma is crucially important for tokamak and stellarator/helical plasmas and in the future for deuterium-tritium (DT) burning plasma experiments. Various types of global modes and their associated enhanced energetic ion transport are commonly observed in toroidal plasmas. Toroidicity-induced AEs and ellipticity-induced AEs, whose gaps are generated through poloidal mode coupling, are observed in both tokamak and stellarator/helical plasmas. Global AEs and reversed shear AEs, where toroidal couplings are not as dominant were also observed in those plasmas. Helicity induced AEs thatmore » exist only in 3D plasmas are observed in the large helical device (LHD) and Wendelstein 7 Advanced Stellarator plasmas. In addition, the geodesic acoustic mode that comes from plasma compressibility is destabilized by energetic ions in both tokamak and LHD plasmas. Nonlinear interaction of these modes and their influence on the confinement of the bulk plasma as well as energetic ions are observed in both plasmas. In this paper, the similarities and differences in these instabilities and their consequences for tokamak and stellarator/helical plasmas are summarized through comparison with the data sets obtained in LHD. In particular, this paper focuses on the differences caused by the rotational transform profile and the 2D or 3D geometrical structure of the plasma equilibrium. Important issues left for future study are listed.« less

Identification of a localized core mode in a helicon plasma

NASA Astrophysics Data System (ADS)

Green, Daniel A.; Chakraborty Thakur, Saikat; Tynan, George R.; Light, Adam D.

2017-10-01

We present imaging measurements of a newly observed mode in the core of the Controlled Shear Decorrelation Experiment - Upgrade (CSDX-U). CSDX-U is a well-characterized linear machine producing dense plasmas relevant to the tokamak edge (Te 3 eV, ne 1013 /cc). Typical fluctuations are dominated by electron drift waves, with evidence for Kelvin-Helmholtz vortices appearing near the plasma edge. A new mode has been observed using high-speed imaging that appears at high magnetic field strengths and is confined to the inner third of the plasma column. A cross-spectral phase technique allows direct visualization of dominant spatial structures as a function of frequency. Experimental dispersion curve estimates are constructed from imaging data alone, and allow direct comparison of theoretical dispersion relations to the observed mode. We present preliminary identification of the mode based on its dispersion curve, and compare the results with electrostatic probe measurements.

Kinetic shear Alfvén instability in the presence of impurity ions in tokamak plasmas

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

Lu, Gaimin; Shen, Y.; Xie, T.

2013-10-15

The effects of impurity ions on the kinetic shear Alfvén (KSA) instability in tokamak plasmas are investigated by numerically solving the integral equations for the KSA eigenmode in the toroidal geometry. The kinetic effects of hydrogen and impurity ions, including transit motion, finite ion Larmor radius, and finite-orbit-width, are taken into account. Toroidicity induced linear mode coupling is included through the ballooning-mode representation. Here, the effects of carbon, oxygen, and tungsten ions on the KSA instability in toroidal plasmas are investigated. It is found that, depending on the concentration and density profile of the impurity ions, the latter can bemore » either stabilizing or destabilizing for the KSA modes. The results here confirm the importance of impurity ions in tokamak experiments and should be useful for analyzing experimental data as well as for understanding anomalous transport and control of tokamak plasmas.« less

Development of the radial neutron camera system for the HL-2A tokamak

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

Zhang, Y. P., E-mail: zhangyp@swip.ac.cn; Yang, J. W.; Liu, Yi

2016-06-15

A new radial neutron camera system has been developed and operated recently in the HL-2A tokamak to measure the spatial and time resolved 2.5 MeV D-D fusion neutron, enhancing the understanding of the energetic-ion physics. The camera mainly consists of a multichannel collimator, liquid-scintillation detectors, shielding systems, and a data acquisition system. Measurements of the D-D fusion neutrons using the camera have been successfully performed during the 2015 HL-2A experiment campaign. The measurements show that the distribution of the fusion neutrons in the HL-2A plasma has a peaked profile, suggesting that the neutral beam injection beam ions in the plasmamore » have a peaked distribution. It also suggests that the neutrons are primarily produced from beam-target reactions in the plasma core region. The measurement results from the neutron camera are well consistent with the results of both a standard {sup 235}U fission chamber and NUBEAM neutron calculations. In this paper, the new radial neutron camera system on HL-2A and the first experimental results are described.« less

Can a Penning ionization discharge simulate the tokamak scrape-off plasma conditions?

NASA Technical Reports Server (NTRS)

Finkenthal, M.; Littman, A.; Stutman, D.; Kovnovich, S.; Mandelbaum, P.; Schwob, J. L.; Bhatia, A. K.

1990-01-01

The tokamak scrape-off (the region between the vacuum vessel wall and the magnetically confined fusion plasma edge), represents a source/sink for the hot fusion plasma. The electron densities and temperatures are in the ranges 10 to the 11th - 10 to the 13th/cu cm and 1-40 eV, respectively (depending on the size, magnetic field intensity and configuration, plasma current, etc). In the work reported, the electron temperature and density have been estimated in a Penning ionization discharge by comparing its spectroscopic emission in the VUV with that predicted by a collisional radiative model. An attempt to directly compare this emission with that of the tokamak edge is briefly described.

Measurement of H/D ratio and ion temperature on a HT-6M Tokamak

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

Wei, L.; Lin, X.

1997-01-01

By combining optical fibers with piezoelectric scanning Fabry{endash}Perot interferometer, the profiles of H{sub {alpha}} and D{sub {alpha}} have been determined simultaneously in a single Tokamak discharge. Consequently, the ratio of hydrogen to deuterium and ion temperature are obtained. Not only is the uncertainty of shot-to-shot avoided, the results of the experiment indicate that this instrumentation has the advantage of rapid wavelength scanning, large dispersion, high resolution, and good adaptability of working in adverse circumstances such as at a Tokamak site. {copyright} {ital 1997 American Institute of Physics.}

Gyrokinetic-Vlasov simulations of the ion temperature gradient turbulence in tokamak and helical systems

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

Watanabe, T.-H.; Sugama, H.; Graduate University for Advanced Studies

2006-11-30

Recent progress of the gyrokinetic-Vlasov simulations on the ion temperature gradient (ITG) turbulence in tokamak and helical systems is reported, where the entropy balance is checked as a reference for the numerical accuracy. The tokamak ITG turbulence simulation carried out on the Earth Simulator clearly captures a nonlinear generation process of zonal flows. The tera-flops and tera-bytes scale simulation is also applied to a helical system with the same poloidal and toroidal periodicities of L = 2 and M = 10 as in the Large Helical Device.

Tritium pellet injector for the tokamak fusion test reactor

NASA Astrophysics Data System (ADS)

Gouge, M. J.; Baylor, L. R.; Combs, S. K.; Fisher, P. W.; Foust, C. R.; Milora, S. L.

The tritium pellet injector (TPI) for the Tokamak Fusion Test Reactor (TFTR) will provide a tritium pellet fueling capability with pellet speeds in the 1- to 3-km/s range for the TFTR deuterium-tritium (D-T) plasma phase. An existing deuterium pellet injector (DPI) was modified at Oak Ridge National Laboratory (ORNL) to provide a four-shot, tritium-compatible, pipe-gun configuration with three upgraded single-stage pneumatic guns and a two-stage light gas gun driver. The TPI was designed for frozen pellets ranging in size from 3 to 4 mm in diameter in arbitrarily programmable firing sequences at tritium pellet speeds up to approximately 1.5 km/s for the three single-stage drivers and 2.5 to 3 km/s for the two-stage driver. Injector operation is controlled by a programmable logic controller (PLC). The new pipe-gun injector assembly was installed in the modified DPI guard vacuum box, and modifications were also made to the internals of the DPI vacuum injection line, including a new pellet diagnostics package. Assembly of these modified parts with existing DPI components was then completed and the TPI was tested at ORNL with deuterium pellets. Results of the testing program at ORNL are described. The TPI has been installed and operated on TFTR in support of the FY-92 deuterium plasma run period. In 1993, the tritium pellet injector will be retrofitted with a D-T fuel manifold and tritium gloveboxes and integrated into TFTR tritium processing systems to provide full tritium pellet capability.

RF assisted Glow Discharge Condition experiment for SST-1 Tokamak

NASA Astrophysics Data System (ADS)

Raval, Dilip; Khan, Ziauddin; George, Siju; Dhanani, Kalpeshkumar R.; Paravastu, Yuvakiran; Semwal, Pratibha; Thankey, Prashant; Shoaib Khan, Mohammad; Kakati, Bharat; Pradhan, Subrata

2017-04-01

Impurity control reduces the radiation loss from plasma and hence enhances the plasma operation. Oxygen and water vapors are the most common impurities in tokamak devices. Water vapour can be reduced with extensive baking while in order to have a significant reduction in oxygen it is necessary to use glow discharge condition (GDC). RF assisted glow discharge cleaning system will be implemented to remove low z impurities at PFC installed SST-1 vacuum vessel. A RF assisted Glow discharge conditioning is studied at laboratory to find the optimum operating parameters in a view to implement at SST-1 tokamak. Helium is used as a fuel gas in the present experiment. It is observed that the ultimate impurity level is reduced significantly below to the accepted level for plasma operation after RF assisted GDC. The experimental findings of RF assisted Glow discharge conditioning is discussed in details in this paper.

Neutral-beam deposition in large, finite-beta noncircular tokamak plasmas

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

Wieland, R.M.; Houlberg, W.A.

1982-02-01

A parametric pencil beam model is introduced for describing the attenuation of an energetic neutral beam moving through a tokamak plasma. The nonnegligible effects of a finite beam cross section and noncircular shifted plasma cross sections are accounted for in a simple way by using a smoothing algorithm dependent linearly on beam radius and by including information on the plasma flux surface geometry explicitly. The model is benchmarked against more complete and more time-consuming two-dimensional Monte Carlo calculations for the case of a large D-shaped tokamak plasma with minor radius a = 120 cm and elongation b/a = 1.6. Depositionmore » profiles are compared for deuterium beam energies of 120 to 150 keV, central plasma densities of 8 x 10/sup 13/ - 2 x 10/sup 14/ cm/sup -3/, and beam orientation ranging from perpendicular to tangential to the inside wall.« less

GEM detector development for tokamak plasma radiation diagnostics: SXR poloidal tomography

NASA Astrophysics Data System (ADS)

Chernyshova, Maryna; Malinowski, Karol; Ziółkowski, Adam; Kowalska-Strzeciwilk, Ewa; Czarski, Tomasz; Poźniak, Krzysztof T.; Kasprowicz, Grzegorz; Zabołotny, Wojciech; Wojeński, Andrzej; Kolasiński, Piotr; Krawczyk, Rafał D.

2015-09-01

An increased attention to tungsten material is related to a fact that it became a main candidate for the plasma facing material in ITER and future fusion reactor. The proposed work refers to the studies of W influence on the plasma performances by developing new detectors based on Gas Electron Multiplier GEM) technology for tomographic studies of tungsten transport in ITER-oriented tokamaks, e.g. WEST project. It presents current stage of design and developing of cylindrically bent SXR GEM detector construction for horizontal port implementation. Concept to overcome an influence of constraints on vertical port has been also presented. It is expected that the detecting unit under development, when implemented, will add to the safe operation of tokamak bringing creation of sustainable nuclear fusion reactors a step closer.

Dynamic optimization of open-loop input signals for ramp-up current profiles in tokamak plasmas

NASA Astrophysics Data System (ADS)

Ren, Zhigang; Xu, Chao; Lin, Qun; Loxton, Ryan; Teo, Kok Lay

2016-03-01

Establishing a good current spatial profile in tokamak fusion reactors is crucial to effective steady-state operation. The evolution of the current spatial profile is related to the evolution of the poloidal magnetic flux, which can be modeled in the normalized cylindrical coordinates using a parabolic partial differential equation (PDE) called the magnetic diffusion equation. In this paper, we consider the dynamic optimization problem of attaining the best possible current spatial profile during the ramp-up phase of the tokamak. We first use the Galerkin method to obtain a finite-dimensional ordinary differential equation (ODE) model based on the original magnetic diffusion PDE. Then, we combine the control parameterization method with a novel time-scaling transformation to obtain an approximate optimal parameter selection problem, which can be solved using gradient-based optimization techniques such as sequential quadratic programming (SQP). This control parameterization approach involves approximating the tokamak input signals by piecewise-linear functions whose slopes and break-points are decision variables to be optimized. We show that the gradient of the objective function with respect to the decision variables can be computed by solving an auxiliary dynamic system governing the state sensitivity matrix. Finally, we conclude the paper with simulation results for an example problem based on experimental data from the DIII-D tokamak in San Diego, California.

Radial force on the vacuum chamber wall during thermal quench in tokamaks

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

Pustovitov, V. D., E-mail: pustovitov-vd@nrcki.ru

The radial force balance during a thermal quench in tokamaks is analyzed. As a rule, the duration τ{sub tp} of such events is much shorter than the resistive time τ{sub w} of the vacuum chamber wall. Therefore, the perturbations of the magnetic field B produced by the evolving plasma cannot penetrate the wall, which makes different the magnetic pressures on its inner and outer sides. The goal of this work is the analytical estimation of the resulting integral radial force on the wall. The plasma is considered axially symmetric; for the description of radial forces on the wall, the resultsmore » of V.D. Shafranov’s classical work [J. Nucl. Energy C 5, 251 (1963)] are used. Developed for tokamaks, the standard equilibrium theory considers three interacting systems: plasma, poloidal field coils, and toroidal field coils. Here, the wall is additionally incorporated with currents driven by ∂B/∂t≠0 accompanying the fast loss of the plasma thermal energy. It is shown that they essentially affect the force redistribution, thereby leading to large loads on the wall. The estimates prove that these loads have to be accounted for in the disruptive scenarios in large tokamaks.« less

The influence of microvascular injury on native T1 and T2* relaxation values after acute myocardial infarction: implications for non-contrast-enhanced infarct assessment.

PubMed

Robbers, Lourens F H J; Nijveldt, Robin; Beek, Aernout M; Teunissen, Paul F A; Hollander, Maurits R; Biesbroek, P Stefan; Everaars, Henk; van de Ven, Peter M; Hofman, Mark B M; van Royen, Niels; van Rossum, Albert C

2018-02-01

Native T1 mapping and late gadolinium enhancement (LGE) imaging offer detailed characterisation of the myocardium after acute myocardial infarction (AMI). We evaluated the effects of microvascular injury (MVI) and intramyocardial haemorrhage on local T1 and T2* values in patients with a reperfused AMI. Forty-three patients after reperfused AMI underwent cardiovascular magnetic resonance imaging (CMR) at 4 [3-5] days, including native MOLLI T1 and T2* mapping, STIR, cine imaging and LGE. T1 and T2* values were determined in LGE-defined regions of interest: the MI core incorporating MVI when present, the core-adjacent MI border zone (without any areas of MVI), and remote myocardium. Average T1 in the MI core was higher than in the MI border zone and remote myocardium. However, in the 20 (47%) patients with MVI, MI core T1 was lower than in patients without MVI (MVI 1048±78ms, no MVI 1111±89ms, p=0.02). MI core T2* was significantly lower in patients with MVI than in those without (MVI 20 [18-23]ms, no MVI 31 [26-39]ms, p<0.001). The presence of MVI profoundly affects MOLLI-measured native T1 values. T2* mapping suggested that this may be the result of intramyocardial haemorrhage. These findings have important implications for the interpretation of native T1 values shortly after AMI. • Microvascular injury after acute myocardial infarction affects local T1 and T2* values. • Infarct zone T1 values are lower if microvascular injury is present. • T2* mapping suggests that low infarct T1 values are likely haemorrhage. • T1 and T2* values are complimentary for correctly assessing post-infarct myocardium.

Just-in-time vaccines: Biomineralized calcium phosphate core-immunogen shell nanoparticles induce long-lasting CD8+ T cell responses in mice

PubMed Central

Zhou, Weibin; Moguche, Albanus; Chiu, David; Murali-Krishna, Kaja; Baneyx, François

2014-01-01

Distributed and on-demand vaccine production could be game-changing for infectious disease treatment in the developing world by providing new therapeutic opportunities and breaking the refrigeration “cold chain”. Here, we show that a fusion protein between a calcium phosphate binding domain and the model antigen ovalbumin can mineralize a biocompatible adjuvant in a single step. The resulting 50 nm calcium phosphate core-immunogen shell particles are comparable to soluble protein in inducing ovalbumin-specific antibody response and class switch recombination in mice. However, single dose vaccination with nanoparticles leads to higher expansion of ovalbumin-specific CD8+ T cells upon challenge with an influenza virus bearing the ovalbumin-derived SIINFEKL peptide, and these cells produce high levels of IFN-γ. Furthermore, mice exhibit a robust antigen-specific CD8+ T cell recall response when challenged with virus 8 months post-immunization. These results underscore the promise of immunogen-controlled adjuvant mineralization for just-in-time manufacturing of effective T cell vaccines. PMID:24275478

Core shift effect in blazars

NASA Astrophysics Data System (ADS)

Agarwal, A.; Mohan, P.; Gupta, Alok C.; Mangalam, A.; Volvach, A. E.; Aller, M. F.; Aller, H. D.; Gu, M. F.; Lähteenmäki, A.; Tornikoski, M.; Volvach, L. N.

2017-07-01

We studied the pc-scale core shift effect using radio light curves for three blazars, S5 0716+714, 3C 279 and BL Lacertae, which were monitored at five frequencies (ν) between 4.8 and 36.8 GHz using the University of Michigan Radio Astronomical Observatory (UMRAO), the Crimean Astrophysical Observatory (CrAO) and Metsähovi Radio Observatory for over 40 yr. Flares were Gaussian fitted to derive time delays between observed frequencies for each flare (Δt), peak amplitude (A) and their half width. Using A ∝ να, we infer α in the range of -16.67-2.41 and using Δ t ∝ ν ^{1/k_r}, we infer kr ∼ 1, employed in the context of equipartition between magnetic and kinetic energy density for parameter estimation. From the estimated core position offset (Ωrν) and the core radius (rcore), we infer that opacity model may not be valid in all cases. The mean magnetic field strengths at 1 pc (B1) and at the core (Bcore) are in agreement with previous estimates. We apply the magnetically arrested disc model to estimate black hole spins in the range of 0.15-0.9 for these blazars, indicating that the model is consistent with expected accretion mode in such sources. The power-law-shaped power spectral density has slopes -1.3 to -2.3 and is interpreted in terms of multiple shocks or magnetic instabilities.

Fast torsional waves and strong magnetic field within the Earth's core.

PubMed

Gillet, Nicolas; Jault, Dominique; Canet, Elisabeth; Fournier, Alexandre

2010-05-06

The magnetic field inside the Earth's fluid and electrically conducting outer core cannot be directly probed. The root-mean-squared (r.m.s.) intensity for the resolved part of the radial magnetic field at the core-mantle boundary is 0.3 mT, but further assumptions are needed to infer the strength of the field inside the core. Recent diagnostics obtained from numerical geodynamo models indicate that the magnitude of the dipole field at the surface of a fluid dynamo is about ten times weaker than the r.m.s. field strength in its interior, which would yield an intensity of the order of several millitesla within the Earth's core. However, a 60-year signal found in the variation in the length of day has long been associated with magneto-hydrodynamic torsional waves carried by a much weaker internal field. According to these studies, the r.m.s. strength of the field in the cylindrical radial direction (calculated for all length scales) is only 0.2 mT, a figure even smaller than the r.m.s. strength of the large-scale (spherical harmonic degree n core-mantle boundary. Here we reconcile numerical geodynamo models with studies of geostrophic motions in the Earth's core that rely on geomagnetic data. From an ensemble inversion of core flow models, we find a torsional wave recurring every six years, the angular momentum of which accounts well for both the phase and the amplitude of the six-year signal for change in length of day detected over the second half of the twentieth century. It takes about four years for the wave to propagate throughout the fluid outer core, and this travel time translates into a slowness for Alfvén waves that corresponds to a r.m.s. field strength in the cylindrical radial direction of approximately 2 mT. Assuming isotropy, this yields a r.m.s. field strength of 4 mT inside the Earth's core.

Accelerator Facilities for Radiation Research

NASA Technical Reports Server (NTRS)

Cucinotta, Francis A.

1999-01-01

HSRP Goals in Accelerator Use and Development are: 1.Need for ground-based heavy ion and proton facility to understand space radiation effects discussed most recently by NAS/NRC Report (1996). 2. Strategic Program Goals in facility usage and development: -(1) operation of AGS for approximately 600 beam hours/year; (2) operation of Loma Linda University (LLU) proton facility for approximately 400 beam hours/year; (3) construction of BAF facility; and (4) collaborative research at HIMAC in Japan and with other existing or potential international facilities. 3. MOA with LLU has been established to provide proton beams with energies of 40-250 important for trapped protons and solar proton events. 4. Limited number of beam hours available at Brookhaven National Laboratory's (BNL) Alternating Gradient Synchrotron (AGS).

Control system of neoclassical tearing modes in real time on HL-2A tokamak.

PubMed

Yan, Longwen; Ji, Xiaoquan; Song, Shaodong; Xia, Fan; Xu, Yuan; Ye, Jiruo; Jiang, Min; Chen, Wenjin; Sun, Tengfei; Liang, Shaoyong; Ling, Fei; Ma, Rui; Huang, Mei; Qu, Hongpeng; Song, Xianming; Yu, Deliang; Shi, Zhongbin; Liu, Yi; Yang, Qingwei; Xu, Min; Duan, Xuru; Liu, Yong

2017-11-01

The stability and performance of tokamak plasmas are routinely limited by various magneto-hydrodynamic instabilities, such as neoclassical tearing modes (NTMs). This paper presents a rather simple method to control the NTMs in real time (RT) on a tokamak, including the control principle of a feedback approach for RT suppression and stabilization for the NTMs. The control system combines Mirnov, electron cyclotron emission, and soft X-ray diagnostics used for determining the NTM positions. A methodology for fast detection of 2/1 or 3/2 NTM positions with 129 × 129 grid reconstruction is elucidated. The forty poloidal angles for steering the electron cyclotron resonance heating (ECRH)/electron cyclotron current drive launcher are used to establish the alignment of antenna mirrors with the center of the NTM and to ensure launcher emission intersecting with the rational surface of a magnetic island. Pilot experiments demonstrate the RT control capability to trace the conventional tearing modes (CTMs) in the HL-2A tokamak. The 2/1 CTMs have been suppressed or stabilized by the ECRH power deposition on site or with the steerable launcher.

Control system of neoclassical tearing modes in real time on HL-2A tokamak

NASA Astrophysics Data System (ADS)

Yan, Longwen; Ji, Xiaoquan; Song, Shaodong; Xia, Fan; Xu, Yuan; Ye, Jiruo; Jiang, Min; Chen, Wenjin; Sun, Tengfei; Liang, Shaoyong; Ling, Fei; Ma, Rui; Huang, Mei; Qu, Hongpeng; Song, Xianming; Yu, Deliang; Shi, Zhongbin; Liu, Yi; Yang, Qingwei; Xu, Min; Duan, Xuru; Liu, Yong

2017-11-01

The stability and performance of tokamak plasmas are routinely limited by various magneto-hydrodynamic instabilities, such as neoclassical tearing modes (NTMs). This paper presents a rather simple method to control the NTMs in real time (RT) on a tokamak, including the control principle of a feedback approach for RT suppression and stabilization for the NTMs. The control system combines Mirnov, electron cyclotron emission, and soft X-ray diagnostics used for determining the NTM positions. A methodology for fast detection of 2/1 or 3/2 NTM positions with 129 × 129 grid reconstruction is elucidated. The forty poloidal angles for steering the electron cyclotron resonance heating (ECRH)/electron cyclotron current drive launcher are used to establish the alignment of antenna mirrors with the center of the NTM and to ensure launcher emission intersecting with the rational surface of a magnetic island. Pilot experiments demonstrate the RT control capability to trace the conventional tearing modes (CTMs) in the HL-2A tokamak. The 2/1 CTMs have been suppressed or stabilized by the ECRH power deposition on site or with the steerable launcher.

Measurement of LHCD antenna position in Aditya tokamak

NASA Astrophysics Data System (ADS)

Ambulkar, K. K.; Sharma, P. K.; Virani, C. G.; Parmar, P. R.; Thakur, A. L.; Kulkarni, S. V.

2010-02-01

To drive plasma current non-inductively in ADITYA tokamak, 120 kW pulsed Lower Hybrid Current Drive (LHCD) system at 3.7 GHz has been designed, fabricated and installed on ADITYA tokamak. In this system, the antenna consists of a grill structure, having two rows, each row comprising of four sub-waveguides. The coupling of LHCD power to the plasma strongly depends on the plasma density near the mouth of grill antenna. Thus the grill antenna has to be precisely positioned for efficient coupling. The movement of mechanical bellow, which contracts or expands up to 50mm, governs the movement of antenna. In order to monitor the position of the antenna precisely, the reference position of the antenna with respect to the machine/plasma position has to be accurately determined. Further a mechanical system or an electronic system to measure the relative movement of the antenna with respect to the reference position is also desired. Also due to poor accessibility inside the ADITYA machine, it is impossible to measure physically the reference position of the grill antenna with respect to machine wall, taken as reference position and hence an alternative method has to be adopted to establish these measurements reliably. In this paper we report the design and development of a mechanism, using which the antenna position measurements are made. It also describes a unique method employing which the measurements of the reference position of the antenna with respect to the inner edge of the tokamak wall is carried out, which otherwise was impossible due to poor accessibility and physical constraints. The position of the antenna is monitored using an electronic scale, which is developed and installed on the bellow. Once the reference position is derived, the linear potentiometer, attached to the bellow, measures the linear distance using position transmitter. The accuracy of measurement obtained in our setup is within +/- 0.5 % and the linearity, along with repeatability is excellent.

Overview of physics research on the TCV tokamak

NASA Astrophysics Data System (ADS)

Fasoli, A.; TCV Team

2009-10-01

The Tokamak à Configuration Variable (TCV) tokamak is equipped with high-power (4.5 MW), real-time-controllable EC systems and flexible shaping, and plays an important role in fusion research by broadening the parameter range of reactor relevant regimes, by investigating tokamak physics questions and by developing new control tools. Steady-state discharges are achieved, in which the current is entirely self-generated through the bootstrap mechanism, a fundamental ingredient for ITER steady-state operation. The discharge remains quiescent over several current redistribution times, demonstrating that a self-consistent, 'bootstrap-aligned' equilibrium state is possible. Electron internal transport barrier regimes sustained by EC current drive have also been explored. MHD activity is shown to be crucial in scenarios characterized by large and slow oscillations in plasma confinement, which in turn can be modified by small Ohmic current perturbations altering the barrier strength. In studies of the relation between anomalous transport and plasma shape, the observed dependences of the electron thermal diffusivity on triangularity (direct) and collisionality (inverse) are qualitatively reproduced by non-linear gyro-kinetic simulations and shown to be governed by TEM turbulence. Parallel SOL flows are studied for their importance for material migration. Flow profiles are measured using a reciprocating Mach probe by changing from lower to upper single-null diverted equilibria and shifting the plasmas vertically. The dominant, field-direction-dependent Pfirsch-Schlüter component is found to be in good agreement with theoretical predictions. A field-direction-independent component is identified and is consistent with flows generated by transient over-pressure due to ballooning-like interchange turbulence. Initial high-resolution infrared images confirm that ELMs have a filamentary structure, while fast, localized radiation measurements reveal that ELM activity first appears

Ignition in tokamaks with modulated source of auxiliary heating

NASA Astrophysics Data System (ADS)

Morozov, D. Kh

2017-12-01

It is shown that the ignition may be achieved in tokamaks with the modulated power source. The time-averaged source power may be smaller than the steady-state source power, which is sufficient for the ignition. Nevertheless, the maximal power must be large enough, because the ignition must be achieved within a finite time interval.

Evaluation of two cold thermoregulatory models for prediction of core temperature during exercise in cold water.

PubMed

Castellani, John W; O'Brien, Catherine; Tikuisis, Peter; Sils, Ingrid V; Xu, Xiaojiang

2007-12-01

Cold thermoregulatory models (CTM) have primarily been developed to predict core temperature (T(core)) responses during sedentary immersion. Few studies have examined their efficacy to predict T(core) during exercise cold exposure. The purpose of this study was to compare observed T(core) responses during exercise in cold water with the predicted T(core) from a three-cylinder (3-CTM) and a six-cylinder (6-CTM) model, adjusted to include heat production from exercise. A matrix of two metabolic rates (0.44 and 0.88 m/s walking), two water temperatures (10 and 15 degrees C), and two immersion depths (chest and waist) were used to elicit different rates of T(core) changes. Root mean square deviation (RMSD) and nonparametric Bland-Altman tests were used to test for acceptable model predictions. Using the RMSD criterion, the 3-CTM did not fit the observed data in any trial, whereas the 6-CTM fit the data (RMSD less than standard deviation) in four of eight trials. In general, the 3-CTM predicted a rapid decline in core temperature followed by a plateau. For the 6-CTM, the predicted T(core) appeared relatively tight during the early part of immersion, but was much lower during the latter portions of immersion, accounting for the nonagreement between RMSD and SD values. The 6-CTM was rerun with no adjustment for exercise metabolism, and core temperature and heat loss predictions were tighter. In summary, this study demonstrated that both thermoregulatory models designed for sedentary cold exposure, currently, cannot be extended for use during partial immersion exercise in cold water. Algorithms need to be developed to better predict heat loss during exercise in cold water.

Electromagnetic banana kinetic equation and its applications in tokamaks

NASA Astrophysics Data System (ADS)

Shaing, K. C.; Chu, M. S.; Sabbagh, S. A.; Seol, J.

2018-03-01

A banana kinetic equation in tokamaks that includes effects of the finite banana width is derived for the electromagnetic waves with frequencies lower than the gyro-frequency and the bounce frequency of the trapped particles. The radial wavelengths are assumed to be either comparable to or shorter than the banana width, but much wider than the gyro-radius. One of the consequences of the banana kinetics is that the parallel component of the vector potential is not annihilated by the orbit averaging process and appears in the banana kinetic equation. The equation is solved to calculate the neoclassical quasilinear transport fluxes in the superbanana plateau regime caused by electromagnetic waves. The transport fluxes can be used to model electromagnetic wave and the chaotic magnetic field induced thermal particle or energetic alpha particle losses in tokamaks. It is shown that the parallel component of the vector potential enhances losses when it is the sole transport mechanism. In particular, the fact that the drift resonance can cause significant transport losses in the chaotic magnetic field in the hitherto unknown low collisionality regimes is emphasized.

The GBS code for tokamak scrape-off layer simulations

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

Halpern, F.D., E-mail: federico.halpern@epfl.ch; Ricci, P.; Jolliet, S.

2016-06-15

We describe a new version of GBS, a 3D global, flux-driven plasma turbulence code to simulate the turbulent dynamics in the tokamak scrape-off layer (SOL), superseding the code presented by Ricci et al. (2012) [14]. The present work is driven by the objective of studying SOL turbulent dynamics in medium size tokamaks and beyond with a high-fidelity physics model. We emphasize an intertwining framework of improved physics models and the computational improvements that allow them. The model extensions include neutral atom physics, finite ion temperature, the addition of a closed field line region, and a non-Boussinesq treatment of the polarizationmore » drift. GBS has been completely refactored with the introduction of a 3-D Cartesian communicator and a scalable parallel multigrid solver. We report dramatically enhanced parallel scalability, with the possibility of treating electromagnetic fluctuations very efficiently. The method of manufactured solutions as a verification process has been carried out for this new code version, demonstrating the correct implementation of the physical model.« less

Ice cores and SeaRISE: What we do (and don't) know

NASA Technical Reports Server (NTRS)

Alley, Richard B.

1991-01-01

Ice core analyses are needed in SeaRISE to learn what the West Antarctic ice sheet and other marine ice sheets were like in the past, what climate changes led to their present states, and how they behave. The major results of interest to SeaRISE from previous ice core analyses in West Antarctic are that the end of the last ice age caused temperature and accumulation rate increases in inland regions, leading to ice sheet thickening followed by thinning to the present.

γδ T cells producing interleukin-17A regulate adipose regulatory T cell homeostasis and thermogenesis.

PubMed

Kohlgruber, Ayano C; Gal-Oz, Shani T; LaMarche, Nelson M; Shimazaki, Moto; Duquette, Danielle; Nguyen, Hung N; Mina, Amir I; Paras, Tyler; Tavakkoli, Ali; von Andrian, Ulrich; Banks, Alexander S; Shay, Tal; Brenner, Michael B; Lynch, Lydia

2018-05-01

γδ T cells are situated at barrier sites and guard the body from infection and damage. However, little is known about their roles outside of host defense in nonbarrier tissues. Here, we characterize a highly enriched tissue-resident population of γδ T cells in adipose tissue that regulate age-dependent regulatory T cell (T reg ) expansion and control core body temperature in response to environmental fluctuations. Mechanistically, innate PLZF + γδ T cells produced tumor necrosis factor and interleukin (IL) 17 A and determined PDGFRα + and Pdpn + stromal-cell production of IL-33 in adipose tissue. Mice lacking γδ T cells or IL-17A exhibited decreases in both ST2 + T reg cells and IL-33 abundance in visceral adipose tissue. Remarkably, these mice also lacked the ability to regulate core body temperature at thermoneutrality and after cold challenge. Together, these findings uncover important physiological roles for resident γδ T cells in adipose tissue immune homeostasis and body-temperature control.

A frequency tunable, eight-channel correlation ECE system for electron temperature turbulence measurements on the DIII-D tokamak

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

Sung, C., E-mail: csung@physics.ucla.edu; Peebles, W. A.; Wannberg, C.

2016-11-15

A new eight-channel correlation electron cyclotron emission diagnostic has recently been installed on the DIII-D tokamak to study both turbulent and coherent electron temperature fluctuations under various plasma conditions and locations. This unique system is designed to cover a broad range of operation space on DIII-D (1.6-2.1 T, detection frequency: 72-108 GHz) via four remotely selected local oscillators (80, 88, 96, and 104 GHz). Eight radial locations are measured simultaneously in a single discharge covering as much as half the minor radius. In this paper, we present design details of the quasi-optical system, the receiver, as well as representative datamore » illustrating operation of the system.« less

Adaptive optimal stochastic state feedback control of resistive wall modes in tokamaks

NASA Astrophysics Data System (ADS)

Sun, Z.; Sen, A. K.; Longman, R. W.

2006-01-01

An adaptive optimal stochastic state feedback control is developed to stabilize the resistive wall mode (RWM) instability in tokamaks. The extended least-square method with exponential forgetting factor and covariance resetting is used to identify (experimentally determine) the time-varying stochastic system model. A Kalman filter is used to estimate the system states. The estimated system states are passed on to an optimal state feedback controller to construct control inputs. The Kalman filter and the optimal state feedback controller are periodically redesigned online based on the identified system model. This adaptive controller can stabilize the time-dependent RWM in a slowly evolving tokamak discharge. This is accomplished within a time delay of roughly four times the inverse of the growth rate for the time-invariant model used.

Adaptive Optimal Stochastic State Feedback Control of Resistive Wall Modes in Tokamaks

NASA Astrophysics Data System (ADS)

Sun, Z.; Sen, A. K.; Longman, R. W.

2007-06-01

An adaptive optimal stochastic state feedback control is developed to stabilize the resistive wall mode (RWM) instability in tokamaks. The extended least square method with exponential forgetting factor and covariance resetting is used to identify the time-varying stochastic system model. A Kalman filter is used to estimate the system states. The estimated system states are passed on to an optimal state feedback controller to construct control inputs. The Kalman filter and the optimal state feedback controller are periodically redesigned online based on the identified system model. This adaptive controller can stabilize the time dependent RWM in a slowly evolving tokamak discharge. This is accomplished within a time delay of roughly four times the inverse of the growth rate for the time-invariant model used.

Collisionless microtearing modes in hot tokamaks: Effect of trapped electrons

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

Swamy, Aditya K.; Ganesh, R., E-mail: ganesh@ipr.res.in; Brunner, S.

2015-07-15

Collisionless microtearing modes have recently been found linearly unstable in sharp temperature gradient regions of large aspect ratio tokamaks. The magnetic drift resonance of passing electrons has been found to be sufficient to destabilise these modes above a threshold plasma β. A global gyrokinetic study, including both passing electrons as well as trapped electrons, shows that the non-adiabatic contribution of the trapped electrons provides a resonant destabilization, especially at large toroidal mode numbers, for a given aspect ratio. The global 2D mode structures show important changes to the destabilising electrostatic potential. The β threshold for the onset of the instabilitymore » is found to be generally downshifted by the inclusion of trapped electrons. A scan in the aspect ratio of the tokamak configuration, from medium to large but finite values, clearly indicates a significant destabilizing contribution from trapped electrons at small aspect ratio, with a diminishing role at larger aspect ratios.« less

Hybrid model for simulation of plasma jet injection in tokamak

NASA Astrophysics Data System (ADS)

Galkin, Sergei A.; Bogatu, I. N.

2016-10-01

Hybrid kinetic model of plasma treats the ions as kinetic particles and the electrons as charge neutralizing massless fluid. The model is essentially applicable when most of the energy is concentrated in the ions rather than in the electrons, i.e. it is well suited for the high-density hyper-velocity C60 plasma jet. The hybrid model separates the slower ion time scale from the faster electron time scale, which becomes disregardable. That is why hybrid codes consistently outperform the traditional PIC codes in computational efficiency, still resolving kinetic ions effects. We discuss 2D hybrid model and code with exact energy conservation numerical algorithm and present some results of its application to simulation of C60 plasma jet penetration through tokamak-like magnetic barrier. We also examine the 3D model/code extension and its possible applications to tokamak and ionospheric plasmas. The work is supported in part by US DOE DE-SC0015776 Grant.

An efficient transport solver for tokamak plasmas

DOE PAGES

Park, Jin Myung; Murakami, Masanori; St. John, H. E.; ...

2017-01-03

A simple approach to efficiently solve a coupled set of 1-D diffusion-type transport equations with a stiff transport model for tokamak plasmas is presented based on the 4th order accurate Interpolated Differential Operator scheme along with a nonlinear iteration method derived from a root-finding algorithm. Here, numerical tests using the Trapped Gyro-Landau-Fluid model show that the presented high order method provides an accurate transport solution using a small number of grid points with robust nonlinear convergence.

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.

Lower hybrid wave edge power loss quantification on the Alcator C-Mod tokamak

NASA Astrophysics Data System (ADS)

Faust, I. C.; Brunner, D.; LaBombard, B.; Parker, R. R.; Terry, J. L.; Whyte, D. G.; Baek, S. G.; Edlund, E.; Hubbard, A. E.; Hughes, J. W.; Kuang, A. Q.; Reinke, M. L.; Shiraiwa, S.; Wallace, G. M.; Walk, J. R.

2016-05-01

For the first time, the power deposition of lower hybrid RF waves into the edge plasma of a diverted tokamak has been systematically quantified. Edge deposition represents a parasitic loss of power that can greatly impact the use and efficiency of Lower Hybrid Current Drive (LHCD) at reactor-relevant densities. Through the use of a unique set of fast time resolution edge diagnostics, including innovative fast-thermocouples, an extensive set of Langmuir probes, and a Lyα ionization camera, the toroidal, poloidal, and radial structure of the power deposition has been simultaneously determined. Power modulation was used to directly isolate the RF effects due to the prompt ( t < τ E ) response of the scrape-off-layer (SOL) plasma to Lower Hybrid Radiofrequency (LHRF) power. LHRF power was found to absorb more strongly in the edge at higher densities. It is found that a majority of this edge-deposited power is promptly conducted to the divertor. This correlates with the loss of current drive efficiency at high density previously observed on Alcator C-Mod, and displaying characteristics that contrast with the local RF edge absorption seen on other tokamaks. Measurements of ionization in the active divertor show dramatic changes due to LHRF power, implying that divertor region can be a key for the LHRF edge power deposition physics. These observations support the existence of a loss mechanism near the edge for LHRF at high density ( n e > 1.0 × 10 20 (m-3)). Results will be shown addressing the distribution of power within the SOL, including the toroidal symmetry and radial distribution. These characteristics are important for deducing the cause of the reduced LHCD efficiency at high density and motivate the tailoring of wave propagation to minimize SOL interaction, for example, through the use of high-field-side launch.

Design of geometric phase measurement in EAST Tokamak

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

Lan, T.; Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui 230031; Liu, H. Q., E-mail: hqliu@ipp.ac.cn

2016-07-15

The optimum scheme for geometric phase measurement in EAST Tokamak is proposed in this paper. The theoretical values of geometric phase for the probe beams of EAST Polarimeter-Interferometer (POINT) system are calculated by path integration in parameter space. Meanwhile, the influences of some controllable parameters on geometric phase are evaluated. The feasibility and challenge of distinguishing geometric effect in the POINT signal are also assessed in detail.

Electron Injection by E-Field Drift and its Application in Starting-up Tokamaks at Low Loop Voltage

NASA Astrophysics Data System (ADS)

Pan, Yuan; Yan, Xiao-Lin; Liu, Bao-Hua

2003-05-01

We propose an innovative method of electron injection by E-field drift into a plasma device and discuss its application in starting-up tokamak plasmas at low loop voltage. The experimental results obtained from HT-6M Tokamak are also presented. The breakdown loop voltage is obviously reduced and the discharge performance is improved by using the electron injection method. It could be applied to some other types of plasma device.

Development of a Magnetic-Core, Transverse-Field AF Demagnetizer

NASA Astrophysics Data System (ADS)

Schillinger, W. E.; Morris, E. R.; Coe, R. S.; Finn, D. R.

2016-12-01

A standard cleaning technique in the study of a rock's natural remanent magnetization (NRM) is progressive Alternating Field Demagnetization (AFD). However, for a significant fraction of samples, demagnetization is not completed by the maximum field of 200 mT or less available in commercial instruments; a field at least two or three times higher is needed. The data from 0 to 160 mT for a resistant red bed sample from Tibet is shown below. It just starts to reveal the sample's characteristic component, but this interpretation would have been tenuous, since 85% of the NRM remained untouched. Continued demagnetization to 500 mT helps a great deal, reducing the NRM to just 30% of its initial value and proving that the segment from 160 to 500 mT indeed trends toward the origin. We have constructed an alternating field (AF) demagnetizer that can routinely operate at fields of up to 0.6 Tesla. It uses a magnetic core in an air-cooled coil and is compatible with our existing sample handler for automated demagnetization and measurement experiments. Nonlinearities of the magnetic core are not a significant problem; even harmonics of the magnetic field are ≤1 ppm of the fundamental and so generate negligible anhysteretic remanence. A surprising result during the testing was that the coil's inductance changed with magnetic field. This made it necessary to add an auto-tuning feature, to keep the drive's frequency on the coil's resonance. We have recently added the ability to include a DC field of up to 0.5 mT, parallel to the alternating field, to perform Anhysteretic Remanent Magnetization (ARM), partial ARM experiments and anisotropy of ARM. We will report on these ARM results at the AGU meeting. Currently the maximum field we can obtain is 600 mT, but by reshaping the core to minimize flux leakage, significantly higher fields should be attainable, since the saturation flux density of the core material is 1.5T.

Balanced Technology Initiative Briefing to Industry.

DTIC Science & Technology

1987-08-26

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Final Environmental Impact Statement for the Kauai Acoustic Thermometry of Ocean Climate Project and its associated Marine Mammal Research Program: Vol 2

DTIC Science & Technology

1995-05-01

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Empirical and Hierarchical Bayes Competitors of Preliminary Test Estimators in Two Sample Problems.

DTIC Science & Technology

1987-09-01

wy.; 7 -J-.j V JF. ~LLU4II UnLclassified AIVW,,I.. REPORT DOCUMENTATION PAGE READ INSTRUCTIONS IIE*ONE CMP1LE’rIN; F’ORM I EP’) RT NUMBER GOVT...V2, I +)," P - - ("., "’-’ ?12, (24.5) V f " * .,(., ...... *%*)*v,"*,, ., + .(, -; ( . . D 1 =v +0 rt ’??2v2-t 1 (4 .7R ) % We now write G’ as (;I I...forimt. R - (ek(,F) / F) ( , , W ) Consider for exam plec the sit . ’mat .ion w hen o - 0 I.e., I? has the Improper prior /1 2(r) - r’ - I. Now w

Neural network evaluation of tokamak current profiles for real time control (abstract)

NASA Astrophysics Data System (ADS)

Wróblewski, Dariusz

1997-01-01

Active feedback control of the current profile, requiring real-time determination of the current profile parameters, is envisioned for tokamaks operating in enhanced confinement regimes. The distribution of toroidal current in a tokamak is now routinely evaluated based on external (magnetic probes, flux loops) and internal (motional Stark effect) measurements of the poloidal magnetic field. However, the analysis involves reconstruction of magnetohydrodynamic equilibrium and is too intensive computationally to be performed in real time. In the present study, a neural network is used to provide a mapping from the magnetic measurements (internal and external) to selected parameters of the safety factor profile. The single-pass, feedforward calculation of output of a trained neural network is very fast, making this approach particularly suitable for real-time applications. The network was trained on a large set of simulated equilibrium data for the DIII-D tokamak. The database encompasses a large variety of current profiles including the hollow current profiles important for reversed central shear operation. The parameters of safety factor profile (a quantity related to the current profile through the magnetic field tilt angle) estimated by the neural network include central safety factor, q0, minimum value of q, qmin, and the location of qmin. Very good performance of the trained neural network both for simulated test data and for experimental data is demonstrated.

Conceptual design study of the moderate size superconducting spherical tokamak power plant

NASA Astrophysics Data System (ADS)

Gi, Keii; Ono, Yasushi; Nakamura, Makoto; Someya, Youji; Utoh, Hiroyasu; Tobita, Kenji; Ono, Masayuki

2015-06-01

A new conceptual design of the superconducting spherical tokamak (ST) power plant was proposed as an attractive choice for tokamak fusion reactors. We reassessed a possibility of the ST as a power plant using the conservative reactor engineering constraints often used for the conventional tokamak reactor design. An extensive parameters scan which covers all ranges of feasible superconducting ST reactors was completed, and five constraints which include already achieved plasma magnetohydrodynamic (MHD) and confinement parameters in ST experiments were established for the purpose of choosing the optimum operation point. Based on comparison with the estimated future energy costs of electricity (COEs) in Japan, cost-effective ST reactors can be designed if their COEs are smaller than 120 mills kW-1 h-1 (2013). We selected the optimized design point: A = 2.0 and Rp = 5.4 m after considering the maintenance scheme and TF ripple. A self-consistent free-boundary MHD equilibrium and poloidal field coil configuration of the ST reactor were designed by modifying the neutral beam injection system and plasma profiles. The MHD stability of the equilibrium was analysed and a ramp-up scenario was considered for ensuring the new ST design. The optimized moderate-size ST power plant conceptual design realizes realistic plasma and fusion engineering parameters keeping its economic competitiveness against existing energy sources in Japan.

Virus-producing cells determine the host protein profiles of HIV-1 virion cores

PubMed Central

2012-01-01

Background Upon HIV entry into target cells, viral cores are released and rearranged into reverse transcription complexes (RTCs), which support reverse transcription and also protect and transport viral cDNA to the site of integration. RTCs are composed of viral and cellular proteins that originate from both target and producer cells, the latter entering the target cell within the viral core. However, the proteome of HIV-1 viral cores in the context of the type of producer cells has not yet been characterized. Results We examined the proteomic profiles of the cores purified from HIV-1 NL4-3 virions assembled in Sup-T1 cells (T lymphocytes), PMA and vitamin D3 activated THP1 (model of macrophages, mMΦ), and non-activated THP1 cells (model of monocytes, mMN) and assessed potential involvement of identified proteins in the early stages of infection using gene ontology information and data from genome-wide screens on proteins important for HIV-1 replication. We identified 202 cellular proteins incorporated in the viral cores (T cells: 125, mMΦ: 110, mMN: 90) with the overlap between these sets limited to 42 proteins. The groups of RNA binding (29), DNA binding (17), cytoskeleton (15), cytoskeleton regulation (21), chaperone (18), vesicular trafficking-associated (12) and ubiquitin-proteasome pathway-associated proteins (9) were most numerous. Cores of the virions from SupT1 cells contained twice as many RNA binding proteins as cores of THP1-derived virus, whereas cores of virions from mMΦ and mMN were enriched in components of cytoskeleton and vesicular transport machinery, most probably due to differences in virion assembly pathways between these cells. Spectra of chaperones, cytoskeletal proteins and ubiquitin-proteasome pathway components were similar between viral cores from different cell types, whereas DNA-binding and especially RNA-binding proteins were highly diverse. Western blot analysis showed that within the group of overlapping proteins, the level of

Setup for potential bias experiments on the Saha Institute of Nuclear Physics tokamak

NASA Astrophysics Data System (ADS)

Ghosh, J.; Pal, R.; Chattopadhyay, P. K.

1999-12-01

An experimental setup for studying the influence of the radial electric field on very low qa plasma on the Saha Institute of Nuclear Physics tokamak is presented. A high current, high voltage pulsed power supply, using a semiconductor controlled rectifier (SCR) as a dc switch is developed and used to bias a tungsten electrode inserted inside the plasma. The electrode's exposed length and its position inside the plasma are controlled by a double bellows assembly to optimize the electrode-exposed length. We show that using the force commutation method to turn the SCR off to get the power pulse desired has good potential for carrying out similar kinds of studies, especially in a low budget small tokamak.

IL-15 regulates memory CD8+ T cell O-glycan synthesis and affects trafficking

PubMed Central

Nolz, Jeffrey C.; Harty, John T.

2014-01-01

Memory and naive CD8+ T cells exhibit distinct trafficking patterns. Specifically, memory but not naive CD8+ T cells are recruited to inflamed tissues in an antigen-independent manner. However, the molecular mechanisms that regulate memory CD8+ T cell trafficking are largely unknown. Here, using murine models of infection and T cell transfer, we found that memory but not naive CD8+ T cells dynamically regulate expression of core 2 O-glycans, which interact with P- and E-selectins to modulate trafficking to inflamed tissues. Following infection, antigen-specific effector CD8+ T cells strongly expressed core 2 O-glycans, but this glycosylation pattern was lost by most memory CD8+ T cells. After unrelated infection or inflammatory challenge, memory CD8+ T cells synthesized core 2 O-glycans independently of antigen restimulation. The presence of core 2 O-glycans subsequently directed these cells to inflamed tissue. Memory and naive CD8+ T cells exhibited the opposite pattern of epigenetic modifications at the Gcnt1 locus, which encodes the enzyme that initiates core 2 O-glycan synthesis. The open chromatin configuration in memory CD8+ T cells permitted de novo generation of core 2 O-glycans in a TCR-independent, but IL-15–dependent, manner. Thus, IL-15 stimulation promotes antigen-experienced memory CD8+ T cells to generate core 2 O-glycans, which subsequently localize them to inflamed tissues. These findings suggest that CD8+ memory T cell trafficking potentially can be manipulated to improve host defense and immunotherapy. PMID:24509081

Heat Deposition and Heat Removal in the UCLA Continuous Current Tokamak

NASA Astrophysics Data System (ADS)

Brown, Michael Lee

1990-01-01

Energy transfer processes in a steady-state tokamak are examined both theoretically and experimentally in order to determine the patterns of plasma heat deposition to material surfaces and the methods of heat removal. Heat transfer experiments involving actively cooled limiters and heat flux probes were performed in the UCLA Continuous Current Tokamak (CCT). The simple exponential model of plasma power deposition was extended to describe the global heat deposition to the first wall of a steady-state tokamak. The heat flux distribution in CCT was determined from measurements of heat flow to 32 large-area water-cooled Faraday shield panels. Significant toroidal and poloidal asymmetries were observed, with the maximum heat fluxes tending to fall on the lower outside panels. Heat deposition to the water-cooled guard limiters of an ion Bernstein wave antenna in CCT was measured during steady-state operation. Very strong asymmetries were observed. The heat distribution varied greatly with magnetic field. Copper heat flux sensors incorporating internal thermocouples were developed to measure plasma power deposition to exterior probe surfaces and heat removal from water -cooled interior surfaces. The resulting inverse heat conduction problem was solved using the function specification method. Cooling by an impinging liquid jet was investigated. One end of a cylindrical copper heat flux sensor was heated by a DC electrical arc and the other end was cooled by a low velocity water jet at 1 atm. Critical heat flux (CHF) values for the 55-80 ^circC sub-cooled free jets were typically 2.5 times published values for saturated free jets. For constrained jets, CHF values were about 20% lower. Heat deposition and heat removal in thick (3/4 inch diameter) cylindrical metal probes (SS304 or copper) inserted into a steady-state tokamak plasma were measured for a broad range of heat loads. The probes were cooled internally by a constrained jet of either air or water. Steady -state heat

Synchronous oscillation prior to disruption caused by kink modes in HL-2A tokamak plasmas

NASA Astrophysics Data System (ADS)

Jiang, M.; Hu, D.; Wang, X. G.; Shi, Z. B.; Xu, Y.; Chen, W.; Ding, X. T.; Zhong, W. L.; Dong, Y. B.; Ji, X. Q.; Zhang, Y. P.; Gao, J. M.; Li, J. X.; Yang, Z. C.; Li, Y. G.; Liu, Y.

2015-08-01

A class of evident MHD activities prior to major disruption has been observed during recent radiation induced disruptions of the HL-2A tokamak discharges. It can be named SOD, synchronous oscillations prior to disruption, characterized by synchronous oscillation of electron cyclotron emission (ECE), core soft x-ray, Mirnov coil, and {{D}α} radiation signals at the divertor plate. The SOD activity is mostly observed in a parametric regime where the poloidal beta is low enough before disruption, typically corresponding to those radiation-induced disruptions. It has been found that the m/n = 2/1 mode is dominant during the SODs, and consequently it is the drop of the mode frequency and the final mode locking that lead to thermal quench. The mode frequency before the mode locking corresponds to the toroidal rotation frequency of the edge plasma. It is also found that during SODs, the location of the q = 2 surface is moving outward, and most of the plasma current is enclosed within the surface. This demonstrates that the current channel lies inside the rational surface during SOD, and thus the resistive kink mode is unstable. Further analysis of the electron temperature perturbation structure shows that the plasma is indeed dominated by the resistive kink mode, with kink-like perturbation in the core plasma region. It suggests that it is the nonlinear growth of the m/n = 2/1 resistive kink mode and its higher order harmonics, rather than the spontaneous overlapping of multiple neighboring islands, that ultimately triggered the disruption.

Stabilization of the Vertical Mode in Tokamaks by Localized Nonaxisymmetric Fields

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

Reiman, A.

Vertical instability of a tokamak plasma can be controlled by nonaxisymmetric magnetic fields localized near the plasma edge at the bottom and top of the torus. The required magnetic fields can be produced by a relatively simple set of parallelogram-shaped coils.

Stability analysis of ELMs in long-pulse discharges with ELITE code on EAST tokamak

NASA Astrophysics Data System (ADS)

Wang, Y. F.; Xu, G. S.; Wan, B. N.; Li, G. Q.; Yan, N.; Li, Y. L.; Wang, H. Q.; Peng, Y.-K. Martin; Xia, T. Y.; Ding, S. Y.; Chen, R.; Yang, Q. Q.; Liu, H. Q.; Zang, Q.; Zhang, T.; Lyu, B.; Xu, J. C.; Feng, W.; Wang, L.; Chen, Y. J.; Luo, Z. P.; Hu, G. H.; Zhang, W.; Shao, L. M.; Ye, Y.; Lan, H.; Chen, L.; Li, J.; Zhao, N.; Wang, Q.; Snyder, P. B.; Liang, Y.; Qian, J. P.; Gong, X. Z.; EAST team

2018-05-01

One challenge in long-pulse and high performance tokamak operation is to control the edge localized modes (ELMs) to reduce the transient heat load on plasma facing components. Minute-scale discharges in H-mode have been achieved repeatedly on Experimental Advanced Superconducting Tokamak (EAST) since the 2016 campaign and understanding the characteristics of the ELMs in these discharges can be helpful for effective ELM control in long-pulse discharges. The kinetic profile diagnostics recently developed on EAST make it possible to perform the pedestal stability analysis quantitatively. Pedestal stability calculation of a typical long-pulse discharge with ELITE code is presented. The ideal linear stability results show that the ELM is dominated by toroidal mode number n around 10–15 and the most unstable mode structure is mainly localized in the steep pressure gradient region, which is consistent with experimental results. Compared with a typical type-I ELM discharge with larger total plasma current (I p = 600 kA), pedestal in the long-pulse H-mode discharge (I p = 450 kA) is more stable in peeling-ballooning instability and its critical peak pressure gradient is evaluated to be 65% of the former. Two important features of EAST tokamak in the long-pulse discharge are presented by comparison with other tokamaks, including a wider pedestal correlated with the poloidal pedestal beta and a smaller inverse aspect ratio and their effects on the pedestal stability are discussed. The effects of uncertainties in measurements on the linear stability results are also analyzed, including the edge electron density profile position, the separatrix position and the line-averaged effective ion charge {Z}{{e}{{f}}{{f}}} value.

Active core profile and transport modification by application of ion Bernstein wave power in the Princeton Beta Experiment-Modification

NASA Astrophysics Data System (ADS)

LeBlanc, B.; Batha, S.; Bell, R.; Bernabei, S.; Blush, L.; de la Luna, E.; Doerner, R.; Dunlap, J.; England, A.; Garcia, I.; Ignat, D.; Isler, R.; Jones, S.; Kaita, R.; Kaye, S.; Kugel, H.; Levinton, F.; Luckhardt, S.; Mutoh, T.; Okabayashi, M.; Ono, M.; Paoletti, F.; Paul, S.; Petravich, G.; Post-Zwicker, A.; Sauthoff, N.; Schmitz, L.; Sesnic, S.; Takahashi, H.; Talvard, M.; Tighe, W.; Tynan, G.; von Goeler, S.; Woskov, P.; Zolfaghari, A.

1995-03-01

Application of Ion Bernstein Wave Heating (IBWH) into the Princeton Beta Experiment-Modification (PBX-M) [Phys. Fluids B 2, 1271 (1990)] tokamak stabilizes sawtooth oscillations and generates peaked density profiles. A transport barrier, spatially correlated with the IBWH power deposition profile, is observed in the core of IBWH-assisted neutral beam injection (NBI) discharges. A precursor to the fully developed barrier is seen in the soft x-ray data during edge localized mode (ELM) activity. Sustained IBWH operation is conducive to a regime where the barrier supports large ∇ne, ∇Te, ∇νφ, and ∇Ti, delimiting the confinement zone. This regime is reminiscent of the H(high) mode, but with a confinement zone moved inward. The core region has better than H-mode confinement while the peripheral region is L(low)-mode-like. The peaked profile enhances NBI core deposition and increases nuclear reactivity. An increase in central Ti results from χi reduction (compared to the H mode) and better beam penetration. Bootstrap current fractions of up to 0.32-0.35 locally and 0.28 overall were obtained when an additional NBI burst is applied to this plasma.

Scoping and sensitivity analyses for the Demonstration Tokamak Hybrid Reactor (DTHR)

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

Sink, D.A.; Gibson, G.

1979-03-01

The results of an extensive set of parametric studies are presented which provide analytical data of the effects of various tokamak parameters on the performance and cost of the DTHR (Demonstration Tokamak Hybrid Reactor). The studies were centered on a point design which is described in detail. Variations in the device size, neutron wall loading, and plasma aspect ratio are presented, and the effects on direct hardware costs, fissile fuel production (breeding), fusion power production, electrical power consumption, and thermal power production are shown graphically. The studies considered both ignition and beam-driven operations of DTHR and yielded results based onmore » two empirical scaling laws presently used in reactor studies. Sensitivity studies were also made for variations in the following key parameters: the plasma elongation, the minor radius, the TF coil peak field, the neutral beam injection power, and the Z/sub eff/ of the plasma.« less

Workshop on High Power ICH Antenna Designs for High Density Tokamaks

NASA Astrophysics Data System (ADS)

Aamodt, R. E.

1990-02-01

A workshop in high power ICH antenna designs for high density tokamaks was held to: (1) review the data base relevant to the high power heating of high density tokamaks; (2) identify the important issues which need to be addressed in order to ensure the success of the ICRF programs on CIT and Alcator C-MOD; and (3) recommend approaches for resolving the issues in a timely realistic manner. Some specific performance goals for the antenna system define a successful design effort. Simply stated these goals are: couple the specified power per antenna into the desired ion species; produce no more than an acceptable level of RF auxiliary power induced impurities; and have a mechanical structure which safely survives the thermal, mechanical and radiation stresses in the relevant environment. These goals are intimately coupled and difficult tradeoffs between scientific and engineering constraints have to be made.

Spatial calibration of a tokamak neutral beam diagnostic using in situ neutral beam emission

DOE PAGES

Chrystal, Colin; Burrell, Keith H.; Grierson, Brian A.; ...

2015-10-20

Neutral beam injection is used in tokamaks to heat, apply torque, drive non-inductive current, and diagnose plasmas. Neutral beam diagnostics need accurate spatial calibrations to benefit from the measurement localization provided by the neutral beam. A new technique has been developed that uses in-situ measurements of neutral beam emission to determine the spatial location of the beam and the associated diagnostic views. This technique was developed to improve the charge exchange recombination diagnostic (CER) at the DIII-D tokamak and uses measurements of the Doppler shift and Stark splitting of neutral beam emission made by that diagnostic. These measurements contain informationmore » about the geometric relation between the diagnostic views and the neutral beams when they are injecting power. This information is combined with standard spatial calibration measurements to create an integrated spatial calibration that provides a more complete description of the neutral beam-CER system. The integrated spatial calibration results are very similar to the standard calibration results and derived quantities from CER measurements are unchanged within their measurement errors. Lastly, the methods developed to perform the integrated spatial calibration could be useful for tokamaks with limited physical access.« less

Spatial calibration of a tokamak neutral beam diagnostic using in situ neutral beam emission

NASA Astrophysics Data System (ADS)

Chrystal, C.; Burrell, K. H.; Grierson, B. A.; Pace, D. C.

2015-10-01

Neutral beam injection is used in tokamaks to heat, apply torque, drive non-inductive current, and diagnose plasmas. Neutral beam diagnostics need accurate spatial calibrations to benefit from the measurement localization provided by the neutral beam. A new technique has been developed that uses in situ measurements of neutral beam emission to determine the spatial location of the beam and the associated diagnostic views. This technique was developed to improve the charge exchange recombination (CER) diagnostic at the DIII-D tokamak and uses measurements of the Doppler shift and Stark splitting of neutral beam emission made by that diagnostic. These measurements contain information about the geometric relation between the diagnostic views and the neutral beams when they are injecting power. This information is combined with standard spatial calibration measurements to create an integrated spatial calibration that provides a more complete description of the neutral beam-CER system. The integrated spatial calibration results are very similar to the standard calibration results and derived quantities from CER measurements are unchanged within their measurement errors. The methods developed to perform the integrated spatial calibration could be useful for tokamaks with limited physical access.

Basic requirements for a 1000-MW(electric) class tokamak fusion-fission hybrid reactor and its blanket concept

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

Hatayama, Ariyoshi; Ogasawara, Masatada; Yamauchi, Michinori

1994-08-01

Plasma size and other basic performance parameters for 1000-MW(electric) power production are calculated with the blanket energy multiplication factor, the M value, as a parameter. The calculational model is base don the International Thermonuclear Experimental Reactor (ITER) physics design guidelines and includes overall plant power flow. Plasma size decreases as the M value increases. However, the improvement in the plasma compactness and other basic performance parameters, such as the total plant power efficiency, becomes saturated above the M = 5 to 7 range. THus, a value in the M = 5 to 7 range is a reasonable choice for 1000-MW(electric)more » hybrids. Typical plasma parameters for 1000-MW(electric) hybrids with a value of M = 7 are a major radius of R = 5.2 m, minor radius of a = 1.7 m, plasma current of I{sub p} = 15 MA, and toroidal field on the axis of B{sub o} = 5 T. The concept of a thermal fission blanket that uses light water as a coolant is selected as an attractive candidate for electricity-producing hybrids. An optimization study is carried out for this blanket concept. The result shows that a compact, simple structure with a uniform fuel composition for the fissile region is sufficient to obtain optimal conditions for suppressing the thermal power increase caused by fuel burnup. The maximum increase in the thermal power is +3.2%. The M value estimated from the neutronics calculations is {approximately}7.0, which is confirmed to be compatible with the plasma requirement. These studies show that it is possible to use a tokamak fusion core with design requirements similar to those of ITER for a 1000-MW(electric) power reactor that uses existing thermal reactor technology for the blanket. 30 refs., 22 figs., 4 tabs.« less

Inner core tilt and polar motion

NASA Astrophysics Data System (ADS)

Dumberry, Mathieu; Bloxham, Jeremy

2002-11-01

A tilted inner core permits exchange of angular momentum between the core and the mantle through gravitational and pressure torques and, as a result, changes in the direction of Earth's axis of rotation with respect to the mantle. We have developed a model to calculate the amplitude of the polar motion that results from an equatorial torque at the inner core boundary which tilts the inner core out of alignment with the mantle. We specifically address the issue of the role of the inner core tilt in the decade polar motion known as the Markowitz wobble. We show that a decade polar motion of the same amplitude as the observed Markowitz wobble requires a torque of 1020 N m which tilts the inner core by 0.07 degrees. This result critically depends on the viscosity of the inner core; for a viscosity less than 5 × 1017 Pa s, larger torques are required. We investigate the possibility that a torque of 1020 N m with decadal periodicity can be produced by electromagnetic coupling between the inner core and torsional oscillations of the flow in the outer core. We demonstrate that a radial magnetic field at the inner core boundary of 3 to 4 mT is required to obtain a torque of such amplitude. The resulting polar motion is eccentric and polarized, in agreement with the observations. Our model suggests that equatorial torques at the inner core boundary might also excite the Chandler wobble, provided there exists a physical mechanism that can generate a large torque at a 14 month period.

Use of Vibratory Coring Samplers for Sediment Surveys.

DTIC Science & Technology

1981-07-01

PAGE ( Man Data R,10#0d,REPORT~~EA INST".T0.P0E’RUCTIONS REPORT DOCUMENTATION PAGE BEFORE COMPLETING FORM I. REPORT NUMBER 2. GOVT ACCESSION NO. 3...COR COARSE 3BRYOZOA S ND AND LARGE SUAND ANSHELS _ CALCAREOUS SANDSTONE ELTH SHELL b CALCAREOUS SANDSTONE PEBBLES SM AND SHELLSb CASTS AND ABUNDAT...GLAUCONITE FINE ANY LMESTOE WITH S SNELL CASTS AND ABUNDANT GLAUCONlITE 1o 4 15-2 5 206 CORE 34 CORE 3S CORE 36 0 D 48 WO 49 WD 26 jFINE SHELF FACIES

Refractive index and strain sensor based on twin-core fiber with a novel T-shaped taper

NASA Astrophysics Data System (ADS)

Zhang, Chuanbiao; Ning, Tigang; Li, Jing; Zheng, JingJing; Gao, Xuekai; Pei, Li

2018-06-01

A compact in-fiber Mach-Zehnder interferometer (MZI) based on twin-core fiber (TCF) with a novel T-shaped taper is proposed and demonstrated. The taper was firstly fabricated by a short section of TCF, and then spliced with a section of cleaved single mode fiber (SMF). When the light transmit into the TCF, multiple modes will be excited and will propagate within the TCF. In experiment, the proposed device had a maximum interferometric extinction ratio about 17 dB. And the refractive index (RI), strain, and temperature response properties of the sensor have been investigated, which show a relatively high RI, strain sensitivity and low temperature cross sensitivity. Hence, the sensor can be a suitable candidate in the biochemical and physical sensing applications. And due to its easy and controllable fabrication, the novel drawing technology can be applied to more multicore optical fibers.

Plasma Physics Lab and the Tokamak Fusion Test Reactor, 1989

ScienceCinema

None

2018-01-16

From the Princeton University Archives: Promotional video about the Plasma Physics Lab and the new Tokamak Fusion Test Reactor (TFTR), with footage of the interior, machines, and scientists at work. This film is discussed in the audiovisual blog of the Seeley G. Mudd Manuscript Library, which holds the archives of Princeton University.

Equilibrium, confinement and stability of runaway electrons in tokamaks

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

Spong, D A

1976-03-01

Some of the ramifications of the runaway population in tokamak experiments are investigated. Consideration is given both to the normal operating regime of tokamaks where only a small fraction of high energy runaways are present and to the strong runaway regime where runaways are thought to carry a significant portion of the toroidal current. In particular, the areas to be examined are the modeling of strong runaway discharges, single particle orbit characteristics of runaways, macroscopic beam-plasma equilibria, and stability against kink modes. A simple one-dimensional, time-dependent model has been constructed in relation to strong runaway discharges. Single particle orbits aremore » analyzed in relation to both the strong runaway regime and the weak regime. The effects of vector E x vector B drifts are first considered in strong runaway discharges and are found to lead to a slow inward shrinkage of the beam. Macroscopic beam-plasma equilibria are treated assuming a pressureless relativistic beam with inertia and using an ideal MHD approximation for the plasma. The stability of a toroidal relativistic beam against kink perturbations is examined using several models. (MOW)« 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

Electronic Structure Calculations and Adaptation Scheme in Multi-core Computing Environments

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

Seshagiri, Lakshminarasimhan; Sosonkina, Masha; Zhang, Zhao

2009-05-20

Multi-core processing environments have become the norm in the generic computing environment and are being considered for adding an extra dimension to the execution of any application. The T2 Niagara processor is a very unique environment where it consists of eight cores having a capability of running eight threads simultaneously in each of the cores. Applications like General Atomic and Molecular Electronic Structure (GAMESS), used for ab-initio molecular quantum chemistry calculations, can be good indicators of the performance of such machines and would be a guideline for both hardware designers and application programmers. In this paper we try to benchmarkmore » the GAMESS performance on a T2 Niagara processor for a couple of molecules. We also show the suitability of using a middleware based adaptation algorithm on GAMESS on such a multi-core environment.« less

Development of the scintillator-based probe for fast-ion losses in the HL-2A tokamak

NASA Astrophysics Data System (ADS)

Zhang, Y. P.; Liu, Yi; Luo, X. B.; Isobe, M.; Yuan, G. L.; Liu, Y. Q.; Hua, Y.; Song, X. Y.; Yang, J. W.; Li, X.; Chen, W.; Li, Y.; Yan, L. W.; Song, X. M.; Yang, Q. W.; Duan, X. R.

2014-05-01

A new scintillator-based lost fast-ion probe (SLIP) has been developed and operated in the HL-2A tokamak [L. W. Yan, X. R. Duan, X. T. Ding, J. Q. Dong, Q. W. Yang, Yi Liu, X. L. Zou, D. Q. Liu, W. M. Xuan, L. Y. Chen, J. Rao, X. M. Song, Y. Huang, W. C. Mao, Q. M. Wang, Q. Li, Z. Cao, B. Li, J. Y. Cao, G. J. Lei, J. H. Zhang, X. D. Li, W. Chen, J. Chen, C. H. Cui, Z. Y. Cui, Z. C. Deng, Y. B. Dong, B. B. Feng, Q. D. Gao, X. Y. Han, W. Y. Hong, M. Huang, X. Q. Ji, Z. H. Kang, D. F. Kong, T. Lan, G. S. Li, H. J. Li, Qing Li, W. Li, Y. G. Li, A. D. Liu, Z. T. Liu, C. W. Luo, X. H. Mao, Y. D. Pan, J. F. Peng, Z. B. Shi, S. D. Song, X. Y. Song, H. J. Sun, A. K. Wang, M. X. Wang, Y. Q. Wang, W. W. Xiao, Y. F. Xie, L. H. Yao, D. L. Yu, B. S. Yuan, K. J. Zhao, G. W. Zhong, J. Zhou, J. C. Yan, C. X. Yu, C. H. Pan, Y. Liu, and the HL-2A Team, Nucl. Fusion 51, 094016 (2011)] to measure the losses of neutral beam ions. The design of the probe is based on the concept of the α-particle detectors on Tokamak Fusion Test Reactor (TFTR) using scintillator plates. The probe is capable of traveling across an equatorial plane port and sweeping the aperture angle rotationally with respect to the axis of the probe shaft by two step motors, in order to optimize the radial position and the collimator angle. The energy and the pitch angle of the lost fast ions can be simultaneously measured if the two-dimensional image of scintillation light intensity due to the impact of the lost fast ions is detected. Measurements of the fast-ion losses using the probe have been performed during HL-2A neutral beam injection discharges. The clear experimental evidence of enhanced losses of beam ions during disruptions has been obtained by means of the SLIP system. A detailed description of the probe system and the first experimental results are reported.

Development of the scintillator-based probe for fast-ion losses in the HL-2A tokamak

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

Zhang, Y. P., E-mail: zhangyp@swip.ac.cn; Liu, Yi; Yuan, G. L.

A new scintillator-based lost fast-ion probe (SLIP) has been developed and operated in the HL-2A tokamak [L. W. Yan, X. R. Duan, X. T. Ding, J. Q. Dong, Q. W. Yang, Yi Liu, X. L. Zou, D. Q. Liu, W. M. Xuan, L. Y. Chen, J. Rao, X. M. Song, Y. Huang, W. C. Mao, Q. M. Wang, Q. Li, Z. Cao, B. Li, J. Y. Cao, G. J. Lei, J. H. Zhang, X. D. Li, W. Chen, J. Chen, C. H. Cui, Z. Y. Cui, Z. C. Deng, Y. B. Dong, B. B. Feng, Q. D. Gao, X. Y.more » Han, W. Y. Hong, M. Huang, X. Q. Ji, Z. H. Kang, D. F. Kong, T. Lan, G. S. Li, H. J. Li, Qing Li, W. Li, Y. G. Li, A. D. Liu, Z. T. Liu, C. W. Luo, X. H. Mao, Y. D. Pan, J. F. Peng, Z. B. Shi, S. D. Song, X. Y. Song, H. J. Sun, A. K. Wang, M. X. Wang, Y. Q. Wang, W. W. Xiao, Y. F. Xie, L. H. Yao, D. L. Yu, B. S. Yuan, K. J. Zhao, G. W. Zhong, J. Zhou, J. C. Yan, C. X. Yu, C. H. Pan, Y. Liu, and the HL-2A Team , Nucl. Fusion 51, 094016 (2011)] to measure the losses of neutral beam ions. The design of the probe is based on the concept of the α-particle detectors on Tokamak Fusion Test Reactor (TFTR) using scintillator plates. The probe is capable of traveling across an equatorial plane port and sweeping the aperture angle rotationally with respect to the axis of the probe shaft by two step motors, in order to optimize the radial position and the collimator angle. The energy and the pitch angle of the lost fast ions can be simultaneously measured if the two-dimensional image of scintillation light intensity due to the impact of the lost fast ions is detected. Measurements of the fast-ion losses using the probe have been performed during HL-2A neutral beam injection discharges. The clear experimental evidence of enhanced losses of beam ions during disruptions has been obtained by means of the SLIP system. A detailed description of the probe system and the first experimental results are reported.« less

Magnetic diagnostics for equilibrium reconstructions with eddy currents on the lithium tokamak experimenta)

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

Schmitt, J. C.; Bialek, J.; Lazerson, S.

2014-11-01

The Lithium Tokamak eXperiment is a spherical tokamak with a close-fitting low-recycling wall composed of thin lithium layers evaporated onto a stainless steel-lined copper shell. Long-lived non-axisymmetric eddy currents are induced in the shell and vacuum vessel by transient plasma and coil currents and these eddy currents influence both the plasma and the magnetic diagnositc signals that are used as constraints for equilibrium reconstruction. A newly installed set of re-entrant magnetic diagnostics and internal saddle flux loops, compatible with high-temperatures and lithium environments, is discussed. Details of the axisymmetric (2D) and non-axisymmetric (3D) treatments of the eddy currents and themore » equilibrium reconstruction are presented.« less

Magnetic diagnostics for equilibrium reconstructions with eddy currents on the lithium tokamak experiment

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

Schmitt, J. C., E-mail: jschmitt@pppl.gov; Lazerson, S.; Majeski, R.

2014-11-15

The Lithium Tokamak eXperiment is a spherical tokamak with a close-fitting low-recycling wall composed of thin lithium layers evaporated onto a stainless steel-lined copper shell. Long-lived non-axisymmetric eddy currents are induced in the shell and vacuum vessel by transient plasma and coil currents and these eddy currents influence both the plasma and the magnetic diagnostic signals that are used as constraints for equilibrium reconstruction. A newly installed set of re-entrant magnetic diagnostics and internal saddle flux loops, compatible with high-temperatures and lithium environments, is discussed. Details of the axisymmetric (2D) and non-axisymmetric (3D) treatments of the eddy currents and themore » equilibrium reconstruction are presented.« less

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.

An active magnetic bearing with high T(sub c) superconducting coils and ferromagnetic cores